ML21210A137

2 to Updated Final Safety Analysis Report, Chapter 5, Containment System (EPID L-2020-LRO-0076) - Redacted

ML21210A137
Person / Time
Site:	Surry
Issue date:	09/30/2020
From:	Virginia Electric & Power Co (VEPCO)
To:	Office of Nuclear Reactor Regulation
Thomas V
Shared Package
ML21208A006	List: ML21208A367 ML21208A409 ML21208A418 ML21208A424 ML21208A429 ML21210A125 ML21210A126 ML21210A127 ML21210A128 ML21210A129 ML21210A133 ML21210A135 ML21210A136 ML21210A137 ML21210A138 ML21210A139 ML21210A140
References
20-325", EPID L-2020-LRO-0076
Download: ML21210A137 (126)

Text

Surry Power Station Updated Final Safety Analysis Report Chapter 5

Intentionally Blank Revision 52Updated Online 09/30/20 SPS UFSAR 5-i Chapter 5: Containment System Table of Contents Section Title Page 5.1 GENERAL DESCRIPTION. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.1-1 5.1 References. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.1-2 5.2 CONTAINMENT ISOLATION. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.2-1 5.2.1 Design Bases . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.2-1 5.2.2 Isolation Design. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.2-4 5.2 References. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.2-8 5.3 CONTAINMENT SYSTEMS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.3-1 5.3.1 Ventilation Systems. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.3-1 5.3.1.1 General Description . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.3-1 5.3.1.2 Design Basis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.3-1 5.3.1.3 System Descriptions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.3-2 5.3.1.4 Design Evaluation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.3-4 5.3.1.5 Tests and Inspections. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.3-5 5.3.2 Leakage-Monitoring System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.3-5 5.3.2.1 Design Basis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.3-5 5.3.2.2 Description. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.3-6 5.3.2.3 Design Evaluation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.3-6 5.3.2.4 Tests and Inspections. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.3-7 5.3.3 Spray Systems . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.3-7 5.3.4 Vacuum System. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.3-7 5.3.4.1 Design Bases . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.3-9 5.3.4.2 Description. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.3-9 5.3.4.3 Evaluation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.3-10 5.3.4.4 Tests and Inspections. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.3-11 5.3.5 Hydrogen Analyzer System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.3-11 5.3 References. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.3-12 5.3 Reference Drawings . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.3-13 5.4 CONTAINMENT DESIGN EVALUATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.4-1 5.4.1 LOCA Mass and Energy Release Analysis. . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.4-2 5.4.1.1 Purpose of Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.4-2 5.4.1.2 System Characteristics and Modeling Assumptions . . . . . . . . . . . . . . . . . . . . 5.4-2 5.4.1.3 Long Term LOCA Mass and Energy Release Analysis . . . . . . . . . . . . . . . . . 5.4-5

Revision 52Updated Online 09/30/20 SPS UFSAR 5-ii Chapter 5: Containment System Table of Contents (continued)

Section Title Page 5.4.1.4 Mass and Energy Releases for Available NPSH Analysis (Hot Leg Double Ended Rupture, Post-Blowdown) . . . . . . . . . . . . . . . . . . . . 5.4-12 5.4.2 LOCA Containment Pressure and Temperature Response . . . . . . . . . . . . . . . . . 5.4-13 5.4.2.1 Containment Response Analytical Model . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.4-14 5.4.3 MSLB Containment Pressure and Temperature Response . . . . . . . . . . . . . . . . . 5.4-22 5.4.3.1 MSLB Mass and Energy Release to Containment . . . . . . . . . . . . . . . . . . . . . 5.4-22 5.4.3.2 MSLB Pressure and Temperature Analysis. . . . . . . . . . . . . . . . . . . . . . . . . . . 5.4-22 5.4 References. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.4-24 5.5 CONTAINMENT TESTS AND INSPECTIONS . . . . . . . . . . . . . . . . . . . . . . . . . 5.5-1 5.5.1 Initial Containment Testing. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.5-1 5.5.2 Continuing Containment Leakage Testing . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.5-1 5.5.3 Containment Integrated Leakage Rate Test (Type A) . . . . . . . . . . . . . . . . . . . . 5.5-2 5.5.4 Containment Penetration Leakage Rate Test (Type B). . . . . . . . . . . . . . . . . . . . 5.5-3 5.5.5 Containment Isolation Valve Leakage Rate Test (Type C) . . . . . . . . . . . . . . . . 5.5-3 5.5.6 Scheduling and Recordkeeping of Periodic Tests. . . . . . . . . . . . . . . . . . . . . . . . 5.5-6 5.5.7 Special Testing Requirements. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.5-6

Revision 52Updated Online 09/30/20 SPS UFSAR 5-iii Chapter 5: Containment System List of Tables Table Title Page Table 5.2-1 Unit 1 Containment Penetrations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.2-9 Table 5.2-2 Unit 2 Containment Penetrations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.2-22 Table 5.3-1 Principal Component Data - Containment Systems. . . . . . . . . . . . . . . . 5.3-14 Table 5.3-2 Containment Cooling Design Heat Loads . . . . . . . . . . . . . . . . . . . . . . . 5.3-15 Table 5.3-3 Leakage-Monitoring System Component Design Data . . . . . . . . . . . . . 5.3-16 Table 5.3-4 Containment Vacuum System Component Design Data . . . . . . . . . . . . 5.3-18 Table 5.4-1 LOCA Mass & Energy Release Analysis System Parameters Initial Conditions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.4-25 Table 5.4-2 LOCA Mass & Energy Release Analysis System Parameters Core Decay Heat Fraction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.4-26 Table 5.4-3 LOCA Mass & Energy Release Analysis Safety Injection Flow Maximum SI - Single Train . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.4-27 Table 5.4-4 LOCA Mass & Energy Release Analysis Safety Injection Flow Minimum SI - Single Train . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.4-28 Table 5.4-5 LOCA Mass & Energy Release Analysis Safety Injection Flow Maximum SI - Two Train . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.4-29 Table 5.4-6 DEHLG, Maximum SI Two Train Mass and Energy Releases for Containment Analysis. . . . . . . . . . . . . . 5.4-30 Table 5.4-7 DEPSG, Minimum SI Single Train Mass and Energy Releases for Containment Analysis. . . . . . . . . . . . . . 5.4-39 Table 5.4-8 Thermophysical Properties of Passive Heat Sink Materials . . . . . . . . . 5.4-42 Table 5.4-9 Passive Heat Sinks . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.4-43 a

Table 5.4-10 Containment LOCA Analysis Initial Conditions . . . . . . . . . . . . . . . . . 5.4-44 Table 5.4-11 Containment LOCA Analysis Peak Pressure Results . . . . . . . . . . . . . . 5.4-44 Table 5.4-12 Containment Depressurization Results DEPSG . . . . . . . . . . . . . . . . . . 5.4-44 Table 5.4-13 Accident Chronology DEPSG, Minimum ESF . . . . . . . . . . . . . . . . . . . 5.4-45 Table 5.4-14 MSLB Containment Peak Pressure Analysis. . . . . . . . . . . . . . . . . . . . . 5.4-46 Table 5.4-15 Accident Chronology MSLB Peak Pressure Analysis. . . . . . . . . . . . . . 5.4-46 Table 5.4-16 MSLB Containment Peak Temperature Analysis . . . . . . . . . . . . . . . . . 5.4-47 Table 5.4-17 Key Parameters in the Containment Analysis . . . . . . . . . . . . . . . . . . . . 5.4-48

Revision 52Updated Online 09/30/20 SPS UFSAR 5-iv Chapter 5: Containment System List of Figures Figure Title Page Figure 5.2-1 Example of Valve Arrangement Containment Isolation System . . . . . 5.2-38 Figure 5.3-1 Typical Containment Pressure Transient Curves; Surry Power Station 5.3-19 Figure 5.3-2 Containment Vacuum System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.3-20 Figure 5.4-1 Containment Pressure DEHLG Peak Pressure Analysis . . . . . . . . . . . 5.4-50 Figure 5.4-2 Containment Vapor Temperature DEHLG Peak Pressure Analysis . . 5.4-50 Figure 5.4-3 Containment Pressure DEPSG Depressurization Analysis . . . . . . . . . 5.4-51 Figure 5.4-4 Containment Temperatures DEPSG Depressurization Analysis . . . . . 5.4-51 Figure 5.4-5 Total RS Heat Exchanger Heat Rate DEPSG Depressurization Analysis 5.4-52 Figure 5.4-6 Containment Pressure 1.4 ft2 MSLB Peak Pressure Analysis. . . . . . . 5.4-52 Figure 5.4-7 Containment Temperature 1.4 ft2 MSLB Peak Pressure Analysis . . . 5.4-53 Figure 5.4-8 Containment Pressure 0.6 ft2 MSLB Peak Temperature Analysis . . . 5.4-53 Figure 5.4-9 Containment Temperature 0.6 ft2 MSLB Peak Temperature Analysis 5.4-54

Revision 52Updated Online 09/30/20 SPS UFSAR 5.1-1 CHAPTER 5 CONTAINMENT SYSTEM Volume II This section describes the containment system for either unit. The containment systems for the two units are similar and completely independent.

Note: As required by the Renewed Operating Licenses for Surry Units 1 and 2, issued March 20, 2003, various systems, structures, and components discussed within this chapter are subject to aging management. The programs and activities necessary to manage the aging of these systems, structures, and components are discussed in Chapter 18.

5.1 GENERAL DESCRIPTION The containment system, together with the engineered safeguards (Chapter 6), is designed to limit radiation doses under conditions resulting from design-basis accident (Chapter 14) to less than or equal to the limits specified in 10 CFR 50.67 at the site boundary and beyond.

The steel-lined, reinforced-concrete containment structure, including foundations, access openings, and penetrations are designed and constructed to maintain full containment integrity when subjected to the temperatures, pressures, potential missiles resulting from the design-basis accident, and the earthquake conditions and tornados described in Chapter 2. Systems are provided to remove heat from the containment and to ensure against breaching containment integrity at the time of, or following, the design-basis accident, or any lesser accident.

The original containment concept includes provisions for routine operation at a reduced internal pressure in which the air partial pressure varies between about 9.0 and 10.3 psia, and for the return to subatmospheric pressure within 60 minutes after the design-basis accident through the use of multiple spray systems. This concept provides for positive termination of outleakage of fission products from the containment, since the containment is maintained at subatmospheric pressure after depressurization. The pressure following depressurization is maintained at less than 14.7 psia. The current concept for the design basis accident containment internal pressure reduction, consistent with alternate source term (AST) analysis, is discussed in Section 5.4.

Provisions have been made for the leak testing of liner seams during construction; for air pressure and leak testing of the containment structure at the completion of construction; for leak testing of the penetrations and access openings at any time; for continuous leak monitoring of the containment structure while at subatmospheric pressure; and for periodic pressure testing of the containment structure throughout station life.

Leak tightness testing of liner welds during construction was performed by welding a structural steel gas test channel over each weld. The test channels of the dome of the containment are located outside the liner plate, with test holes tapped and plugged from the inside. Test channels of the floor liner are piped through the concrete, which covers the floor, to test port panels, and are plugged. Test channels on the straight side walls of the containment are located inside the liner plate, and are tapped and plugged from the inside.

Revision 52Updated Online 09/30/20 SPS UFSAR 5.1-2 The containment weld test channels used for leak testing of the containment liner welds as described in Section 5.5 were left in place at the completion of construction. Therefore, the test channel welds are tested as an integral part of the containment liner plate welds during the operational phase leak rate testing.

Although the leak test channels were not designed as structural elements, nor as a pressure boundary, they do provide additional leak protection. The test channels are capable of withstanding all loads that might be imposed on them during normal, test, and design basis accident conditions without any loss of function, and the presence of the test channels does not in any way impair the performance of the containment liner itself.

Details of containment structural design are given in Chapter 15, and details related to performance during postulated accident situations are given in Section 5.4 and Chapter 14.

5.1 REFERENCES

1. Virginia Power letter dated August 5, 1988, (Serial No. 88-707B), Containment Liner Test Channels.

2. NRC SER dated March 6, 1989, (Serial No.89-184), Surry Units 1 & 2 Containment Liner Weld Leak Chase Channels.

Revision 52Updated Online 09/30/20 SPS UFSAR 5.2-1 5.2 CONTAINMENT ISOLATION 5.2.1 Design Bases The containment isolation system has the following design bases:

1. During incident conditions, at least two barriers exist between the atmosphere outside the containment structure and:

a. The atmosphere inside the containment structure.

b. The reactor coolant and connecting systems.

2. The design pressure of all piping and connecting components within the isolation boundary is greater than the design pressure of the containment, 45 psig.

3. The failure of one valve or barrier does not prevent isolation.

4. The operation of the containment isolation system is automatic.

5. All isolation valves and equipment are protected from missiles and water jets originating from the reactor coolant system.

6. All remotely actuated and automatically operated isolation valves have their positions indicated in, and can be operated from, the control room.

7. Containment isolation system valves are located so as to require a minimum length of piping between the isolation valves and their penetrations.

8. Special consideration is given to the design of the low-head safety injection and recirculation spray pump inlet lines, in that highly reliable components are used in a single valve arrangement, which is enclosed in a special valve pit.

For isolation, the two-barrier valving arrangements consist of the following:

1. Two automatic isolation valves one on each side of the containment wall.

2. Two automatic isolation valves located outside the containment wall.

3. An automatic isolation valve and a membrane barrier. A membrane barrier consists of either pipe, tubing, or a component wall.

4. An administratively-controlled, manually-operated valve outside the containment, and a closed system inside the containment.

5. Two administratively-controlled, manually-operated valves, one on each side of the containment wall.

Revision 52Updated Online 09/30/20 SPS UFSAR 5.2-2

6. A sump recirculation pipe and valve arrangement, conservatively designed and fabricated, and enclosed by a special valve pit. The suction lines for the low-head safety injection pumps and the recirculation-spray pumps are designed to prevent gross system leakage. The major portion of this piping is buried in the reinforced-concrete base mat, and only a short length of piping exists between the mat and the isolation valve. This valve is equipped with a reliable remote operator. The design of this portion of the installation is compatible with letters from the Advisory Committee on Reactor Safeguards to the U.S. Atomic Energy Commission (References 1 & 2). Provisions for detecting leaks in these suction lines are described in Chapter 6.

The criteria applied to the various functional classes of piping to implement the design bases are as follows:

1. Class I piping is open to the outside atmosphere and is connected to the reactor coolant system, or a connecting system, or is open to the containment atmosphere. An example is the line from the containment sump pumps to the waste drain tanks. For Class I piping, the following is provided for isolation subsequent to a LOCA:

a. Incoming lines with one check valve inside the containment wall and an automatic isolation valve outside the containment.

b. Outgoing lines with one automatic isolation valve inside and one automatic isolation valve outside the containment wall or two automatic isolation valves outside the containment wall.

2. Class II piping is connected to a closed system outside the containment and is connected to the reactor coolant system, or a connecting system, or is open to the containment atmosphere.

An example is the excess letdown line. For Class II piping, the following is provided for isolation subsequent to a LOCA:

a. Incoming lines with one check valve inside the containment wall and one automatic isolation valve outside the containment wall.

b. Outgoing lines with one automatic isolation valve.

3. Class III piping is connected to open systems outside the containment and is separated from the reactor coolant system, or a connecting system, and the containment atmosphere by a valve under administrative control or by a membrane barrier. Examples are the component cooling-water lines. For Class III piping, the following is provided for isolation subsequent to a LOCA:

a. Incoming lines with one check valve inside the containment wall and a valve under administrative control outside the containment wall.

b. Outgoing lines with one automatic isolation valve or a valve under administrative control outside the containment wall.

Revision 52Updated Online 09/30/20 SPS UFSAR 5.2-3 In the case of the main feedwater and auxiliary feedwater systems, isolation of the Class III lines is provided by one check valve inside and one check valve outside the containment wall. Isolation of the steam supply to the turbine driven auxiliary feedwater pump is provided by a normally open manual valve under procedural control. Isolation of the service water lines to the Recirculation Spray Heat Exchangers is provided by the closed membrane system inside of the containment wall and a remote manual isolation valve outside the containment wall in each line.

4. Class IV piping must remain open after a LOCA. An example is the high-head safety injection/charging pump header to the reactor coolant system. For Class IV piping, the following is provided for isolation subsequent to a LOCA:

a. Incoming lines with one check valve inside the containment wall and one remote manual valve outside the containment wall.

b. Outgoing lines with one automatic isolation valve outside the containment wall.

Isolation for the seal water supply to the Reactor Coolant Pumps (RCP) is provided by two check valves inside the containment wall and one administratively controlled manual valve outside the containment. Isolation barriers are provided by the check valves inside containment and the closed portion of the chemical and volume control system on the discharge of the charging pumps. These lines remain open after a safety injection signal, and the flow contributes to the total injection flow while cooling the RCP seals.

5. Class V piping is connected to systems outside the containment wall, which are normally not in service and are isolated by a normally closed isolation valve under administrative control.

A Class V line is separated from the reactor coolant system, connecting systems, and the containment atmosphere by a closed valve and/or by a membrane barrier. An example is the service air line.

The isolation valve configuration for each penetration is provided in Tables 5.2-1 and 5.2-2 for Units 1 and 2, respectively.

Where check valves are used as isolation valves, consideration has been given to the ability of these check valves to prevent the leakage of air into the containment when the containment atmospheric pressure is negative.

Check valves in the containment spray and recirculation spray systems are positive-closure check valves. These valves have an external, adjustable counterweight, set to maintain the disk tightly sealed during certain phases of accident conditions when the containment atmospheric pressure is slightly negative. These types of check valves are provided in these systems because they are open to the containment atmosphere through the spray nozzles when the systems are isolated after an accident.

Check valves used for isolation purposes in other pipelines, normally those containing water, are ordinary check valves. They do not have the positive closure feature because they are in

Revision 52Updated Online 09/30/20 SPS UFSAR 5.2-4 series with an automatic trip valve or a valve under administrative control. This arrangement would require a double failure; that is, a failure of the automatic trip valve to close and a rupture in the line downstream from the check valve, which would cause the water leg normally holding the check valve closed and sealed to be drained. This allows outward leakage past the check valve if the check valve fails to seal tightly with a small differential air pressure.

A monitoring arrangement is provided to test the leaktightness of each automatically actuated trip valve and check valve. Examples of valve arrangements for each class of penetration are depicted in Figure 5.2-1.

Instrumentation and adjunct control circuits associated with automatic isolation valve closure are fail-safe (initiate closure) upon loss of voltage and/or control air. Most isolation valves are air-to-open/spring-return-closure diaphragm-operated, piston-operated or direct acting electric solenoid valves thus providing a fail-safe design. The automatic isolation valves inside the containment will function properly under all containment atmospheric pressures.

Under accident conditions, the containment pressure is positive and the solenoid valve vents the control air to the containment atmosphere. Because both sides of the isolation valve diaphragm are vented, balanced forces on either side of the diaphragm result, allowing the spring to close the automatic isolation valve. Circuits that control redundant automatic valves are redundant in the sense that no single failure will preclude isolation. Means are provided to periodically test the functioning of the automatic isolation equipment such as the setpoint of sensors, the speed of response, and the operability of fail-safe features. The containment isolation instrumentation is discussed in Section 7.5.

It should be noted that isolation valves actuated by electric motors upon electrical failure fail in the as-is position.

The trip valves in the reactor coolant sample system and the residual heat removal sample systems are direct acting electric solenoid valves. This ensures that the valves could be reopened to draw a sample under single failure criteria, after an accident.

The steam generator blowdown trip valves are 2-inch, double disk, pressure seal-type gate valves. The valves are of sufficient size to meet the maximum allowable pressure-drop requirement at the design flow rate and will minimize the occurrence of cavitation.

5.2.2 Isolation Design The general criteria covering the number and location of isolation valves required to ensure containment integrity during LOCA conditions are provided in Section 5.2.1. Tables 5.2-1 and 5.2-2 for Units 1 and 2, respectively, summarizes the major piping penetrations through the containment for each fluid system as to the type of valves that are provided, their position under various plant conditions, the fluid they contain and the systems they connect. The tables also identify if the system is essential or non-essential and the isolation actuation signals. In addition,

Revision 52Updated Online 09/30/20 SPS UFSAR 5.2-5 the tables identify those valves that are required to be leak tested in accordance with 10 CFR 50 Appendix J.

The isolation valves tested in accordance with 10 CFR 50 Appendix J provide containment integrity during LOCA conditions. The remaining isolation valves, which are not required to be tested in accordance with 10 CFR 50 Appendix J, provide isolation to mitigate the consequences of other accident conditions (e.g., mainsteam, feedwater and blowdown valves are not tested, but provide isolation in the event of a steam generator tube rupture).

Containment isolation is accomplished under the following conditions:

1. Phase-1 isolation is initiated by a safety injection actuation signal. Safety injection is actuated by any one of the following input signals (Section 7.5):

a. High steam-line flow with low steam-line pressure or low-low Tavg.

b. High steam-line differential pressure.

c. Low-low pressurizer pressure.

d. High containment pressure.

e. Manual initiation.

These input signals provide the diversity required by Section 6.2.4 of the Standard Review Plan (Reference 3).

2. Phase 2 - Isolation is initiated by a high containment pressure signal, and closes the automatic trip valves in all normally open lines penetrating the containment that are not required to be open to control containment pressure to perform an orderly shutdown without actuation of the consequence limiting safeguards in case of a small reactor coolant system leak.

3. Phase 3 - Isolation is initiated by a high-high containment pressure signal, which is indicative of a major LOCA. Remaining automatic trip valves normally open lines that penetrate the containment which have not been shut by 2. above are shut by this signal.

Plant systems with containment penetrations have been categorized as essential or non-essential. There are, in turn, two levels of essential systems:

Level 1 - Systems required to mitigate the consequences of an accident.

Level 2 - Systems required to maintain the operability of critical systems or functions.

Level 1 essential systems (Tables 5.2-1 & 5.2-2) include the engineered safety features (such as containment spray, recirculation spray, and the safety injection system) and the service-water system used to cool the recirculation-spray heat exchangers. Level 2 essential systems (Tables 5.2-1 & 5.2-2) include the auxiliary feedwater system, the component cooling-water system associated with reactor coolant pump operation, containment air cooling,

Revision 52Updated Online 09/30/20 SPS UFSAR 5.2-6 and residual heat removal. Non-essential systems (Tables 5.2-1 & 5.2-2) include the other systems not required for the Level 1 and Level 2 functions described above.

Level 1 essential systems are required to operate after a LOCA. Level 2 essential systems remain unisolated from containment unless they are not required, or until a LOCA is indicated by Phase 2 isolation. Non-essential systems are either isolated during normal operation or they are isolated by a Phase 1 isolation signal. Some non-essential systems may be operated manually following a LOCA if conditions warrant their use.

Once Phase 1 containment isolation has been initiated by a safety injection actuation signal, the automatic isolation valves can be opened only after the manual reset of the actuating signal and the deliberate remote manual operation of the individual valve (an exception is the condenser air ejector containment isolation valve described in the next paragraph). There are no valve control switches that control the reopening of more than one valve.

Under normal conditions, the condenser air ejector discharge is vented to the atmosphere and the containment discharge divert valve is closed. When high radioactivity is detected by the condenser air ejector radiation monitors, the normal condenser air ejector discharge flow path to atmosphere is isolated and the containment divert valve opens to divert the condenser air ejector discharge to containment. If a containment isolation (safety injection) signal occurs, the condenser air ejector containment isolation trip valve will close. When the containment isolation signal resets, the condenser air ejector containment isolation trip valve will open and divert the condenser air ejector discharge back to the containment if the high radioactivity signal is still present. The isolation valve has an electrical interlock, however, that prevents reset until containment pressure is subatmospheric. Normal flow to the atmosphere is not restored until the high radiation signal is cleared. See Sections 10.3.8.2 and 11.3.3.8 for further information.

Diverse isolation signals are provided for the automatic containment isolation valves in non-essential systems. However, some non-essential systems are not automatically isolated by a containment isolation signal. But the staff of the Nuclear Regulatory Commission has agreed that sufficient isolation provisions have been provided at Surry for all non-essential penetrations (Reference 4). Penetrations with normally closed manual isolation valves are locked closed and administratively controlled such that, if a valve is required to be opened during plant operation, a dedicated person is assigned to close it after the evolution requiring it to be open has been completed, or to close it within 60 seconds after the receipt of a containment isolation signal.

The basis for the 60-second limit is that no fuel cladding is expected to melt or fail until after 60 seconds following a loss-of-coolant accident (LOCA). This is verified for PWRs by the FLECHT experimental results (Reference 7). Thus, fission product release from the core to the containment atmosphere or to other portions of the Reactor Coolant System (RCS) could not occur until at least one minute after the event.

If any of the automatic signals fail to actuate the containment isolation trip valves or the remote manual valves, isolation can be accomplished manually from the control room. The

Revision 52Updated Online 09/30/20 SPS UFSAR 5.2-7 solenoid valves that operate the automatic trip valves can be actuated by an electric signal that is produced in the control room.

For lines coming into the containment, check valves are used wherever an additional barrier is provided by either a membrane or an automatic isolation valve. The use of check valves in this service is confined to either liquid lines or lines that are closed outside the containment. These check valves shut under a differential pressure when the higher pressure is on the containment side of the check valve.

The monitoring arrangement provided to test the leaktightness of each automatic trip valve and check valve consists of a monitoring tap on the main line upstream from each isolation valve.

To test for valve tightness, the main piping section upstream from each valve is pressurized and evidence of fluid leakage is checked using the makeup air method. When not in use, the monitoring lines are plugged at the open end. As described in Section 5.5, containment isolation valves are tested to verify their sealing capability and leaktightness.

Several spare containment pipe penetrations of various sizes are provided. All pipes in these spare penetrations are sealed at both ends.

All isolation valves and equipment are protected from missiles and water jets originating from the reactor coolant system. Missile protection for isolation valves, actuators, and controls is provided by locating isolation valves between the steam generator cubicle wall, crane wall, and the containment wall or locating isolation valves outside the containment structure. The pressure-sensing devices that detect high containment pressure are located outside the containment on the leakage-monitoring tubing that is open to the containment. The location of the pressure-sensing devices outside the containment protects them from missiles developed by a LOCA. Details regarding the probability of missile damage and design features to prevent the formation of missiles are given in Section 15.5.1.11.

The fuel transfer penetration between the refueling canal inside the containment and the spent-fuel pit is fitted with a blind flange inside the containment and a normally closed gate valve in the transfer canal outside the containment to prevent leakage through the transfer tube during accident conditions.

The following precautions, which apply to all lines penetrating the containment, are intended to prevent the inadvertent opening of these lines to the atmosphere outside the containment:

1. Automatic isolation valves can be opened only upon the cessation of the actuating signal and the manual reset of controls.

2. Automatic isolation valves are capable of manual actuation from the control room, with the limitations on the opening of the valve discussed in item 1. above.

3. Remote manual valves are closed and opened only under administrative control.

Revision 52Updated Online 09/30/20 SPS UFSAR 5.2-8

4. Local manual valves are closed and opened under administrative control.

5. Check valves open only when the fluid pressure is higher on the side outside the containment.

6. The design pressure of piping and connecting components within the isolation boundary is greater than the design pressure of the containment (45 psig).

7. Remote manual valves, once opened by a high-containment-pressure isolation signal, can only be closed upon the cessation of the actuating signal and the manual reset of controls.

For items 1, 2, 3, and 4 above, and for flanged closures, specific plant procedures define the positioning of these closures in the containment isolation system during normal operation, shutdown, and accident conditions.

5.2 REFERENCES

1. Letter from S. H. Hanauer, Advisory Committee on Reactor Safeguards, to G. T. Seaborg, AEC,

Subject:

Report on Edwin I. Hatch Nuclear Plant, dated May 15, 1969.

2. Letter from S. H. Hanauer, Advisory Committee on Reactor Safeguards, to G. T. Seaborg, AEC,

Subject:

Report on Brunswick Steam Electric Plant.

3. Standard Review Plan, Section 6.2.4, Containment Isolation System.

4. U.S. Nuclear Regulatory Commission, Evaluation of Licensees Compliance with Category A Items of NRC Recommendations Resulting from TMI-2 Lessons Learned, April 24, 1980.

5. U.S. Nuclear Regulatory Commission, Exemption from Appendix J, 10 CFR 50 for Surry Unit 2, dated November 21, 1988.

6. U.S. Nuclear Regulatory Commission, Exemption from Appendix J, 10 CFR 50 for Surry Unit 1, dated August 7, 1990.

7. Westinghouse Report WCAP-7544, PWR FLECHT Group II Test Report, September 1970.

Revision 52Updated Online 09/30/20 Table 5.2-1 UNIT 1 CONTAINMENT PENETRATIONS 10 CFR 50 Essential Line Penetration Configuration Appendix J System Nominal Class Valve Number/Type/Operator Test Required Level Line Statusc, d System Pen Line Per 1, 2, ESF Actuation Shut- Loss of Name No. Size UFSAR Inside Outside Inside Outside or No System Signalf Normal down Incident Pwr (I/O)e Flow Temperatureb Fluid Notes CCW from B 1-CC-TV-109B 001 18" 3 None - No 2 No SI C -/FC Out Cold Liquid 1 RHR HX Butterfly/Air Pilot CCW to A 1-CC-177 1-CC-214 002 18" 3 No No 2 No - O O C NA In Cold Liquid 1 RHR HX Check Butterfly/Manual CCW to B 1-CC-176 1-CC-220 004 18" 3 No No 2 No - O O C NA In Cold Liquid 1 RHR HX Check Butterfly/Manual CCW from A 1- CC-TV-109A 005 18" 3 None - No 2 No SI C NA/FC Out Cold Liquid 1 RHR HX Butterfly/Air Pilot 1-SI-150 No Noa 1 Yes SI LC LC LC NA/LC In Cold Liquid 2, 4, 9 Globe/Manual High Head SI 1-SI-225 1-SI-MOV-1867D 007 3" 4 - Noa C C O -/FAI (Normal header) Check Gate/Motor 1-SI-MOV-1867C

- Noa C C O -/FAI Gate/Motor CCW to C Air 1-CC-224 1-CC-223 009 6" 3 No No 2 No - O O C NA In Cold Liquid 1 Recirc. Fan Check Gate/Manual CCW to B Air 1-CC-233 1-CC-232 010 6" 3 No No 2 No - O O C NA In Cold Liquid 1 Recirc. Fan Check Gate/Manual CCW to A Air 1-CC-242 1-CC-241 011 6" 3 No No 2 No - O O C NA In Cold Liquid 1 SPS UFSAR Recirc. Fan Check Gate/Manual CCW from B 1-CC-TV-110B 012 6" 3 None - No 2 No HH O O C -/FC Out Cold Liquid 1 Air Recirc. Fan Plug/Air Pilot CCW from C 1-CC-TV-110C 013 6" 3 None - No 2 No HH O O C -/FC Out Cold Liquid 1 Air Recirc Fan Plug/Air Pilot CCW from A 1-CC-TV-110A 014 6" 3 None - No 2 No HH O O C -/FC Out Cold Liquid 1 Air Recirc Fan Plug/Air Pilot Chemical and 1-CH-309 1-CH-MOV-1289A 015 3" 2 No Noa No No SI O O C NA/FAI In Cold Liquid 9 Volume Control Check Gate/Motor CCW to C 1-CC-59 1-CC-216 016 6" 3 No No 2 No - O O C NA In Cold Liquid 1 RCP Check Gate/Manual

a. See Note 9

b. Cold < 250°F; Hot > 250°F

c. O = Open; C = Closed; Int = Intermediate; LC = Locked Closed

d. FC = Fails Closed; FAI = Fails As Is; NA = Not Applicable

e. I/O = Inside/Outside 5.2-9

f. HH = Phase 3, H = Phase 2, SI = Phase 1

Table 5.2-1 (CONTINUED)

Revision 52Updated Online 09/30/20 UNIT 1 CONTAINMENT PENETRATIONS 10 CFR 50 Essential Line Penetration Configuration Appendix J System Nominal Class Valve Number/Type/Operator Test Required Level Line Statusc, d System Pen Line Per 1, 2, ESF Actuation Shut- Loss of Name No. Size UFSAR Inside Outside Inside Outside or No System Signalf Normal down Incident Pwr (I/O)e Flow Temperatureb Fluid Notes CCW to B 1-CC-58 1-CC-218 017 6" 3 No No 2 No - O O C NA In Cold Liquid 1 RCP Check Gate/Manual CCW to A 1-CC-1 1-CC-219 018 6" 3 No No 2 No - O O C NA In Cold Liquid 1 RCP Check Gate/Manual Seal Water from 1-CH-MOV-1381 019 3" 2 None - Yes No No SI O O C -/FAI Out Cold Liquid RCPs Gate/Motor Safety Injection 1-SI-HCV-1851A 1-SI-32 Accumulator 020 1" 5 No Yes No No - LC LC LC FC/NA In Cold Liquid 3, 4 Globe/Air Pilot Globe/Manual Makeup 1-SI-HCV-1851B No - FC/-

Globe/Air Pilot 1-SI-HCV-1851C No - FC/-

Globe/Air Pilot High Head Safety Injection 1-SI-224 1-SI-MOV-1842 to Cold Leg 021 3" 4 No Noa 1 Yes - C C INT NA/FAI In Cold Liquid 2, 9 Check Gate/Motor (Alternate header)

High Head 1-SI-226 1-SI-MOV-1869B Safety Injection 023 3" 4 No Noa 1 Yes - C C INT NA/FAI In Cold Liquid 2, 9 Check Gate/Motor SPS UFSAR to Hot Leg 1-RH-47 1-RH-100 RHR to RWST 024 6" 5 Yes Yes No No - LC INT LC NA/FAI Out Cold Liquid 3, 4 Gate/Manual Gate/Manual CCW from A 1-CC-TV-105A 025 6" 3 None - No 2 No HH O O C -/FC Out Cold Liquid 1 RCP Plug/Air Pilot CCW from C 1-CC-TV-105C 026 6" 3 None - No 2 No HH O O C -/FC Out Cold Liquid 1 RCP Plug/Air Pilot CCW from B 1-CC-TV-105B 027 6" 3 None - No 2 No HH O O C -/FC Out Cold Liquid 1 RCP Plug/Air Pilot Reactor Coolant 1-CH-TV-1204A 1-CH-TV-1204B 028 2" 2 Yes Yes No No SI O O C FC/FC Out Hot Liquid Letdown Globe/Air Pilot Globe/Air Pilot

a. See Note 9

b. Cold < 250°F; Hot > 250°F

c. O = Open; C = Closed; Int = Intermediate; LC = Locked Closed

d. FC = Fails Closed; FAI = Fails As Is; NA = Not Applicable 5.2-10

e. I/O = Inside/Outside

f. HH = Phase 3, H = Phase 2, SI = Phase 1

Table 5.2-1 (CONTINUED)

Revision 52Updated Online 09/30/20 UNIT 1 CONTAINMENT PENETRATIONS 10 CFR 50 Essential Line Penetration Configuration Appendix J System Nominal Class Valve Number/Type/Operator Test Required Level Line Statusc, d System Pen Line Per 1, 2, ESF Actuation Shut- Loss of Name No. Size UFSAR Inside Outside Inside Outside or No System Signalf Normal down Incident Pwr (I/O)e Flow Temperatureb Fluid Notes 3/8 valve, 1-GW-TV-106 Gaseous Waste 032 1 None - Yes No No - C C C -/FC In Cold Gas 3" penetration Globe/Air Pilot 1-GW-TV-107

- Yes Globe/Air Pilot Primary Drain 1-DG-TV-108A 1-DG-TV-108B Transfer Tank 033 2" 1 Yes Yes No No SI O O C FC/FC Out Cold Liquid Globe/Air Pilot Globe/Air Pilot Pump Discharge Seal Water to 1-CH-349 1-CH-300 035 2" 4 No No 1 No - O O O NA In Cold Liquid 2, 5 C RCP Check Needle/Manual Seal Water to 1-CH-323 1-CH-294 036 2" 4 No No 1 No - O O O NA In Cold Liquid 2, 5 A RCP Check Needle/Manual Seal Water to 1-CH-333 1-CH-297 037 2" 4 No No 1 No - O O O NA In Cold Liquid 2, 5 B RCP Check Needle/Manual Aerated Drain 1-DA-TV-100A 1-DA-TV-100B Sump Pump 038 2" 1 Yes Yes No No SI O O C FC/FC Out Cold Liquid Ball/Air Pilot Ball/Air Pilot Discharge SG 1-BD-TV-100A 1-BD-TV-100B 039 3" 3 No No No No AFW O O C FC/FC Out Hot Liquid 7 Blowdown-1A Gate/Air Pilot Gate/Air Pilot SG 1-BD-TV-100E 1-BD-TV-100F 040 3" 3 No No No No AFW O O C FC/FC Out Hot Liquid 7 Blowdown-1C Gate/Air Pilot Gate/Air Pilot SPS UFSAR SG 1-BD-TV-100C 1-BD-TV-100D 041 3" 3 No No No No AFW O O C FC/FC Out Hot Liquid 7 Blowdown-1B Gate/Air Pilot Gate/Air Pilot 1-SA-60

- Yes No No - LC INT LC NA In Cold Gas 4 Service Air Gate/Manual 042 2" 5 None Supply 1-SA-62

- Yes Gate Manual Particulate and Gaseous Rad 1-RM-3 1-RM-TV-100A 043 1" 1 Yes Yes No No H O O C NA/FC In Cold Gas Monitoring Check Globe/Air Pilot Return

a. See Note 9

b. Cold < 250°F; Hot > 250°F

c. O = Open; C = Closed; Int = Intermediate; LC = Locked Closed

d. FC = Fails Closed; FAI = Fails As Is; NA = Not Applicable

e. I/O = Inside/Outside 5.2-11

f. HH = Phase 3, H = Phase 2, SI = Phase 1

Table 5.2-1 (CONTINUED)

Revision 52Updated Online 09/30/20 UNIT 1 CONTAINMENT PENETRATIONS 10 CFR 50 Essential Line Penetration Configuration Appendix J System Nominal Class Valve Number/Type/Operator Test Required Level Line Statusc, d System Pen Line Per 1, 2, ESF Actuation Shut- Loss of Name No. Size UFSAR Inside Outside Inside Outside or No System Signalf Normal down Incident Pwr (I/O)e Flow Temperatureb Fluid Notes Particulate and Gaseous Rad 1-RM-TV-100C 1-RM-TV-100B 044 1" 1 Yes Yes No No H O O C FC/FC Out Cold Gas Monitoring Globe/Air Pilot Globe/Air Pilot Supply Primary Grade 1-RC-160 1-RC-TV-1519A 045 3" 1 Yes Yes No No SI INT C C NA/FC In Cold Liquid Water to PRT Check Diaphragm/Air Pilot 1-RC-HCV-1556A No Noa No No - C INT C FC/FC In Cold Liquid 3, 9 Globe/Air Pilot Loop Fill 1-RC-HCV-1556B 1-CH-FCV-1160 046 2" 5 No - FC/-

Header Globe/Air Pilot Globe/Air Pilot 1-RC-HCV-1556C No - FC/-

Globe/Air Pilot 1-IA-TV-100 Yes Yes No No HH O O C NA/FC In Cold Gas 4 Instrument Air 1-IA-939 Globe/Air Pilot 047 2" 1 Supply Check 1-IA-446

- Yes INT/LC O INT/LC -/NA Gate/Manual Primary Vent 1-VG-TV-109A 1-VG-TV-109B 048 2" 1 Yes Yes No No SI O O C FC/FC Out Cold Gas Header Globe/Air Pilot Globe/Air Pilot Accumulator 1-SI-TV-101A 1-SI-TV-101B Vent Header to 050 1" 1 Yes Yes No No SI O O/INT C FC/FC Out Cold Gas SPS UFSAR Globe/Air Pilot Globe/Air Pilot Gaseous Waste Recirc Spray 1-SW-206 1-SW-208 051 2" 5 Yes Yes No No - LC INT LC NA Out Cold Liquid 4 HX SW Drains Gate/Manual Gate/Manual 1-SI-234 1-SI-TV-100 Nitrogen to PRT 053 1" 2 Yes Yes No No SI INT C C NA/FC In Cold Gas Check Globe/Air Pilot Primary Vent 1-VA-6 1-VA-1 054 2" 5 Yes Yes No No - LC O/INT LC NA Out Cold Gas 4 Pot Vent Gate/Manual Gate/Manual 1-LM-TV-100F

- Yes No No SI C C C -/FC Out Cold Gas 4 Leakage Globe/Air Pilot 055A 3/8 1 None Monitoring 1-LM-TV-100E

- Yes Globe/Air Pilot

a. See Note 9

b. Cold < 250°F; Hot > 250°F

c. O = Open; C = Closed; Int = Intermediate; LC = Locked Closed

d. FC = Fails Closed; FAI = Fails As Is; NA = Not Applicable 5.2-12

e. I/O = Inside/Outside

f. HH = Phase 3, H = Phase 2, SI = Phase 1

Table 5.2-1 (CONTINUED)

Revision 52Updated Online 09/30/20 UNIT 1 CONTAINMENT PENETRATIONS 10 CFR 50 Essential Line Penetration Configuration Appendix J System Nominal Class Valve Number/Type/Operator Test Required Level Line Statusc, d System Pen Line Per 1, 2, ESF Actuation Shut- Loss of Name No. Size UFSAR Inside Outside Inside Outside or No System Signalf Normal down Incident Pwr (I/O)e Flow Temperatureb Fluid Notes Pressurizer 1-SS-TV-100A 1-SS-TV-100B 056A 3/8 1 Yes Yes No No SI INT INT C -/FC Out Cold Liquid Liquid Sample Globe/Air Pilot Globe/Air Pilot RCS Cold Leg 1-SS-TV-102A 1-SS-TV-102B 056B 3/8 1 Yes Yes No No SI INT INT C FC/FC Out Cold Liquid Sample Globe/Air Pilot Globe/Air Pilot RCS Hot Leg 1-SS-TV-106A 1-SS-TV-106B 056D 3/8 1 Yes Yes No No SI INT INT C FC/FC Out Cold Liquid Sample Globe/Air Pilot Globe/Air Pilot 1-LM-TV-100H

- Yes No No SI INT INT C -/FC Out Cold Gas 4 Leakage Globe/Air Pilot 057A 3/8 1 None Monitoring 1-LM-TV-100G

- Yes -/FC Globe/Air Pilot 1-SS-TV-104A 1-SS-TV-104B PRT Sample 057B 3/8 1 Yes Yes No No SI INT INT C FC/FC Out Cold Gas Globe/Air Pilot Globe/Air Pilot Pressurizer 1-SS-TV-101A 1-SS-TV-101B Steam Space 057C 3/8 1 Yes Yes No No SI INT INT C FC/FC Out Cold Gas Globe/Air Pilot Globe/Air Pilot Sample 1-DA-TV-103A

- Yes No No SI INT INT C -/FC In Cold Liquid Post Accident Globe/Air Pilot 057D 2" 1 None Sample Return 1-DA-TV-103B

- Yes -/FC Globe/Air Pilot Instrument Air 1-IA-938 2-IA-446 SPS UFSAR 058 2" 5 Yes Yes No No - LC O LC NA In Cold Gas 4 from Unit 2 Check Gate/Manual A Low Head SI Pump 1-SI-229 1-SI-MOV-1890A 060 6" 4 No Noa 1 Yes - C C C/INT NA/FAI In Cold Liquid 2, 9 Discharge to Check Gate/Motor Hot Legs 3" 5 1-SI-500/Globe Valve/Manual No Noa 1 Yes - C C C NA/FAI In Cold Liquid 2, 4, 9

a. See Note 9

b. Cold < 250°F; Hot > 250°F

c. O = Open; C = Closed; Int = Intermediate; LC = Locked Closed

d. FC = Fails Closed; FAI = Fails As Is; NA = Not Applicable

e. I/O = Inside/Outside

f. HH = Phase 3, H = Phase 2, SI = Phase 1 5.2-13

Table 5.2-1 (CONTINUED)

Revision 52Updated Online 09/30/20 UNIT 1 CONTAINMENT PENETRATIONS 10 CFR 50 Essential Line Penetration Configuration Appendix J System Nominal Class Valve Number/Type/Operator Test Required Level Line Statusc, d System Pen Line Per 1, 2, ESF Actuation Shut- Loss of Name No. Size UFSAR Inside Outside Inside Outside or No System Signalf Normal down Incident Pwr (I/O)e Flow Temperatureb Fluid Notes 1-SI-243 No Noa 1 Yes - O - O NA/FAI In Cold Liquid 2, 6, 9 Low Head SI Check Pumps 1-SI-242 1-SI-MOV-1890C 061 6 4 No - NA/-

Discharge to Check Gate/Motor Cold Legs 1-SI-241 No - NA/-

Check B Low Head SI Pump 1-SI-228 1-SI-MOV-1890B 062 6" 4 No Noa 1 Yes - C C C/INT NA/FAI In Cold Liquid 2, 9 Discharge to Check Gate/Motor Hot Legs 1-CS-MOV-101C B CS Pump Yes Yes 1 Yes HH C C O NA/FAI In Cold Liquid 1-CS-24 Butterfly/Motor Discharge to 063 8" 4 Check 1-CS-MOV-101D Spray Ring - Yes NA/FAI Butterfly/Motor 1-CS-MOV-101A A CS Pump Yes Yes 1 Yes HH C C O NA/FAI In Cold Liquid 1-CS-13 Butterfly/Motor Discharge to 064 8" 4 Check 1-CS-MOV-101B Spray Ring - Yes NA/FAI Butterfly/Motor A ORS Pump 1-RS-MOV-155A

- Noa 1 Yes HH O O O -/FAI Out Cold Liquid 4, 9 Suction from Plug/Motor SPS UFSAR 066 12" 4 None Containment 1-RS-52

- Noa LC LC LC NA Sump Gate/Manual B ORS Pump 1-RS-MOV-155B

- Noa 1 Yes HH O O O -/FAI Out Cold Liquid 4, 9 Suction from Plug/Motor 067 12" 4 None Containment 1-RS-46 Sump - Noa LC LC LC NA Gate/Manual B Low Head 1-SI-MOV-1860B

- Noa 1 Yes - C C INT -/FAI Out Cold Liquid 4, 9 SI Suction from Gate/Motor 068 12" 4 None Containment 1-SI-311

- Noa LC LC LC NA Sump Gate/Manual

a. See Note 9

b. Cold < 250°F; Hot > 250°F

c. O = Open; C = Closed; Int = Intermediate; LC = Locked Closed

d. FC = Fails Closed; FAI = Fails As Is; NA = Not Applicable

e. I/O = Inside/Outside 5.2-14

f. HH = Phase 3, H = Phase 2, SI = Phase 1

Table 5.2-1 (CONTINUED)

Revision 52Updated Online 09/30/20 UNIT 1 CONTAINMENT PENETRATIONS 10 CFR 50 Essential Line Penetration Configuration Appendix J System Nominal Class Valve Number/Type/Operator Test Required Level Line Statusc, d System Pen Line Per 1, 2, ESF Actuation Shut- Loss of Name No. Size UFSAR Inside Outside Inside Outside or No System Signalf Normal down Incident Pwr (I/O)e Flow Temperatureb Fluid Notes A Low Head 1-SI-MOV-1860A

- Noa 1 Yes - C C INT -/FAI Out Cold Liquid 4, 9 SI Suction from Gate/Motor 069 12" 4 None Containment 1-SI-301

- Noa LC LC LC NA Sump Gate/Manual B ORS Pump 1-RS-11 1-RS-MOV-156B Discharge to 070 10" 4 Yes Yes 1 Yes HH O O O NA/FAI In Cold Liquid Check Butterfly/Motor Spray Ring A ORS Pump 1-RS-17 1-RS-MOV-156A Discharge 071 10" 4 Yes Yes 1 Yes HH O O O NA/FAI In Cold Liquid Check Butterfly/Motor to Spray Ring A Main Steam 1-MS-TV-101A 073 30" 3 None - No No No SI O C C -/FC Out Hot Steam 7 Header Check/Air Pilot or 1-MS-84

- No HH C C C -/NA Gate/Manual 1-MS-87 4" - None - No O O O -/NA 6 Gate/Manual 1-MS-379

- No C C C -/NA Gate/Manual 1-MS-NRV-102A 3" - None - No C C C -/NA Stop Check/Manual SPS UFSAR 1-GN-1 2" - None - No C C C -/NA Gate Manual 1-MS-266 1-1/2 - None - No C C C -/NA Gate Manual 1-MS-SV-101A 4" - None - No C C C -/NA Safety 1-MS-SV-102A 6" - None - No C C C -/NA Safety 1-MS-SV-103A 6" - None - No C C C -/NA Safety

a. See Note 9

b. Cold < 250°F; Hot > 250°F

c. O = Open; C = Closed; Int = Intermediate; LC = Locked Closed

d. FC = Fails Closed; FAI = Fails As Is; NA = Not Applicable

e. I/O = Inside/Outside 5.2-15

f. HH = Phase 3, H = Phase 2, SI = Phase 1

Table 5.2-1 (CONTINUED)

Revision 52Updated Online 09/30/20 UNIT 1 CONTAINMENT PENETRATIONS 10 CFR 50 Essential Line Penetration Configuration Appendix J System Nominal Class Valve Number/Type/Operator Test Required Level Line Statusc, d System Pen Line Per 1, 2, ESF Actuation Shut- Loss of Name No. Size UFSAR Inside Outside Inside Outside or No System Signalf Normal down Incident Pwr (I/O)e Flow Temperatureb Fluid Notes 1-MS-SV-104A 6" - None - No C C C -/NA Safety 1-MS-SV-105A 6" - None - No C C C -/NA Safety 1-MS-RV-101A 5" - None - No C C C -/FC Relief/Air B Main Steam 1-MS-TV-101B 074 30" 3 None - No No No SI O C C -/FC Out Hot Steam 7 Header Check/Air Pilot or 1-MS-116

- No HH C C C -/NA Gate/Manual 1-MS-120 4" - None - No O O O -/NA 6 Gate/Manual 1-MS-378

- No C C C -/NA Gate/Manual 1-MS-NRV-102B 3" - None - No C C C -/NA Stop Check/Manual 1-GN-2 2" - None - No C C C -/NA Gate/Manual 1-MS-268 1-1/2 - None - No C C C -/NA Gate Manual SPS UFSAR 1-MS-SV-101B 4" - None - No C C C -/NA Safety 1-MS-SV-102B 6" - None - No C C C -/NA Safety 1-MS-SV-103B 6" - None - No C C C -/NA Safety 1-MS-SV-104B 6" - None - No C C C -/NA Safety 1-MS-SV-105B 6" - None - No C C C -/NA Safety

a. See Note 9

b. Cold < 250°F; Hot > 250°F

c. O = Open; C = Closed; Int = Intermediate; LC = Locked Closed

d. FC = Fails Closed; FAI = Fails As Is; NA = Not Applicable

e. I/O = Inside/Outside 5.2-16

f. HH = Phase 3, H = Phase 2, SI = Phase 1

Table 5.2-1 (CONTINUED)

Revision 52Updated Online 09/30/20 UNIT 1 CONTAINMENT PENETRATIONS 10 CFR 50 Essential Line Penetration Configuration Appendix J System Nominal Class Valve Number/Type/Operator Test Required Level Line Statusc, d System Pen Line Per 1, 2, ESF Actuation Shut- Loss of Name No. Size UFSAR Inside Outside Inside Outside or No System Signalf Normal down Incident Pwr (I/O)e Flow Temperatureb Fluid Notes 1-MS-RV-101B 5" - None - No C C C -/FC Relief/Air C Main Steam 1-MS-TV-101C 075 30" 3 None - No No No SI O C C -/FC Out Hot Steam 7 Header Check/Air Pilot or 1-MS-155

- No HH C C C -/NA Gate/Manual 1-MS-158 4" - None - No O O O -/NA 6 Gate/Manual 1-MS-377

- No C C C -/NA Gate/Manual 1-MS-NRV-102C 3" - None - No C C C -/NA Stop Check/Manual 1-GN-3 2" - None - No C C C -/NA Gate/Manual 1-MS-208 1-1/2 - None - No C C C -/NA Gate Manual 1-MS-SV-101C 4" - None - No C C C -/NA Safety 1-MS-SV-102C 6" - None - No C C C -/NA Safety SPS UFSAR 1-MS-SV-103C 6" - None - No C C C -/NA Safety 1-MS-SV-104C 6" - None - No C C C -/NA Safety 1-MS-SV-105C 6" - None - No C C C -/NA Safety 1-MS-RV-101C 5" - None - No C C C -/FC Relief/Air A Feedwater 1-FW-10 1-FW-12 076 14" 3 No No No No - O C O NA/NA In Hot Liquid 7 Header Check Check

a. See Note 9

b. Cold < 250°F; Hot > 250°F

c. O = Open; C = Closed; Int = Intermediate; LC = Locked Closed

d. FC = Fails Closed; FAI = Fails As Is; NA = Not Applicable

e. I/O = Inside/Outside 5.2-17

f. HH = Phase 3, H = Phase 2, SI = Phase 1

Table 5.2-1 (CONTINUED)

Revision 52Updated Online 09/30/20 UNIT 1 CONTAINMENT PENETRATIONS 10 CFR 50 Essential Line Penetration Configuration Appendix J System Nominal Class Valve Number/Type/Operator Test Required Level Line Statusc, d System Pen Line Per 1, 2, ESF Actuation Shut- Loss of Name No. Size UFSAR Inside Outside Inside Outside or No System Signalf Normal down Incident Pwr (I/O)e Flow Temperatureb Fluid Notes C Feedwater 1-FW-72 1-FW-74 077 14" 3 No No No No - O C O NA/NA In Hot Liquid 7 Header Check Check B Feedwater 1-FW-41 1-FW-43 078 14" 3 No No No No - O C O NA/NA In Hot Liquid 7 Header Check Check SW to D RS 1-SW-MOV-104D 079 24" 3 None - No 1 Yes HH C C O -/FAI In Cold Liquid 8 HX Butterfly/Motor SW to C RS 1-SW-MOV-104C 080 24" 3 None - No 1 Yes HH C C O -/FAI In Cold Liquid 8 HX Butterfly/Motor SW to B RS 1-SW-MOV-104B 081 24" 3 None - No 1 Yes HH C C O -/FAI In Cold Liquid 8 HX Butterfly/Motor SW to A RS 1-SW-MOV-104A 082 24" 3 None - No 1 Yes HH C C O -/FAI In Cold Liquid 8 HX Butterfly/Motor SW from D 1-SW-MOV-105D 083 24" 3 None - No 1 Yes HH C C O -/FAI Out Cold Liquid 8 RS HX Butterfly/Motor SW from C RS 1-SW-MOV-105C 084 24" 3 None - No 1 Yes HH C C O -/FAI Out Cold Liquid 8 HX Butterfly/Motor SW from B RS 1-SW-MOV-105B 085 24" 3 None - No 1 Yes HH C C O -/FAI Out Cold Liquid 8 HX Butterfly/Motor SW from A RS 1-SW-MOV-105A 086 24" 3 None - No 1 Yes HH C C O -/FAI Out Cold Liquid 8 HX Butterfly/Motor SPS UFSAR 1-FW-131 1-FW-133 AFW to FW 087 6" 3 No No 2 Yes - C C O NA/NA In Cold Liquid 7 Check Check 1-FW-136 1-FW-138 AFW to FW 088 6" 3 No No 2 Yes - C C O NA/NA In Cold Liquid 7 Check Check Condenser Air 1-VP-12 1-SV-TV-102A Ejector 089 6" 1 Yes Yes No - SI O O C NA/FC In Cold Gas Check Gate/Air Pilot Discharge Containment Purge Exhaust 1-VS-MOV-100C 1-VS-MOV-101 090 36" 5 Yes Yes No - - LC O LC FAI/FAI Out Cold Gas 4, 10 Ventilation Butterfly/Motor Butterfly/Motor System

a. See Note 9

b. Cold < 250°F; Hot > 250°F

c. O = Open; C = Closed; Int = Intermediate; LC = Locked Closed

d. FC = Fails Closed; FAI = Fails As Is; NA = Not Applicable 5.2-18

e. I/O = Inside/Outside

f. HH = Phase 3, H = Phase 2, SI = Phase 1

Table 5.2-1 (CONTINUED)

Revision 52Updated Online 09/30/20 UNIT 1 CONTAINMENT PENETRATIONS 10 CFR 50 Essential Line Penetration Configuration Appendix J System Nominal Class Valve Number/Type/Operator Test Required Level Line Statusc, d System Pen Line Per 1, 2, ESF Actuation Shut- Loss of Name No. Size UFSAR Inside Outside Inside Outside or No System Signalf Normal down Incident Pwr (I/O)e Flow Temperatureb Fluid Notes 1-VS-MOV-100D

- Yes LC O LC -/FAI Butterfly/Motor Containment Purge Supply 1-VS-MOV-100A 1-VS-MOV-100B 091 36" 5 Yes Yes No - - LC O LC FAI/FAI In Cold Gas 4, 10 Ventilation Butterfly/Motor Butterfly/Motor System 1-VS-MOV-102

- Yes LC O LC -/FAI Butterfly/Motor 1-CV-TV-150C Containment 092 2" 1 None - Yes 1 No SI INT C C -/FC Out Cold Gas Globe/Air Pilot Vacuum A 1-CV-TV-150D Pump Suction 3/8 1 None - Yes SI INT C C -/FC Globe/Air Pilot and Gaseous Waste 1-GW-TV-100

- Yes No - C C O -/FC Plug/Air Pilot Hydrogen Analyzer 1-GW-TV-101

- Yes - C C O -/FC Plug/Air Pilot 1-CV-TV-150A Containment 093 2" 1 None - Yes 1 No SI INT C C -/FC Out Cold Gas Globe/Air Pilot Vacuum B 1-CV-TV-150B Pump Suction 3/8 1 None - Yes SI INT C C -/FC Globe/Air Pilot SPS UFSAR and 1-GW-TV-104 Gaseous Waste - Yes No - C C O -/FC Hydrogen Plug/Air Pilot Analyzer 1-GW-TV-105

- Yes - C C O -/FC Plug/Air Pilot Containment Vacuum & 1-CV-HCV-100 1-CV-2 094 8" 5 Yes Yes No No - LC INT LC FC/NA Out Cold Gas 4 Leakage Globe/Air Pilot Gate/Manual Monitoring A SG Recirc & 1-RT-2 1-RT-6 096 3" 3 Yes Yes No No - LC O LC NA Out Cold Liquid 4, 11 Transfer Ball/Manual Ball/Manual

a. See Note 9

b. Cold < 250°F; Hot > 250°F

c. O = Open; C = Closed; Int = Intermediate; LC = Locked Closed

d. FC = Fails Closed; FAI = Fails As Is; NA = Not Applicable

e. I/O = Inside/Outside 5.2-19

f. HH = Phase 3, H = Phase 2, SI = Phase 1

Table 5.2-1 (CONTINUED)

Revision 52Updated Online 09/30/20 UNIT 1 CONTAINMENT PENETRATIONS 10 CFR 50 Essential Line Penetration Configuration Appendix J System Nominal Class Valve Number/Type/Operator Test Required Level Line Statusc, d System Pen Line Per 1, 2, ESF Actuation Shut- Loss of Name No. Size UFSAR Inside Outside Inside Outside or No System Signalf Normal down Incident Pwr (I/O)e Flow Temperatureb Fluid Notes B SG Recirc & 1-RT-21 1-RT-25 022 3" 3 Yes Yes No No - LC O LC NA Out Cold Liquid 4, 11 Transfer Ball/Manual Ball/Manual C SG Recirc & 1-RT-40 1-RT-44 114 3" 3 Yes Yes No No - LC O LC NA Out Cold Liquid 4, 11 Transfer Ball/Manual Ball/Manual 1-SS-TV-103A 1-SS-TV-103B RHR Sample 097B 3/8 1 Yes Yes 1 No SI C INT C FC/FC Out Cold Liquid Globe/Air Pilot Globe/Air Pilot Leakage 1-LM-TV-100A 097C 3/8 1 None - Yes 1 No SI C C C -/FC Out Cold Gas 4 Monitoring Globe/Air Pilot 1-LM-TV-100B

- Yes /FC Globe/Air Pilot 3/8 valve, 1-GW-TV-102 Gaseous Waste 100 1 None - Yes 1 No - C C C -/FC In Cold Gas 3" penetration Globe/Air Pilot 1-GW-TV-103

- Yes /FC Globe/Air Pilot 1-FP-151 Fire Protection 101 4" 5 None - Yes No No - LC INT LC NA In Cold Liquid 4 Ball/Manual 1-FP-152

- Yes NA Ball/Manual Unit 2 - AFW 1-FW-273 1-FW-272 102 6" 3 No No 2 Yes - C C O/INT NA In Cold Liquid 7 Cross Connect Check Check SPS UFSAR Reactor Cavity 1-RL-5 1-RL-3 Purification 103 3" 5 Yes Yes No No - LC O/INT LC NA In Cold Liquid 4, 10 Diaphragm/Manual Diaphragm/Manual Inlet Reactor Cavity 1-RL-13 1-RL-15 Purification 104 3" 5 Yes Yes No No - LC O/INT LC NA Out Cold Liquid 4 Diaphragm/Manual Diaphragm/Manual Outlet Hydrogen 1-GW-TV-111A 1-GW-TV-111B 105C 3/8 1 Yes Yes 1 No - C C O/INT FC/FC Out Cold Gas Sample Globe/Air Pilot Globe/Air Pilot Leakage 1-LM-TV-100C 105D 3/8 1 None - Yes 1 No SI C C C -/FC Out Cold Gas 4 Monitoring Globe/Air Pilot

a. See Note 9

b. Cold < 250°F; Hot > 250°F

c. O = Open; C = Closed; Int = Intermediate; LC = Locked Closed

d. FC = Fails Closed; FAI = Fails As Is; NA = Not Applicable

e. I/O = Inside/Outside 5.2-20

f. HH = Phase 3, H = Phase 2, SI = Phase 1

Table 5.2-1 (CONTINUED)

Revision 52Updated Online 09/30/20 UNIT 1 CONTAINMENT PENETRATIONS 10 CFR 50 Essential Line Penetration Configuration Appendix J System Nominal Class Valve Number/Type/Operator Test Required Level Line Statusc, d System Pen Line Per 1, 2, ESF Actuation Shut- Loss of Name No. Size UFSAR Inside Outside Inside Outside or No System Signalf Normal down Incident Pwr (I/O)e Flow Temperatureb Fluid Notes 1-LM-TV-100D

- Yes -/FC Globe/Air Pilot 1-SI-HCV-1850A 1-SI-73 106 3/4 5 No Yes No No - C/LC LC LC FC/NA Out Cold Liquid 3, 4 Globe/Air Pilot Globe/Manual 1-SI-HCV-1850B No - FC/-

Globe/Air Pilot 1-SI-HCV-1850C No - FC/-

Globe/Air Pilot SI Test Line 1-SI-HCV-1850D No - FC/-

Globe/Air Pilot 1-SI-HCV-1850E No - FC/-

Globe/Air Pilot 1-SI-HCV-1850F No - FC/-

Globe/Air Pilot CCW from RCP 1-CC-TV-140A 1-CC-TV-140B 110 3" 3 No No 2 No HH O O C FC/FC Out Cold Liquid 1 Thermal Barrier Globe/Air Pilot Globe/Air Pilot Containment 1-IA-TV-101A 1-IA-TV-101B 112 3" 1 Yes Yes 1 No H O O C FC/FC Out Cold Gas Instrument Air Plug/Air Pilot Plug/Air Pilot 1-SI-227 1-SI-MOV-1869A 113 3" 4 No Noa 1 Yes - C C O/INT NA/FAI In Cold Liquid 2, 4, 9 Check Gate/Motor SPS UFSAR Safety Injection 1-SI-174

- Noa LC LC LC -/NA Globe/Manual

a. See Note 9

b. Cold < 250°F; Hot > 250°F

c. O = Open; C = Closed; Int = Intermediate; LC = Locked Closed

d. FC = Fails Closed; FAI = Fails As Is; NA = Not Applicable

e. I/O = Inside/Outside

f. HH = Phase 3, H = Phase 2, SI = Phase 1 5.2-21

Revision 52Updated Online 09/30/20 Table 5.2-2 UNIT 2 CONTAINMENT PENETRATIONS 10 CFR 50 Essential Line Penetration Configuration Appendix J System Nominal Class Valve Number/Type/Operator Test Required Level Line Status c,d System Pen Line Per 1, 2, ESF Actuation Shut- Loss of Name No. Size UFSAR Inside Outside Inside Outside or No System Signalf Normal down Incident Pwr (I/O)e Flow Temperatureb Fluid Notes CCW from B 2-CC-TV-209B 001 18" 3 None - No 2 No SI - - C -/FC Out Cold Liquid 1 RHR HX Butterfly/Air Pilot CCW to A 2-CC-177 2-CC-214 002 18" 3 No No 2 No - O O C NA In Cold Liquid 1 RHR HX Check Butterfly/Manual CCW to B 2-CC-176 2-CC-220 004 18" 3 No No 2 No - O O C NA In Cold Liquid 1 RHR HX Check Butterfly/Manual CCW from A 2- CC-TV-209A 005 18" 3 None - No 2 No SI - - C NA/FC Out Cold Liquid 1 RHR HX Butterfly/Air Pilot 2-SI-225 2-SI-150 007 3" 4 No Noa 1 Yes SI LC LC LC NA/LC In Cold Liquid 2, 4, 9 Check Globe/Manual High Head SI 2-SI-MOV-2867D

- Noa C C O -/FAI (Normal header) Gate/Motor 2-SI-MOV-2867C

- Noa C C O -/FAI 1 Gate/Motor CCW to C Air 2-CC-224 2-CC-223 009 6" 3 No No 2 No - O O C NA In Cold Liquid 1 Recirc. Fan Check Gate/Manual CCW to A Air 2-CC-242 2-CC-241 010 6" 3 No No 2 No - O O C NA In Cold Liquid 1 Recirc. Fan Check Gate/Manual CCW to B Air 2-CC-233 2-CC-232 011 6" 3 No No 2 No - O O C NA In Cold Liquid 1 Recirc. Fan Check Gate/Manual SPS UFSAR CCW from B 2-CC-TV-210B Air 012 6" 3 None - No 2 No HH O O C -/FC Out Cold Liquid 1 Plug/Air Pilot Recirc. Fan CCW from C 2-CC-TV-210C Air 013 6" 3 None - No 2 No HH O O C -/FC Out Cold Liquid 1 Plug/Air Pilot Recirc. Fan CCW from A 2-CC-TV-210A Air 014 6" 3 None - No 2 No HH O O C -/FC Out Cold Liquid 1 Plug/Air Pilot Recirc Fan Chemical and 2-CH-309 2-CH-MOV-2289A 015 3" 2 No Noa No No SI O O C NA/FAI In Cold Liquid 9 Volume Control Check Gate/Motor

a. See Note 9

b. Cold < 250°F; Hot > 250°F

c. O = Open; C = Closed; Int = Intermediate; LC = Locked Closed

d. FC = Fails Closed; FAI = Fails As Is; NA = Not Applicable

e. I/O = Inside/Outside 5.2-22

f. HH = Phase 3, H = Phase 2, SI = Phase 1

Table 5.2-2 (CONTINUED)

Revision 52Updated Online 09/30/20 UNIT 2 CONTAINMENT PENETRATIONS 10 CFR 50 Essential Line Penetration Configuration Appendix J System Nominal Class Valve Number/Type/Operator Test Required Level Line Status c,d System Pen Line Per 1, 2, ESF Actuation Shut- Loss of Name No. Size UFSAR Inside Outside Inside Outside or No System Signalf Normal down Incident Pwr (I/O)e Flow Temperatureb Fluid Notes CCW to C 2-CC-59 2-CC-216 016 6" 3 No No 2 No - O O C NA In Cold Liquid 1 RCP Check Gate/Manual CCW to B 2-CC-58 2-CC-218 017 6" 3 No No 2 No - O O C NA In Cold Liquid 1 RCP Check Gate/Manual CCW to A 2-CC-1 2-CC-219 018 6" 3 No No 2 No - O O C NA In Cold Liquid 1 RCP Check Gate/Manual Seal Water from 2-CH-MOV-2381 019 3" 2 None - Yes No No SI O O C -/FAI Out Cold Liquid RCPs Gate/Motor 2-SI-HCV-2851A 2-SI-32 020 1" 5 No Yes No - - LC LC LC FC/NA In Cold Liquid 3, 4 Globe/Air Pilot Globe/Manual Safety Injection 2-SI-HCV-2851B Accumulator No - FC/-

Globe/Air Pilot Makeup 2-SI-HCV-2851C No - FC/-

Globe/Air Pilot High Head Safety Injection 2-SI-224 2-SI-MOV-2842 to Cold Leg 021 3" 4 No Noa 1 Yes - C C INT NA/FAI In Cold Liquid 2, 9 Check Gate/Motor (Alternate header)

High Head 2-SI-226 2-SI-MOV-2869B Noa SPS UFSAR Safety Injection 023 3" 4 No 1 Yes - C C INT NA/FAI In Cold Liquid 2, 9 Check Gate/Motor to Hot Leg 2-RH-29 2-RH-108 RHR to RWST 024 6" 5 No Yes No No - LC INT LC NA/FAI Out Cold Liquid 3, 4 Gate/Manual Gate/Manual CCW from A 2-CC-TV-205A 025 6" 3 None - No 2 No HH O O C -/FC Out Cold Liquid 1 RCP Plug/Air Pilot CCW from C 2-CC-TV-205C 026 6" 3 None - No 2 No HH O O C -/FC Out Cold Liquid 1 RCP Plug/Air Pilot CCW from B 2-CC-TV-205B 027 6" 3 None - No 2 No HH O O C -/FC Out Cold Liquid 1 RCP Plug/Air Pilot Reactor Coolant 2-CH-TV-2204A 2-CH-TV-2204B 028 2" 2 Yes Yes No No SI O O C FC/FC Out Hot Liquid Letdown Globe/Air Pilot Globe/Air Pilot

a. See Note 9

b. Cold < 250°F; Hot > 250°F

c. O = Open; C = Closed; Int = Intermediate; LC = Locked Closed

d. FC = Fails Closed; FAI = Fails As Is; NA = Not Applicable 5.2-23

e. I/O = Inside/Outside

f. HH = Phase 3, H = Phase 2, SI = Phase 1

Table 5.2-2 (CONTINUED)

Revision 52Updated Online 09/30/20 UNIT 2 CONTAINMENT PENETRATIONS 10 CFR 50 Essential Line Penetration Configuration Appendix J System Nominal Class Valve Number/Type/Operator Test Required Level Line Status c,d System Pen Line Per 1, 2, ESF Actuation Shut- Loss of Name No. Size UFSAR Inside Outside Inside Outside or No System Signalf Normal down Incident Pwr (I/O)e Flow Temperatureb Fluid Notes 3/8 valve, 3" 2-GW-TV-202 Gaseous Waste 032 1 None - Yes No No - C C C -/FC In Cold Gas penetration Globe/Air Pilot 2-GW-TV-203

- Yes Globe/Air Pilot Primary Drain 2-DG-TV-208A 2-DG-TV-208B Transfer Tank 033 2" 1 Yes Yes No No SI O O C FC/FC Out Cold Liquid Globe/Air Pilot Globe/Air Pilot Pump Discharge Seal Water to 2-CH-349 2-CH-300 035 2" 4 No No 1 No - O O O NA In Cold Liquid 2, 5 C RCP Check Needle/Manual Seal Water to 2-CH-323 2-CH-294 036 2" 4 No No 1 No - O O O NA In Cold Liquid 2, 5 A RCP Check Needle/Manual Seal Water to 2-CH-333 2-CH-297 037 2" 4 No No 1 No - O O O NA In Cold Liquid 2, 5 B RCP Check Needle/Manual Aerated Drain 2-DA-TV-200A 2-DA-TV-200B Sump Pump 038 2" 1 Yes Yes No No SI O O C FC/FC Out Cold Liquid Ball/Air Pilot Ball/Air Pilot Discharge SG 2-BD-TV-200A 2-BD-TV-200B 039 3" 3 No No No No AFW O O C FC/FC Out Hot Liquid 7 Blowdown-1A Gate/Air Pilot Gate/Air Pilot SG 2-BD-TV-200E 2-BD-TV-200F 040 3" 3 No No No No AFW O O C FC/FC Out Hot Liquid 7 Blowdown-1C Gate/Air Pilot Gate/Air Pilot SPS UFSAR SG 2-BD-TV-200C 2-BD-TV-200D 041 3" 3 No No No No AFW O O C FC/FC Out Hot Liquid 7 Blowdown-1B Gate/Air Pilot Gate/Air Pilot Service Air 2-SA-82 042 2" 5 None - Yes No No - LC INT LC NA In Cold Gas 4 Supply Gate/Manual 2-SA-81

- Yes Gate Manual Particulate and Gaseous Rad 2-RM-3 2-RM-TV-200A 043 1" 1 Yes Yes No No H O O C NA/FC In Cold Gas Monitoring Check Globe/Air Pilot Return

a. See Note 9

b. Cold < 250°F; Hot > 250°F

c. O = Open; C = Closed; Int = Intermediate; LC = Locked Closed

d. FC = Fails Closed; FAI = Fails As Is; NA = Not Applicable

e. I/O = Inside/Outside

f. HH = Phase 3, H = Phase 2, SI = Phase 1 5.2-24

Table 5.2-2 (CONTINUED)

Revision 52Updated Online 09/30/20 UNIT 2 CONTAINMENT PENETRATIONS 10 CFR 50 Essential Line Penetration Configuration Appendix J System Nominal Class Valve Number/Type/Operator Test Required Level Line Status c,d System Pen Line Per 1, 2, ESF Actuation Shut- Loss of Name No. Size UFSAR Inside Outside Inside Outside or No System Signalf Normal down Incident Pwr (I/O)e Flow Temperatureb Fluid Notes Particulate and Gaseous 2-RM-TV-200C 2-RM-TV-200B 044 1" 1 Yes Yes No No H O O C FC/FC Out Cold Gas Rad Monitoring Globe/Air Pilot Globe/Air Pilot Supply 2-RC-TV-2519A Primary Grade 2-RC-160 045 3" 1 Diaphragm/Air Yes Yes No No SI INT C C NA/FC In Cold Liquid Water to PRT Check Pilot Loop Fill 2-RC-HCV-2556A 2-CH-FCV-2160 046 2" 5 No Noa No No - C INT C FC/FC In Cold Liquid 3, 9 Header Globe/Air Pilot Globe/Air Pilot 2-RC-HCV-2556B No - FC/

Globe/Air Pilot 2-RC-HCV-2556C No - FC/

Globe/Air Pilot Instrument Air 2-IA-864 2-IA-TV-200 047 2" 1 Yes Yes No No HH O O C NA/FC In Cold Gas 4 Supply Check Globe/Air Pilot 2-IA-704

- Yes INT/LC O INT/LC -/NA Gate/Manual Primary Vent 2-VG-TV-209A 2-VG-TV-209B 048 2" 1 Yes Yes No No SI O O C FC/FC Out Cold Gas Header Globe/Air Pilot Globe/Air Pilot Accumulator 2-SI-TV-201A 2-SI-TV-201B SPS UFSAR Vent Header to 050 1" 1 Yes Yes No No SI O O/INT C FC/FC Out Cold Gas Globe/Air Pilot Globe/Air Pilot Gaseous Waste Recirc Spray 2-SW-206 2-SW-208 051 2" 5 Yes Yes No No - LC INT LC NA Out Cold Liquid 4 HX SW Drains Gate/Manual Gate/Manual 2-SI-304 2-SI-TV-200 Nitrogen to PRT 053 1" 2 Yes Yes No No SI INT C C NA/FC In Cold Gas Check Globe/Air Pilot Primary Vent 2-VA-9 2-VA-1 054 2" 5 Yes Yes No No - LC O/INT LC NA Out Cold Gas 4 Pot Vent Gate/Manual Gate/Manual Leakage 2-LM-TV-200G 055D 3/8 1 None - Yes No No SI C C C -/FC Out Cold Gas 4 Monitoring Globe/Air Pilot 2-LM-TV-200H

- Yes Globe/Air Pilot

a. See Note 9

b. Cold < 250°F; Hot > 250°F

c. O = Open; C = Closed; Int = Intermediate; LC = Locked Closed

d. FC = Fails Closed; FAI = Fails As Is; NA = Not Applicable 5.2-25

e. I/O = Inside/Outside

f. HH = Phase 3, H = Phase 2, SI = Phase 1

Table 5.2-2 (CONTINUED)

Revision 52Updated Online 09/30/20 UNIT 2 CONTAINMENT PENETRATIONS 10 CFR 50 Essential Line Penetration Configuration Appendix J System Nominal Class Valve Number/Type/Operator Test Required Level Line Status c,d System Pen Line Per 1, 2, ESF Actuation Shut- Loss of Name No. Size UFSAR Inside Outside Inside Outside or No System Signalf Normal down Incident Pwr (I/O)e Flow Temperatureb Fluid Notes RCS Hot Leg 2-SS-TV-206A 2-SS-TV-206B 056A 3/8 1 Yes Yes No No SI INT INT C -/FC Out Cold Liquid Sample Globe/Air Pilot Globe/Air Pilot RCS Cold Leg 2-SS-TV-202A 2-SS-TV-202B 056B 3/8 1 Yes Yes No No SI INT INT C FC/FC Out Cold Liquid Sample Globe/Air Pilot Globe/Air Pilot Pressurizer 2-SS-TV-200A 2-SS-TV-200B 056D 3/8 1 Yes Yes No No SI INT INT C FC/FC Out Cold Liquid Liquid Sample Globe/Air Pilot Globe/Air Pilot Pressurizer 2-SS-TV-201A 2-SS-TV-201B Steam Space 057A 3/8 1 Yes Yes No No SI INT INT C FC/FC Out Cold Gas Globe/Air Pilot Globe/Air Pilot Sample Post Accident 2-DA-TV-203B 057B 2" 1 None - Yes No No SI INT INT C -/FC In Cold Liquid Sample Return Globe/Air Pilot 2-DA-TV-203A

- Yes -/FC Globe/Air Pilot Leakage 2-LM-TV-200F 057C 3/8 1 None - Yes No No SI INT INT C -/FC Out Cold Gas 4 Monitoring Globe/Air Pilot 2-LM-TV-200E

- Yes -/FC Globe/Air Pilot 2-SS-TV-204A 2-SS-TV-204B PRT Sample 057D 3/8 1 Yes Yes No No SI INT INT C FC/FC Out Cold Gas Globe/Air Pilot Globe/Air Pilot Instrument Air 2-IA-868 1-IA-704 058 2" 5 Yes Yes No No - LC O LC NA In Cold Gas 4 SPS UFSAR from Unit 1 Check Gate/Manual A Low Head SI Pump 2-SI-229 2-SI-MOV-2890A 060 6" 4 No Noa 1 Yes - C C C/INT NA/FAI In Cold Liquid 2, 9 Discharge to Check Gate/Motor Hot Legs 3" 5 2-SI-500/Globe Valve/Manual No Noa 1 Yes - C C C NA/FAI In Cold Liquid 2, 4, 9 Low Head SI Pumps 2-SI-243 2-SI-MOV-2890C 061 6" 4 No Noa 1 Yes - O - O NA/FAI In Cold Liquid 2, 6, 9 Discharge to Check Gate/Motor Cold Legs 2-SI-242 No - NA/

Check

a. See Note 9

b. Cold < 250°F; Hot > 250°F

c. O = Open; C = Closed; Int = Intermediate; LC = Locked Closed

d. FC = Fails Closed; FAI = Fails As Is; NA = Not Applicable 5.2-26

e. I/O = Inside/Outside

f. HH = Phase 3, H = Phase 2, SI = Phase 1

Table 5.2-2 (CONTINUED)

Revision 52Updated Online 09/30/20 UNIT 2 CONTAINMENT PENETRATIONS 10 CFR 50 Essential Line Penetration Configuration Appendix J System Nominal Class Valve Number/Type/Operator Test Required Level Line Status c,d System Pen Line Per 1, 2, ESF Actuation Shut- Loss of Name No. Size UFSAR Inside Outside Inside Outside or No System Signalf Normal down Incident Pwr (I/O)e Flow Temperatureb Fluid Notes 2-SI-241 No - NA/

Check B Low Head SI Pump 2-SI-228 2-SI-MOV-2890B 062 6" 4 No Noa 1 Yes - C C C/INT NA/FAI In Cold Liquid 2, 9 Discharge to Check Gate/Motor Hot Legs B CS Pump 2-CS-24 2-CS-MOV-201C Discharge 063 8" 4 Yes Yes 1 Yes HH C C O NA/FAI In Cold Liquid Check Butterfly/Motor to Spray Ring 2-CS-MOV-201D

- Yes NA/FAI Butterfly/Motor A CS Pump 2-CS-13 2-CS-MOV-201A Discharge 064 8" 4 Yes Yes 1 Yes HH C C O NA/FAI In Cold Liquid Check Butterfly/Motor to Spray Ring 2-CS-MOV-201B

- Yes NA/FAI Butterfly/Motor A ORS Pump Suction from 2-RS-MOV-255A 066 12" 4 None - Noa 1 Yes HH O O O -/FAI Out Cold Liquid 4, 9 Containment Plug/Motor Sump 2-RS-53 SPS UFSAR

- Noa LC LC LC NA Gate/Manual B ORS Pump Suction from 2-RS-MOV-255B 067 12" 4 None - Noa 1 Yes HH O O O -/FAI Out Cold Liquid 4, 9 Containment Plug/Motor Sump 2-RS-46

- Noa LC LC LC NA Gate/Manual

a. See Note 9

b. Cold < 250°F; Hot > 250°F

c. O = Open; C = Closed; Int = Intermediate; LC = Locked Closed

d. FC = Fails Closed; FAI = Fails As Is; NA = Not Applicable

e. I/O = Inside/Outside

f. HH = Phase 3, H = Phase 2, SI = Phase 1 5.2-27

Table 5.2-2 (CONTINUED)

Revision 52Updated Online 09/30/20 UNIT 2 CONTAINMENT PENETRATIONS 10 CFR 50 Essential Line Penetration Configuration Appendix J System Nominal Class Valve Number/Type/Operator Test Required Level Line Status c,d System Pen Line Per 1, 2, ESF Actuation Shut- Loss of Name No. Size UFSAR Inside Outside Inside Outside or No System Signalf Normal down Incident Pwr (I/O)e Flow Temperatureb Fluid Notes B Low Head SI Suction from 2-SI-MOV-2860B 068 12" 4 None - Noa 1 Yes - C C INT -/FAI Out Cold Liquid 4, 9 Containment Gate/Motor Sump 2-SI-321

- Noa LC LC LC NA Gate/Manual A Low Head SI Suction from 2-SI-MOV-2860A 069 12" 4 None - Noa 1 Yes - C C INT -/FAI Out Cold Liquid 4, 9 Containment Gate/Motor Sump 2-SI-322

- Noa LC LC LC NA Gate/Manual B ORS Pump 2-RS-11 2-RS-MOV-256B Discharge 070 10" 4 Yes Yes 1 Yes HH O O O NA/FAI In Cold Liquid Check Butterfly/Motor to Spray Ring A ORS Pump 2-RS-017 2-RS-MOV-256A Discharge 071 10" 4 Yes Yes 1 Yes HH O O O NA/FAI In Cold Liquid Check Butterfly/Motor to Spray Ring A Main Steam 2-MS-TV-201A 073 30" 3 None - No No No SI O C C -/FC Out Hot Steam 7 Header Check/Air Pilot or 2-MS-84 SPS UFSAR

- No HH C C C -/NA Gate/Manual 2-MS-87 4" - None - No O O O -/NA 6 Gate/Manual 2-MS-379

- No C C C -/NA Gate/Manual 2-MS-NRV-202A 3" - None Stop - No C C C -/NA Check/Manual 2-GN-1 2" - None - No C C C -/NA Gate Manual

a. See Note 9

b. Cold < 250°F; Hot > 250°F

c. O = Open; C = Closed; Int = Intermediate; LC = Locked Closed

d. FC = Fails Closed; FAI = Fails As Is; NA = Not Applicable

e. I/O = Inside/Outside 5.2-28

f. HH = Phase 3, H = Phase 2, SI = Phase 1

Table 5.2-2 (CONTINUED)

Revision 52Updated Online 09/30/20 UNIT 2 CONTAINMENT PENETRATIONS 10 CFR 50 Essential Line Penetration Configuration Appendix J System Nominal Class Valve Number/Type/Operator Test Required Level Line Status c,d System Pen Line Per 1, 2, ESF Actuation Shut- Loss of Name No. Size UFSAR Inside Outside Inside Outside or No System Signalf Normal down Incident Pwr (I/O)e Flow Temperatureb Fluid Notes 2-MS-266 1-1/2 - None - No C C C -/NA Gate Manual 2-MS-SV-201A 4" - None - No C C C -/NA Safety 2-MS-SV-202A 6" - None - No C C C -/NA Safety 2-MS-SV-203A 6" - None - No C C C -/NA Safety 2-MS-SV-204A 6" - None - No C C C -/NA Safety 2-MS-SV-205A 6" - None - No C C C -/NA Safety 2-MS-RV-201A 5" - None - No C C C -/FC Relief/Air C Main Steam 2-MS-TV-201C 074 30" 3 None - No No No SI O C C -/FC Out Hot Steam 7 Header Check/Air Pilot or 2-MS-155

- No HH C C C -/NA Gate/Manual 2-MS-158 4" - None - No O O O -/NA 6 Gate/Manual SPS UFSAR 2-MS-377

- No C C C -/NA Gate/Manual 2-MS-NRV-202C 3" - None Stop - No C C C -/NA Check/Manual 2-GN-3 2" - None - No C C C -/NA Gate/Manual 2-MS-208 1-1/2 - None - No C C C -/NA Gate Manual 2-MS-SV-201C 4" - None - No C C C -/NA Safety

a. See Note 9

b. Cold < 250°F; Hot > 250°F

c. O = Open; C = Closed; Int = Intermediate; LC = Locked Closed

d. FC = Fails Closed; FAI = Fails As Is; NA = Not Applicable

e. I/O = Inside/Outside 5.2-29

f. HH = Phase 3, H = Phase 2, SI = Phase 1

Table 5.2-2 (CONTINUED)

Revision 52Updated Online 09/30/20 UNIT 2 CONTAINMENT PENETRATIONS 10 CFR 50 Essential Line Penetration Configuration Appendix J System Nominal Class Valve Number/Type/Operator Test Required Level Line Status c,d System Pen Line Per 1, 2, ESF Actuation Shut- Loss of Name No. Size UFSAR Inside Outside Inside Outside or No System Signalf Normal down Incident Pwr (I/O)e Flow Temperatureb Fluid Notes 2-MS-SV-202C 6" - None - No C C C -/NA Safety 2-MS-SV-203C 6" - None - No C C C -/NA Safety 2-MS-SV-204C 6" - None - No C C C -/NA Safety 2-MS-SV-205C 6" - None - No C C C -/NA Safety 2-MS-RV-201C 5" - None - No C C C -/FC Relief/Air B Main Steam 2-MS-TV-201B 075 30" 3 None - No No No SI O C C -/FC Out Hot Steam 7 Header Check/Air Pilot or 2-MS-116

- No HH C C C -/NA Gate/Manual 2-MS-120 4" - None - No O O O -/NA 6 Gate/Manual 2-MS-378

- No C C C -/NA Gate/Manual 2-MS-NRV-202B 3" - None Stop - No C C C -/NA SPS UFSAR Check/Manual 2-GN-2 2" - None - No C C C -/NA Gate/Manual 2-MS-268 1-1/2 - None - No C C C -/NA Gate Manual 2-MS-SV-201B 4" - None - No C C C -/NA Safety 2-MS-SV-202B 6" - None - No C C C -/NA Safety 2-MS-SV-203B 6" - None - No C C C -/NA Safety

a. See Note 9

b. Cold < 250°F; Hot > 250°F

c. O = Open; C = Closed; Int = Intermediate; LC = Locked Closed

d. FC = Fails Closed; FAI = Fails As Is; NA = Not Applicable

e. I/O = Inside/Outside 5.2-30

f. HH = Phase 3, H = Phase 2, SI = Phase 1

Table 5.2-2 (CONTINUED)

Revision 52Updated Online 09/30/20 UNIT 2 CONTAINMENT PENETRATIONS 10 CFR 50 Essential Line Penetration Configuration Appendix J System Nominal Class Valve Number/Type/Operator Test Required Level Line Status c,d System Pen Line Per 1, 2, ESF Actuation Shut- Loss of Name No. Size UFSAR Inside Outside Inside Outside or No System Signalf Normal down Incident Pwr (I/O)e Flow Temperatureb Fluid Notes 2-MS-SV-204B 6" - None - No C C C -/NA Safety 2-MS-SV-205B 6" - None - No C C C -/NA Safety 2-MS-RV-201B 5" - None - No C C C -/FC Relief/Air C Feedwater 2-FW-72 2-FW-74 076 14" 3 No No No No - O C O NA/NA In Hot Liquid 7 Header Check Check B Feedwater 2-FW-41 2-FW-43 077 14" 3 No No No No - O C O NA/NA In Hot Liquid 7 Header Check Check A Feedwater 2-FW-10 2-FW-12 078 14" 3 No No No No - O C O NA/NA In Hot Liquid 7 Header Check Check SW to D RS 2-SW-MOV-204D 079 24" 3 None - No 1 Yes HH C C O -/FAI In Cold Liquid 8 HX Butterfly/Motor SW to C RS 2-SW-MOV-204C 080 24" 3 None - No 1 Yes HH C C O -/FAI In Cold Liquid 8 HX Butterfly/Motor SW to B RS 2-SW-MOV-204B 081 24" 3 None - No 1 Yes HH C C O -/FAI In Cold Liquid 8 HX Butterfly/Motor SW to A RS 2-SW-MOV-204A 082 24" 3 None - No 1 Yes HH C C O -/FAI In Cold Liquid 8 HX Butterfly/Motor SPS UFSAR SW from D 2-SW-MOV-205D 083 24" 3 None - No 1 Yes HH C C O -/FAI Out Cold Liquid 8 RS HX Butterfly/Motor SW from C 2-SW-MOV-205C 084 24" 3 None - No 1 Yes HH C C O -/FAI Out Cold Liquid 8 RS HX Butterfly/Motor SW from B 2-SW-MOV-205B 085 24" 3 None - No 1 Yes HH C C O -/FAI Out Cold Liquid 8 RS HX Butterfly/Motor SW from A RS 2-SW-MOV-205A 086 24" 3 None - No 1 Yes HH C C O -/FAI Out Cold Liquid 8 HX Butterfly/Motor 2-FW-131 2-FW-133 AFW to FW 087 6" 3 No No 2 Yes - C C O NA/NA In Cold Liquid 7 Check Check 2-FW-136 2-FW-138 AFW to FW 088 6" 3 No No 2 Yes - C C O NA/NA In Cold Liquid 7 Check Check

a. See Note 9

b. Cold < 250°F; Hot > 250°F

c. O = Open; C = Closed; Int = Intermediate; LC = Locked Closed

d. FC = Fails Closed; FAI = Fails As Is; NA = Not Applicable 5.2-31

e. I/O = Inside/Outside

f. HH = Phase 3, H = Phase 2, SI = Phase 1

Table 5.2-2 (CONTINUED)

Revision 52Updated Online 09/30/20 UNIT 2 CONTAINMENT PENETRATIONS 10 CFR 50 Essential Line Penetration Configuration Appendix J System Nominal Class Valve Number/Type/Operator Test Required Level Line Status c,d System Pen Line Per 1, 2, ESF Actuation Shut- Loss of Name No. Size UFSAR Inside Outside Inside Outside or No System Signalf Normal down Incident Pwr (I/O)e Flow Temperatureb Fluid Notes Condenser Air 2-VP-12 2-SV-TV-202A Ejector 089 6" 1 Yes Yes No - SI O O C NA/FC In Cold Gas Check Gate/Air Pilot Discharge Containment Purge Exhaust 2-VS-MOV-200C 2-VS-MOV-201 090 36" 5 Yes Yes No - - LC O LC FAI/FAI Out Cold Gas 4, 10 Ventilation Butterfly/Motor Butterfly/Motor System 2-VS-MOV-200D

- Yes LC O LC -/FAI Butterfly/Motor Containment Purge Supply 2-VS-MOV-200A 2-VS-MOV-200B 091 36" 5 Yes Yes No - - LC O LC FAI/FAI In Cold Gas 4, 10 Ventilation Butterfly/Motor Butterfly/Motor System 2-VS-MOV-202

- Yes LC O LC -/FAI Butterfly/Motor 2-CV-TV-250C Containment 092 2" 1 None - Yes 1 No SI INT C C -/FC Out Cold Gas Globe/Air Pilot Vacuum A 2-CV-TV-250D Pump Suction 3/8 1 None - Yes SI INT C C -/FC Globe/Air Pilot and 2-GW-TV-204 Gaseous Waste - Yes No - C C O -/FC Plug/Air Pilot SPS UFSAR Hydrogen Analyzer 2-GW-TV-205

- Yes - C C O -/FC Plug/Air Pilot 2-CV-TV-250A Containment 093 2" 1 None - Yes 1 No SI INT C C -/FC Out Cold Gas Globe/Air Pilot Vacuum B 2-CV-TV-250B Pump Suction 3/8 1 None - Yes SI INT C C -/FC Globe/Air Pilot and 2-GW-TV-200 Gaseous Waste - Yes No - C C O -/FC Plug/Air Pilot Hydrogen Analyzer 2-GW-TV-201

- Yes - C C O -/FC Plug/Air Pilot

a. See Note 9

b. Cold < 250°F; Hot > 250°F

c. O = Open; C = Closed; Int = Intermediate; LC = Locked Closed

d. FC = Fails Closed; FAI = Fails As Is; NA = Not Applicable

e. I/O = Inside/Outside 5.2-32

f. HH = Phase 3, H = Phase 2, SI = Phase 1

Table 5.2-2 (CONTINUED)

Revision 52Updated Online 09/30/20 UNIT 2 CONTAINMENT PENETRATIONS 10 CFR 50 Essential Line Penetration Configuration Appendix J System Nominal Class Valve Number/Type/Operator Test Required Level Line Status c,d System Pen Line Per 1, 2, ESF Actuation Shut- Loss of Name No. Size UFSAR Inside Outside Inside Outside or No System Signalf Normal down Incident Pwr (I/O)e Flow Temperatureb Fluid Notes Containment Vacuum & 2-CV-HCV-200 2-CV-2 094 8" 5 Yes Yes No No - LC INT LC FC/NA Out Cold Gas 4 Leakage Globe/Air Pilot Gate/Manual Monitoring A SG Recirc & 2-RT-2 2-RT-6 096 3" 3 Yes Yes No No - LC O LC NA Out Cold Liquid 4, 11 Transfer Ball/Manual Ball/Manual B SG Recirc 2-RT-21 2-RT-25 022 3" 3 Yes Yes No No - LC O LC NA Out Cold Liquid 4, 11

& Transfer Ball/Manual Ball/Manual C SG Recirc 2-RT-40 2-RT-44 114 3" 3 Yes Yes No No - LC O LC NA Out Cold Liquid 4, 11

& Transfer Ball/Manual Ball/Manual 2-SS-TV-203A 2-SS-TV-203B RHR Sample 097B 3/8 1 Yes Yes 1 No SI C INT C FC/FC Out Cold Liquid Globe/Air Pilot Globe/Air Pilot Leakage 2-LM-TV-200A 097C 3/8 1 None - Yes 1 No SI C C C -/FC Out Cold Gas 4 Monitoring Globe/Air Pilot 2-LM-TV-200B

- Yes -/FC Globe/Air Pilot 3/8 valve, 2-GW-TV-206 Gaseous Waste 100 1 None - Yes 1 No - C C C -/FC In Cold Gas 3" penetration Globe/Air Pilot 2-GW-TV-207

- Yes -/FC Globe/Air Pilot SPS UFSAR 2-FP-151 Fire Protection 101 4" 5 None - Yes No No - LC INT LC NA In Cold Liquid 4 Ball/Manual 2-FP-152

- Yes NA Ball/Manual Unit 1- AFW 2-FW-273 2-FW-272 102 6" 3 No No 2 Yes - C C O/INT NA In Cold Liquid 7 Cross Connect Check Check Reactor Cavity 2-RL-5 2-RL-3 Purification 103 3" 5 Yes Yes No No - LC O/INT LC NA In Cold Liquid 4, 10 Diaphragm/Manual Diaphragm/Manual Inlet Reactor Cavity 2-RL-13 2-RL-15 Purification 104 3" 5 Yes Yes No No - LC O/INT LC NA Out Cold Liquid 4 Diaphragm/Manual Diaphragm/Manual Outlet

a. See Note 9

b. Cold < 250°F; Hot > 250°F

c. O = Open; C = Closed; Int = Intermediate; LC = Locked Closed

d. FC = Fails Closed; FAI = Fails As Is; NA = Not Applicable 5.2-33

e. I/O = Inside/Outside

f. HH = Phase 3, H = Phase 2, SI = Phase 1

Table 5.2-2 (CONTINUED)

Revision 52Updated Online 09/30/20 UNIT 2 CONTAINMENT PENETRATIONS 10 CFR 50 Essential Line Penetration Configuration Appendix J System Nominal Class Valve Number/Type/Operator Test Required Level Line Status c,d System Pen Line Per 1, 2, ESF Actuation Shut- Loss of Name No. Size UFSAR Inside Outside Inside Outside or No System Signalf Normal down Incident Pwr (I/O)e Flow Temperatureb Fluid Notes Leakage 2-LM-TV-200C 105B 3/8 1 None - Yes 1 No SI C C C -/FC Out Cold Gas 4 Monitoring Globe/Air Pilot 2-LM-TV-200D

- Yes -/FC Globe/Air Pilot Hydrogen 2-GW-TV-211A 2-GW-TV-211B 105C 3/8 1 Yes Yes 1 No - C C O/INT FC/FC Out Cold Gas Sample Globe/Air Pilot Globe/Air Pilot 2-SI-HCV-2850A 2-SI-73 SI Test Line 106 3/4 5 No Yes No No - C/LC LC LC FC/NA Out Cold Liquid 3, 4 Globe/Air Pilot Globe/Manual 2-SI-HCV-2850B No - FC/-

Globe/Air Pilot 2-SI-HCV-2850C No - FC/-

Globe/Air Pilot 2-SI-HCV-2850D No - FC/-

Globe/Air Pilot 2-SI-HCV-2850E No - FC/-

Globe/Air Pilot 2-SI--HCV-2850F No - FC/-

Globe/Air Pilot CCW from RCP 2-CC-TV-240A 2-CC-TV-240B 110 3" 3 No No 2 No HH O O C FC/FC Out Cold Liquid 1 Thermal Barrier Globe/Air Pilot Globe/Air Pilot SPS UFSAR Containment 2-IA-TV-201A 2-IA-TV-201B 112 3" 1 Yes Yes 1 No H O O C FC/FC Out Cold Gas Instrument Air Plug/Air Pilot Plug/Air Pilot 2-SI-227 2-SI-MOV-2869A Safety Injection 113 3" 4 No Noa 1 Yes - C C O/INT NA/FAI In Cold Liquid 2, 4, 9 Check Gate/Motor 2-SI-174

- Noa LC LC LC -/NA Globe/Manual

a. See Note 9

b. Cold < 250°F; Hot > 250°F

c. O = Open; C = Closed; Int = Intermediate; LC = Locked Closed

d. FC = Fails Closed; FAI = Fails As Is; NA = Not Applicable

e. I/O = Inside/Outside

f. HH = Phase 3, H = Phase 2, SI = Phase 1 5.2-34

Revision 52Updated Online 09/30/20 SPS UFSAR 5.2-35 Tables 5.2-1 and 5.2-2 Notes

1. Component Cooling Containment Penetration Isolation:

Penetration #s: 1, 2, 4, 5, 9, 10, 11, 12, 13, 14, 16, 17, 18, 25, 26, 27, 110 These penetrations are in closed systems. Containment penetration check valves and trip valves are leak tested, but the leakage is not included in the 10 CFR 50 App. J Type B and C total leakage. During the associated penetration check valve test, the containment penetration manual isolation valve is leak tested in the reverse direction. The valve is tested with system pressure on the upstream side and the downstream side vented.

Reference:

T. S. Amendment 72/73 dated September 29, 1981.

2. Safety Injection Inside Containment Penetration Isolation:

Penetration #s: 7, 21, 23, 35, 36, 37, 60, 61, 62, 113 Inside containment penetration check valves are not Type C tested. A single valve is acceptable because the system is closed outside of containment and a single active failure does not prevent isolation.

Penetration numbers 7 and 113 have locked closed outside containment isolation valves under administrative control.

3. General Design Criteria (GDC) Compliance:

Penetration #s: 20, 24, 46, 106 Containment isolation for the above penetrations is consistent with the original design basis of the UFSAR for applicable class 5 lines. GDC 55 & 56 were not promulgated when these containment isolation configurations were designed. These penetrations are considered to meet the requirements of GDC 53 (July 1967). Type C testing is performed on the outside isolation valve only.

4. Locked Closed Containment Penetration Isolation Valves:

The following Penetrations have either one or two Containment Penetration Isolation Manual Valves locked closed. These valves are maintained under administrative control. The valves outside containment are verified locked closed periodically. The valves inside containment are verified prior to exceeding Refueling shutdown conditions.

Penetration #s:

Outside/Inside Locked: 22, 51, 54, 96, 103, 104, 114 Outside Locked: 7, 20, 24, 42, 47, 55, 57, 58, 66, 67, 68, 69, 94, 97, 101, 105, 106, 113

Revision 52Updated Online 09/30/20 SPS UFSAR 5.2-36 The following penetrations have the containment penetration Isolation MOV Breaker locked open with the valve in the closed position. The containment isolation MOVs also have their handwheel locked. The valves and breakers are verified locked closed periodically.

Penetration #s: 90, 91

5. Seal Water To RCPs:

Penetration #s: 35, 36, 37 Needle valves are throttled open and administratively controlled. These lines remain open after a safety injection signal and contribute to the total injection flow while cooling the RCP seals. The incoming lines have a check valve inside containment and a local manual valve (throttle valve) outside containment combined with both a closed system and continuous water seal at a pressure sufficient to preclude containment atmospheric leakage.

6. Open Containment Penetration Isolation Valves:

Penetration #s: 61,74,75,76 Penetration 61 has its breaker de-energized with the valve in the open position. This penetration is for Low Head Safety Injection Discharge to the Reactor Coolant System Cold Legs. The Safety Injection System outside containment has an external leakage Technical Specification requirement which provides limits to ensure acceptable leakage during accidents.

Penetrations 74, 75, & 76 are the normal steam supply to the turbine driven auxiliary feedwater pump and are normally open. These valves are closed in accordance with emergency procedures to provide steam generator isolation in the event of a steam generator tube rupture.

7. Main Steam Containment Penetration Isolation:

Penetration #s: 39, 40, 41, 73, 74, 75, 76, 77, 78, 87, 88, 102 No testing required - per App. J.

These penetrations are in systems directly connected to the steam generator secondary side and, therefore, are considered a closed system (an extension of the primary containment). In addition, the steam generator remains at a pressure greater than peak accident pressure for at least the first hour and is not considered a credible leakage path from containment.

Reference:

T. S. Amendment 72/73 dated September 29, 1981.

An air test is performed prior to a Type A Test. If a Type A Test is not performed, station procedures verify that no external leakage exists.

8. Service Water To Recirculation Spray Heat Exchanger Containment Penetration Isolation:

Penetration #s: 79, 80, 81, 82, 83, 84, 85, 86

Revision 52Updated Online 09/30/20 SPS UFSAR 5.2-37 These penetrations are in closed systems. Each train is leak tested but the leakage is not included in the 10 CFR 50 App. J Type B and C total leakage. The valves in these lines remain open during a design basis accident.

Reference:

T. S. Amendment 72/73 dated September 29, 1981.

9. Water Filled Penetrations:

Penetration #s: 7, 15, 21, 23, 46, 60, 61, 62, 66, 67, 68, 69, 113 These penetrations are in systems that are water filled and/or normally operating under accident conditions at a pressure greater than peak accident pressure. Therefore, these penetrations are not considered credible leakage paths from containment.

Reference:

NRC SER dated November 21, 1988.

10. Type C Reverse Direction Tests:

Penetration #s: 90, 91, 103 Type C testing of the inboard isolation valve for penetration 90, 91, and 103 is performed in the reverse direction due to the existing piping configuration. For the type of inboard isolation valves used (diaphragm and butterfly), leakage is the same in either direction.

11. Steam Generator Recirc. And Transfer System:

Penetration #s: 22, 96, 114 Due to the piping configuration inside containment, the above penetrations were added to the Type C testing program to ensure that any leakage through these valves is identified and corrected.

Revision 52Updated Online 09/30/20 Figure 5.2-1 EXAMPLE OF VALVE ARRANGEMENT CONTAINMENT ISOLATION SYSTEM SPS UFSAR 5.2-38

Revision 52Updated Online 09/30/20 SPS UFSAR 5.3-1 5.3 CONTAINMENT SYSTEMS 5.3.1 Ventilation Systems 5.3.1.1 General Description Containment ventilation consists of an air cooling recirculation system, an air cooling control rod drive mechanism (CRDM) system, a filter system, and a purge system. They are shown on Reference Drawing 1. A review of the effects of the power uprate to a core power of 2587 MWt was conducted and the containment air recirculation system was found to be adequate.

The reactor coolant pump motors (Section 4.2.2.4) are cooled with an integral component cooling water system. The air-cooling recirculation system, air-cooling control rod drive mechanism system, and reactor pump motor coolers provide the total cooling required to limit the bulk air temperature to 125°F during normal summer operations. The minimum temperature allowed is 75°F.

5.3.1.2 Design Basis The ventilation systems were originally designed to limit the containment bulk air temperature to below 105°F when three of the recirculating fans are running, three CRDM-cooling systems are running, and the cooling systems for the reactor coolant pump motors are functioning. Operating experience has demonstrated that the heat load in containment exceeds the original design estimates but that the ventilation systems are adequate to maintain the containment bulk air temperatures less than 125°F. The value of 125°F is the maximum containment initial temperature assumed in the design basis accident containment response evaluations.

The recirculation fan and cooling coil systems are designed to remove their portion of the heat load, under subatmospheric operating conditions, when supplied with 680 gpm of 70°F water. The relative humidity during both summer and winter operations is about 40%, with 70°F cooling water entering the recirculation coolers and 105°F bulk air temperature.

The control rod drive mechanism cooling system is designed to meet the required heat removal load when three fan cooling coil units are operating.

The two inside containment iodine filtration units are designed to remove airborne activity that may be released by nominal operational reactor coolant system leakage during subatmospheric operations.

The purge system is designed to purge the containment after the pressure has been raised to within 1-inch water gauge of atmospheric. The purging rate can be varied in steps from approximately one change per day to one change per hour. The purge exhaust is charcoal filtered for airborne radioactivity removal. If fuel is being handled within containment, the purge exhaust air flow may be directed through one or two safety related filters depending on the purge flow rate, or it may be routed through a non-safety related charcoal filter. Without fuel handling inside

Revision 52Updated Online 09/30/20 SPS UFSAR 5.3-2 containment, the purge exhaust air may be routed through a nonsafety related charcoal filter. The purge system design also provides ventilation and space heating for cold weather refueling and maintenance.

Principal component data are given in Table 5.3-1.

5.3.1.3 System Descriptions 5.3.1.3.1 Air-Cooling Recirculation System The air-cooling recirculation system consists of three 75,000-cfm (at subatmospheric conditions) fan and cooling coil banks discharging into a common ring duct from which cool supply air is ducted to the various compartments.

A vane-axial fan is installed in ductwork routed from a recirculation system discharge plenum at Elevation -27 ft. 7 in. to the containment dome area. The fan supplies approximately 10,000 cfm at subatmospheric conditions to the containment dome area to prevent warm air stratification.

Return-air transfer ducting is provided to prevent the short-circuiting of return-air flow paths, and thereby ensure cooling-air flow to the area above the operating floor. Three 36-inch-diameter ducts, each with a vane-axial fan, take suction from approximately 30 feet above the operating floor and discharge below Elevation -3 ft. Each duct has an air flow capacity of approximately 25,000 cfm. The vane-axial fans are operated with normal station power.

The recirculation system cooling coils are served primarily by the component cooling water system (Section 9.4.3.1), with backup cooling available from the chilled component cooling water system (Section 9.4.3.3).

Containment cooling design heat loads are given in Table 5.3-2.

5.3.1.3.2 Control-Rod Drive Mechanisms The control-rod drive mechanisms are cooled by three 24,000-cfm (at subatmospheric conditions) fan and coil banks. All three units are required to provide the essential heat removal during normal operation. On Unit 1, air is drawn through the top of the shroud down over the CRDM coil stacks. On Unit 2, air is drawn through the sides of the shroud and up over the CRDM coil stacks. The air then circulates through the mechanisms, and discharges back to the containment through the cooling coils. Each fan and coil unit has two 100% flow capacity fans for redundancy.

5.3.1.3.3 Iodine Filtration Units The inside containment iodine filtration units are self-contained packages installed on the lower level of each containment. Each consists of a 2000-cfm fan with roughing, high efficiency particulate air (HEPA), and charcoal filters installed within concrete shielding. The units are remotely operable from the control room.

Revision 52Updated Online 09/30/20 SPS UFSAR 5.3-3 5.3.1.3.4 Purge System The containment purge supply and exhaust subsystems consist of supply and exhaust ducting arranged to ventilate either containment when the pressure has been raised to within 1-inch of water gauge of atmospheric pressure. The normal station powered supply fans are not operated to assure that the containments are negatively pressurized. The exhaust fan(s) draw outside air through low efficiency filters and a winter heating coil into containment through two isolation butterfly valves. The purge exhaust air is drawn through low-level ducts within the containment. This air flows through two isolation butterfly valves that may connect to one or both safety related filter trains or one non-safety-related filter train (Section 9.13) through two isolation dampers installed in series. The outer exhaust valve is fitted with an 8-inch bypass valve to permit reduced purge flow if required. An 18-inch pressure-equalizing valve is installed on the outside of the containment structure between the supply system penetration valves to bring the containment up to atmospheric pressure on shutdown.

The motor-operated butterfly valves are located on either side of the containment penetrations for pressure integrity. The two isolation trip dampers in series connecting the purge exhaust ducting to the safety-related filter inlet header are air operated and are designed to fail in the closed position on loss of air. The air is supplied from either the station compressed air system or through an air accumulator sized to store sufficient air to keep the dampers open for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

The butterfly valves, air-operated isolation trip dampers, and ducting leading to the safety-related filter inlet header, including the safety-related filter system, are constructed to meet seismic qualification requirements. The valves are normally kept closed except during unit shutdown when they are opened for ventilation, heating, and purging.

The purge system fans, isolation valves, and bypass and pressure equalizing valves are remote manually operated from the control room for system alignments. If a safety-injection signal is received, the purge supply fans will trip off and the isolation valves and dampers will automatically shut to isolate the containment and allow the safety-related filters to treat the air exhausted from the emergency core cooling equipment areas.

Since the Surry units have subatmospheric containments, containment purging operations are not allowed unless the unit is in cold shutdown or refueling conditions. Technical Specifications require that containment integrity be established before increasing reactor coolant temperature above 200°F, and that containment air partial pressure be within specifications before exceeding 350°F. Technical Specifications also require that the containment vacuum be maintained for all plant conditions under which the engineered safeguards systems are required to be operational. Purging is precluded under these conditions because physical limitations prohibit containment purging unless the containment vacuum is broken.

The connection to the non-safety-related charcoal filter is located between the isolation valves and isolation dampers. Although seismically supported, the connection is isolated seismically via a flexible joint from the seismic duct. When not in use, the connection is closed by

Revision 52Updated Online 09/30/20 SPS UFSAR 5.3-4 installing the closed side of a spectacle flange. The maximum purge rate through this path is limited to 20,000 cfm as the filter also serves the Auxiliary Building General Exhaust.

5.3.1.4 Design Evaluation Whenever the three main recirculation fan and coil units, the three CRDM fan and coil units, and the main coolant pump cooling systems are operating, the containment bulk air temperature will be maintained below 125°F. Two of the three fans in the recirculation system will continue to operate under limited main coolant leakage conditions that result in containment pressures up to but not exceeding the Consequence Limiting Safeguards (CLS) high-high containment pressure actuation setpoint (Section 7.5.1.2). The third fan will continue to operate, if normal station power is available, until stopped either manually or by actuation of an electrical fault protection device. This may provide sufficient heat removal to permit reactor shutdown under limited leakage conditions without resorting to caustic spray injection.

The inside containment filter units will remove the airborne iodine and particulate radioactivity that could result from nominal operational leakage during subatmospheric operations.

The purge system provides the capability to change the containment air and remove radioactivity, if required, before entry for refueling and maintenance. The purge system is designed for one air change per hour and to maintain a minimum of 60°F inside the containment.

5.3.1.4.1 Incident Control During normal operation of the plant, the containment purge system is not in use.

After unit shutdown and cooldown, purging of the containment can take place. The purge exhaust air may be directed to either the non-safety-related or safety-related ventilation filters in the auxiliary building if fuel is being handled inside containment, but no filtration is credited in the analysis. The analysis of the fuel handling accident in containment does not require that containment integrity be established prior to fuel movement. The purge design flow through the non-safety-related filter is 20,000 cfm with a limit of 30,000 cfm through the safety-related filters when containment integrity is established. If containment integrity is not established, the maximum purge exhaust rate equals the maximum safety-related fan flow limit of 39,600 cfm.

The physical design and installation of the duct systems preclude exceeding these limits. The discharge of the safety-related filters and non-safety-related filter are monitored by the same system for radioactivity prior to release. Should a LOCA signal from the other unit be received, the air-operated isolation dampers will fail closed and allow the safety-related filters to treat the air exhausted from the ECCS areas. As described in Section 9.13.4.1, if a safety injection actuation occurs and auto alignment of the ventilation system is defeated, manual action is required to realign the system to the ECCS filtration mode. An alarm is received in the main control room if the purge is not realigned following a safety injection signal. This condition is not expected however, since defeating the automatic realignment is no longer credited in the fuel

Revision 52Updated Online 09/30/20 SPS UFSAR 5.3-5 handling accident analysis and procedural controls have been established to eliminate operating with automatic alignment defeated.

5.3.1.4.2 Malfunction Analysis The three air-cooling recirculating subsystems are required to maintain the containment bulk air design temperature during warm weather.

The three CRDM ventilation systems are required to provide the essential cooling.

The two inside containment self-contained particulate and iodine filter packages provide redundancy for small leakage rates.

During refueling, a high-radiation signal from the containment gas or particulate monitors or the manipulator crane area monitor will automatically trip the containment purge supply fans and close the containment isolation control valves. This automatic function is not credited in the fuel handling accident nor is it required to be operable. The operability of the containment gas and particulate monitors and the manipulator crane area monitor is relied upon in conjunction with communications to provide a timely and valid indication of a fuel handling accident in the containment.

5.3.1.5 Tests and Inspections The systems are inspected, tested, and pneumatically balanced upon installation. Particulate and charcoal filters are individually tested before shipment. The filters in the purge exhaust flowpath are tested after installation and can be periodically tested for leakage and dioctylphthalate smoke test efficiency as described in Section 9.13.5.

5.3.2 Leakage-Monitoring System The containment leakage-monitoring system was used for the preoperational integrated test of the containment and is used for the periodic measurement of leakage into the containment during normal unit operation. Section 5.3.2.2 describes two methods that can be used to monitor containment leakage: the reference volume method and the absolute method. The absolute method of leakage rate testing is the preferred method of testing due to overall test measurement accuracy. The reference volume method is not used but could be if necessary.

5.3.2.1 Design Basis The leakage-monitoring system is not operational since the reference volume method is no longer used in containment leakage rate testing. The system can be made operational if necessary; however, the connections to the manometers would have to be reestablished since these are not used in the absolute system. In addition, the system would require necessary maintenance prior to use.

Revision 52Updated Online 09/30/20 SPS UFSAR 5.3-6 The absolute method applies the perfect gas law to measured changes in the containment pressure, temperature, and relative humidity, and reduces the data by means of a mathematical least-squares linear regression calculation.

Design data for the leakage-monitoring system components are given in Table 5.3-3. The system was designed in accordance with an Atomic Energy Commission Safety Guide entitled.

Reactor Containment Leakage Testing and Surveillance Requirements.

5.3.2.2 Description The absolute system consists of instruments to measure and record containment pressure, temperature, and relative humidity. These measurements are recorded at different times and the air mass in the containment is determined by the perfect gas law. Data compiled in this fashion are fitted to a linear regression equation relating time and the mass of air in the containment to leakage. Statistical methods are used to compute the variance in the results and thereby evaluate the error.

The reference volume method is based on determining the change in the pressure differential between the sealed reference system and the open-ended system of containment as caused by containment leakage. This method is limited by the difficulty in ensuring and validating the system integrity. Thus it is difficult to obtain accurate results. For this reason, the absolute method is the method of choice.

The accuracy of the test used to monitor containment leakage (ILRT) is verified by a supplemental test (superimposed method). The supplemental test is conducted for sufficient duration to accurately establish the change in leakage rates between the ILRT and the supplemental test. The results from the supplemental test are acceptable if the difference between the supplemental test data and the data obtained from either the reference volume test method or the absolute test method is within 0.25 La, where La is the maximum allowable leakage rate at the calculated peak accident pressure.

5.3.2.3 Design Evaluation Periodic leakage monitoring is performed as required by Appendix J by the absolute method. This method is verified by a supplemental test (superimposed method). The absolute method is sufficiently accurate to establish that the containment leakage rate is less than the Technical Specification required leakage limit.

As part of the containment isolation system (Section 5.2), each open leakage-monitoring line penetrating the containment structure is provided with two automatic trip valves. In the event of an incident, these lines are closed and no leakage to the environment occurs. The containment leakage-monitoring system tubing, as an extension of the containment, is designed to withstand the pressure and temperature expected during an incident.

Revision 52Updated Online 09/30/20 SPS UFSAR 5.3-7 5.3.2.4 Tests and Inspections The temporary instruments used to perform the Type A test are calibrated before each test as required by Appendix J.

5.3.3 Spray Systems The containment-spray systems, which consist of containment-spray subsystems and recirculation-spray subsystems, are described in detail in Section 6.3.1.

The containment-spray subsystems operate during the depressurization period after a LOCA.

The containment-spray subsystems transfer chilled water from the refueling-water storage tank to the containment through the containment-spray headers. The chilled water removes sensible heat from the containment, resulting in a decrease in containment temperature and pressure. The containment-spray pumps and recirculation-spray pumps are driven by electric motors powered from either normal or emergency power sources as described in Chapter 8. The recirculation-spray subsystems recirculate water from the containment sumps through heat exchangers to the recirculation-spray headers. These subsystems provide a net heat removal from the containment. They are designed to aid in lowering the temperature and pressure within the containment and to maintain the containment subatmospheric once it is depressurized.

5.3.4 Vacuum System The containment vacuum system is used to obtain the initial subatmospheric pressure in containment and to maintain that pressure during unit operation. It consists of a steam jet air ejector, two mechanical vacuum pumps, and the required piping, valves, and instrumentation.

Many of the containments currently in use operate at approximately atmospheric pressure.

Following a LOCA, the containment pressure rises. Although the pressure can be reduced rapidly at first, the pressure-time transient curve is asymptotic to atmospheric pressure and the outleakage of fission products may continue for some time.

The subatmospheric containment concept is based on the normal operation of the containment below atmospheric pressure. Following a LOCA, the pressure will rise to above atmospheric pressure with subsequent outleakage. However, the containment temperature and pressure can be reduced rapidly, returning the containment to subatmospheric pressure and thus terminating outleakage. This is because the pressure-time curve crosses atmospheric pressure rapidly, being asymptotic to the initial subatmospheric pressure. The engineered safeguards used for pressure reduction are designed to limit the outleakage of fission products to an acceptable quantity.

Before unit operation, the containment pressure is at atmospheric pressure. During the reactor system heat-up, the pressure is reduced with a steam ejector so that containment operating pressure is reached before reactor power operation.

Revision 52Updated Online 09/30/20 SPS UFSAR 5.3-8 A subatmospheric containment pressure is maintained whenever the reactor is operating at or near design pressure and temperature. During operation, inleakage occurs and the vacuum system maintains the containment subatmospheric. The air pumped out is metered to provide a constant indication of containment system integrity. After reactor shutdown and reactor coolant system depressurization and before refueling or extended maintenance, the containment pressure is returned to atmospheric.

The initial operating temperatures for the subatmospheric and atmospheric containment are approximately equal. However, the subatmospheric containment has a lower initial air pressure.

It follows that for any containment pressure, the subatmospheric containment must have a higher steam partial pressure, which results in a higher containment temperature. The atmospheric containment would have an initial temperature of about 105°F and a temperature of approximately 273°F if allowed to rise to 50 psig following a LOCA. The subatmospheric containment starting at 10.0 psia would have an initial temperature of about 105°F and a temperature of approximately 285°F if allowed to rise to 50 psig following a LOCA.

The higher temperature for the subatmospheric containment results in the more effective operation of both static heat sinks and engineered heat removal systems. This higher temperature makes it possible to use reasonably sized water spray systems to return the containment to subatmospheric pressure and then to hold the containment subatmospheric.

The subatmospheric containment is based on known technology; 10 psia corresponds to a pressure altitude of about 10,000 feet. No difficulties have been encountered in the design and procurement of equipment for operation at this sub-atmospheric pressure. The containment is designed to resist the external pressure without new technology. The concept provides for an increase in unit safety through the reduction in possible activity release. The need for charcoal air recirculation filters and fans operating in a steam environment is eliminated and dependence is placed on a redundancy of simple spray systems.

The advantages of the subatmospheric containment are best realized if pressure is reduced rapidly to a subatmospheric level. A review of various heat removal systems indicates that the injection of cold water into the containment is the most rapid and dependable means of depressurization. This water must be borated when used in conjunction with shim-controlled reactors, since it will ultimately be recirculated to the reactor for core cooling. Depressurization is accomplished through the combined operation of the containment spray, which provides a cold-water heat sink, and the recirculation spray, which removes heat from the containment.

Transient heat transfer calculations required to size the spray systems and consider the effect of static heat sinks are well understood.

Figure 5.3-1 shows typical pressure transient curves for comparable atmospheric and subatmospheric containments following a LOCA. As shown, the pressure in the atmospheric type of containment design drops to a low value, but outleakage continues indefinitely, while

Revision 52Updated Online 09/30/20 SPS UFSAR 5.3-9 outleakage is terminated in the subatmospheric containment type as soon as pressure is reduced below atmospheric.

5.3.4.1 Design Bases The containment vacuum system is designed to perform the following three functions:

1. Evacuation of the containment from atmospheric pressure and maintenance of the subatmospheric pressure used for normal operation.

2. Removal of air from the containment to compensate for containment inleakage during normal operation.

3. Removal of steam and air from the containment to compensate for leakage following a design-basis accident.

The system is designed to reduce the containment pressure from atmospheric to within the Technical Specification air partial pressure limits in a time period compatible with the station start-up schedule by using the containment vacuum steam jet air ejector. To compensate for inleakage, each vacuum pump is capable of removing 5 cfm. The containment will be at a variable subatmospheric operating air partial pressure of between 10.1 and 11.3 psia whenever the reactor coolant system is at or exceeds hot standby pressure and temperature (450 psig and 350°F, minimum).

5.3.4.2 Description The containment vacuum system consists of a steam jet air ejector and two mechanical vacuum pumps, with the required piping, valves, and instrumentation, as shown in Figure 5.3-2.

Pump control and system instrumentation are discussed in Section 7.5. Containment vacuum system component data are given in Table 5.3-4.

The steam jet air ejector removes air from the containment to create the initial vacuum before operations, and operates on 150-psig steam provided by the auxiliary steam system, as discussed in Section 10.3.2. The air ejector is sized to draw down the containment pressure to 9.5 psia within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

There are two five-cfm mechanical vacuum pumps, each of which can provide more than 100% of the required pumping capacity. The pumps are capable of being operated from the emergency diesel generators discussed in Section 8.5 and discharge to the process vent through the charcoal filters of the gaseous waste disposal system (Section 11.2.5).

Each containment vacuum pump is located inside a leaktight containment vacuum pump tank. A pipe running from the containment to the containment vacuum pump tank transports air from the containment to the containment vacuum pump inside the tank. The leaktight tank prevents seal leakage from the containment vacuum pump from escaping to the atmosphere.

Revision 52Updated Online 09/30/20 SPS UFSAR 5.3-10 5.3.4.3 Evaluation The steam jet air ejector which is used only for initial reduction of the pressure in the containment and the vacuum pumps which are normally used during plant operations are not part of the engineered safeguards. The system is only operated when the reactor coolant system is below a temperature of 200°F. Each of the mechanical vacuum pumps is capable of removing containment inleakage and maintaining the required vacuum. The pumps are sized for intermittent and automatic operation so that, on a continual operational basis, they have the capacity for removing inleakage at a rate of about four times the design value.

The operation of the containment vacuum system is not required for several months after the design-basis accident. The containment is designed and demonstrated to have a leak rate not exceeding 0.1% per day at the design pressure. It is therefore reasonable to assume that the leakage rate, under normal plant operation and during the postaccident period, when the containment has been returned to subatmospheric pressure, will be considerably less. However, assuming the leakage rate to be independent of containment pressure, the rate of 0.1% per day would correspond to a leakage flow of 1.2 scfm and would increase containment pressure approximately 0.01 psi per day (i.e., 100 days are required to increase pressure 1 psi).

Offsetting the tendency for the containment pressure to rise as a result of the inleakage of air after the design-basis accident is the reduction in the containment pressure effected by the recirculation spray subsystems ability to cool down the containment atmosphere further, since decay heat evolution decreases with time.

The effect of long-term leakage must be considered. Ultimately, air inleakage could result in the containment pressure increasing to atmospheric, with temperature fluctuations possibly raising the pressure to slightly above atmospheric. When the vacuum system resumes operation, it discharges through charcoal filters, which are part of the gaseous waste disposal system.

Therefore, the amount of activity released to the environment is minimal.

Excess depressurization of the containment is not considered credible, since the vacuum pumps have such a small capacity when compared to the containment free volume. It requires a vacuum control system signal failure and uninterrupted operation of the vacuum pumps for approximately 17 days to result in a 1-psi decrease in containment pressure with the required capacity of 5 cfm. Alarms are provided in the control room to indicate that the containment pressure has dropped below the normal operating level. The alarm points are set at values above the design loading of the containment. This allows sufficient time for the operator to take corrective action before the pressure drops below the design loading of the containment.

Administrative procedures require that the steam ejector be isolated to prevent its operation at any time other than that required during start-up of the unit. The steam ejectors are secured during normal operation by administrative control to preclude the possibility of excessive depressurization.

Revision 52Updated Online 09/30/20 SPS UFSAR 5.3-11 The minimum credible pressure expected for this containment design is caused by inadvertent operation of the containment spray system.

An evaluation of this event is performed using Charles Law and the following initial conditions.

Minimum air partial pressure (P1) 9.85 psia TS limit of 10.1 psia - 0.25 psi uncertainty Maximum bulk air temperature (T1) 125.5°F TS limit of 125°F + 0.5°F uncertainty Minimum RWST temperature (T2) 32°F Bounding minimum value Saturation pressure at T2 (Psat) 0.09 psia ASME Steam Tables at 32°F Using Charles Law for the air partial pressure (temperatures converted to Rankine), the final pressure in containment is calculated:

T2 P total = P air + P vapor = ------ P 1 + P sat ( T 2 ) = (---------------------------------------------------------

460 + 32 ) ( 10.1 - 0.25 )- + 0.09 = 8.37 psia T1 ( 460 + 125.5 )

For an inadvertent CS actuation starting at the TS minimum air partial pressure of 10.1 psia and TS maximum air temperature of 125°F, the containment liner meets the following criteria without operator action to terminate CS.

1. Minimum containment pressure is greater than the bottom mat liner internal design pressure of 8.0 psia.

2. Minimum containment pressure is greater than the containment shell and dome internal design pressure of 3.0 psia.

5.3.4.4 Tests and Inspections The steam jet air ejector and the mechanical vacuum pumps are not part of the engineered safeguards. Therefore, preoperational inspection of this simple mechanical device is satisfactory.

The mechanical vacuum pumps were operated during the initial containment leakage rate test described in Section 5.3.2 and demonstrated adequate capacity to remove inleakage. During normal unit operation, they are alternated in service periodically, so their performance status is continually available.

5.3.5 Hydrogen Analyzer System The requirements of TMI-2 Short Term Lessons Learned, NUREG 0578 and subsequent clarifications contained in the NRC letter dated October 30, 1979, required that there be a continuous indication of hydrogen concentration in the containment atmosphere provided in the control room. As a result, redundant hydrogen analyzers qualified to IEEE 323-1974 and IEEE 344-1975 were added with the capability of measuring over the range of 0 to 10% hydrogen concentration with containment conditions of 9 psia to 60 psia and 100% humidity. The redundant qualified hydrogen analyzers are shared by Units 1 and 2a transfer switch with control circuitry provides for the capability of Unit 1 to utilize both analyzers or for Unit 2 to utilize both

Revision 52Updated Online 09/30/20 SPS UFSAR 5.3-12 analyzers. This same circuitry allows for the operation of the direct operating electric solenoid containment isolation valves to the hydrogen analyzers.

Hydrogen analyzer H2A-GW-104-1 has the capability of being powered from the Orange Train of Unit 1 or Unit 2 vital power and hydrogen analyzer H2A-GW-204-1 has the capability of being powered from the Purple Train of the Unit 1 or Unit 2 vital power via switchable power sources.

Each analyzer has the capability to obtain an accurate sample within 90 minutes of the initiation of safety injection. Hydrogen concentration measurements will be indicated and recorded in the control room. The 90-minute timeframe is based upon the functional requirements provided in RG 1.7, Revision 3. Compliance with RG 1.7 ensures that indication of hydrogen concentration in the containment atmosphere is available in a timely manner to support the Emergency Plan (and related procedures) and related activities such as guidance for the severe accident management plan (References 1 & 2).

A qualified heat tracing system was added to the sample lines to each hydrogen analyzer in order to maintain a truly representative sample of containment atmosphere. The heat tracing system will be initiated on a safety injection signal.

A supply of oxygen gas is available to the hydrogen analyzer for use as the reagent gas for hydrogen recombining and a supply of hydrogen gas is available as the calibration gas.

5.3 REFERENCES

1. Letter from L.N. Hartz to USNRC, Virginia Electric and Power Company, Dominion Nuclear Connecticut, Inc., Surry Power Station Units 1 and 2, North Anna Power Station Units 1 and 2, Millstone Power Station Units 2 and 3, Application for Technical Specification Improvement to Eliminate Requirements for Hydrogen Recombiners and Hydrogen Monitors Using the Consolidated Line Item Improvement Process, Serial No.04-386, dated September 8, 2004.

2. Letter from USNRC to D.A. Christian, Surry Power Station, Units 1 and 2 - Issuance of Amendments on Elimination of Requirements for Hydrogen Monitors Using the Consolidated Line Item Improvement Process (TAC Nos. MC4393 and MC4394), dated March 22, 2005.

Revision 52Updated Online 09/30/20 SPS UFSAR 5.3-13 5.3 REFERENCE DRAWINGS The list of Station Drawings below is provided for information only. The referenced drawings are not part of the UFSAR. This is not intended to be a complete listing of all Station Drawings referenced from this section of the UFSAR. The contents of Station Drawings are controlled by station procedure.

Drawing Number Description

1. 11448-FB-006A Flow/Valve Operating Numbers Diagram: Air Cooling and Purging System, Unit 1 11548-FB-006A Flow/Valve Operating Numbers Diagram: Air Cooling and Purging System, Unit 2

Revision 52Updated Online 09/30/20 SPS UFSAR 5.3-14 Table 5.3-1 PRINCIPAL COMPONENT DATA - CONTAINMENT SYSTEMS Units for Normal System Units Installed Unit Capacity Operation Containment Recirculating a Cooling coil banks 3 1200 MBh 3 Fans 3 75,000 cfm 3 Fan pressure 5 in. W.G.

Fan motors 3 125 hp 3 Control-Rod Drive Cooling a Cooling coil banks 3 726 MBh 3 Fans 6 24,000 cfm 3 Fan pressure (two-stage fans) 14.0 in. W.G.

Fan motors 6 40 hp 3 Purge Supply Plenum 1 30,000 cfm 1 Fans 2 15,000 cfm 0 Fan pressure 5 in. W.G.

Fan motors 2 15 hp 0 Heating coil 1800 MBh Roughing-filter bank 1 30,000 cfm 1 Carbon Filter Banks Safety related number 2 36,000 cfm 1 Roughing cells, per unit 30 1200 cfm 30 HEPA cells, per unit 30 1200 cfm 30 Charcoal cells, per unit 60 600 cfm 60 Inside containment number 2 2000 cfm 0

a. Fan data is at atmospheric conditions.

Revision 52Updated Online 09/30/20 SPS UFSAR 5.3-15 Table 5.3-2 CONTAINMENT COOLING DESIGN HEAT LOADS Heat Load Air Flow Heat Source (1000 Btu/hr) (1000 cfm)

Steam generator cubicle A 1027.0 58 (Elevation 3 ft. 6 in. to 47 ft. 4 in.)

Steam generator cubicle B 1027.0 58 (Elevation 3 ft. 6 in. to 47 ft. 4 in.)

Steam generator cubicle C 1040.1 59 (Elevation 3 ft. 6 in. to 47 ft. 4 in.)

Pressurizer cubicle 172.7 10.5 (Elevation 3 ft. 6 in. to 47 ft. 4 in.)

Operating floor (Elevation 47 ft. 4 in.) 145.1 8 Annulus area 416.0 0a Reactor cavity (Elevation 27 ft. 7 in.) 375.7 22.5 Incore instrument drive (Elevation 17 ft. 4 in.) 190 9 Elevation 27 ft. 7 in. (general) 100.9 0a Recirculation fans 462.0 0a Leakage Primary water, sensible heat 68.7 0a Primary water, latent heat 474.7 0a Main stream, sensible heat 178.0 0a Main stream, latent heat 1057.0 0a Total 6733.6 225 Recirculation Cooling Coil Conditions Air entering condition 105°F Air leaving condition 75°F dry bulb Cooling water entering condition, per coil 680 gpm Temperature entering 70°F Temperature leaving 74.5°F Chiller Capacity (Units 1 & 2)

For 95°F entering bearing cooling-water 400 tons; 4,800,000 Btu/hr temperature Sensible Heat Loads Calculated equipment loads plus 15% lighting 4 W/ft2

a. Air circulation path

Revision 52Updated Online 09/30/20 SPS UFSAR 5.3-16 Table 5.3-3 LEAKAGE-MONITORING SYSTEM COMPONENT DESIGN DATA Reference Volume System (Installed But Not Used)

Sealed-System Bulb Number Open Tap Number Bulb Location in Containment 1 1 Steam generator cubicle 1A 2 2 Steam generator cubicle 1B 3 3 Steam generator cubicle 1C 4 4 Pressurizer cubicle 5 - Area above 47 ft. 4 in.

6 5 Area above 47 ft. 4 in.

7 - Area above 47 ft. 4 in.

8 6 Area above 47 ft. 4 in.

9 7 Dome 10 - Dome 11 - Dome 12 8 Annulus 13 - Annulus 14 - Annulus 15 9 Annulus 16 - Lower volume 17 - Lower volume 18 10 Lower volume Atmospheric Manometer (Installed But Not Used)

Type Dual Tube, well type Range 0-120 in. Hg Number of scales 2, one vernier for each scale Fill liquid Mercury

Revision 52Updated Online 09/30/20 SPS UFSAR 5.3-17 Table 5.3-3 (CONTINUED)

LEAKAGE-MONITORING SYSTEM COMPONENT DESIGN DATA Differential Manometer (Installed But Not Used)

Type U-tube type, wall mounting Range 0-60 in.

Reading accuracy To 0.01 in. fluid with vernier Fill liquid D-3166 red fluid Absolute Method System (Servomanometer) (Installed But Not Used)

Type Precision cistern Absolute pressure range 0-120 in. Hg Reading accuracy to 0.001 in.

Electrical output 0-5V dc Compensation Temperature

Revision 52Updated Online 09/30/20 SPS UFSAR 5.3-18 Table 5.3-4 CONTAINMENT VACUUM SYSTEM COMPONENT DESIGN DATA Containment Vacuum Pumps Number 4 (2 per unit, 1 required)

Type Rotary vane, oil free Power source Station electric or standby generators Capacity 5 cfm Steam Jet Ejector Number 2 (1 per unit)

Power source 150 psig steam Capacity 52,000 lb of air in 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> Containment Vacuum Pump Tank Number 4 (2 per unit)

Volume 10 ft3 Design pressure Full vacuum Design temperature 300°F Operating pressure 9.5 psia Operating temperature 105°F Material A285 GR C

Revision 52Updated Online 09/30/20 Figure 5.3-1 TYPICAL CONTAINMENT PRESSURE TRANSIENT CURVES; SURRY POWER STATION SPS UFSAR 5.3-19

Revision 52Updated Online 09/30/20 SPS UFSAR 5.3-20 Figure 5.3-2 CONTAINMENT VACUUM SYSTEM

Revision 52Updated Online 09/30/20 SPS UFSAR 5.4-1 5.4 CONTAINMENT DESIGN EVALUATION The reactor containment is maintained at a subatmospheric pressure in which the air partial pressure varies between approximately 10.1 and 11.3 psia. The containment shell and dome are designed to withstand an internal pressure as low as 3 psia, and the containment bottom mat liner is designed to withstand an internal pressure as low as 8 psia. The Technical Specifications specify the partial pressure limitations as a function of service-water temperature. The allowable variation is based on the ability of the containment heat removal systems to depressurize the containment and depends on seasonal temperature changes in the service water and thermal recirculation effects. The containment design pressure is 45 psig, which is greater than the peak post-LOCA pressure, based on a double-ended hot-leg rupture. The containment returns to subatmospheric pressure within 60 minutes of the occurrence of the accident, thus terminating outleakage from the containment (Section 14.5.5). This original design criterion was modified in conjunction with the analyses for implementation of the alternative source term. The modified criteria require that, following the LOCA, the containment pressure be less than 1.0 psig within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and less than 0.0 psig within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. The radiological consequences analysis demonstrates acceptable results provided the containment pressure does not exceed 1.0 psig for the interval from 1 to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> following the Design Basis Accident. Beyond 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, containment pressure is assumed to be less than 0.0 psig, terminating leakage from containment.

A subatmospheric containment limits the outleakage of fission products to meet 10 CFR 50.67 criteria for the design-basis accident as discussed in Section 14.5.6, using conventional spray cooling as described in Section 6.3.1.

At the design containment leak rate of 0.1% of contained volume per day, air inleakage is not significant for a considerable length of time after a LOCA. Ultimately, air inleakage, combined with ambient barometric pressure fluctuations, could result in a containment pressure slightly above atmospheric. To prevent this, the vacuum system maintains the containment pressure at several inches of mercury below the lowest expected atmospheric pressure during normal operation. The vacuum cannot be lost rapidly because of the inherent low-leakage design features of the containment.

The containment isolation features, such as penetrations, access hatches, and isolation valves, have been designed so that double barriers or seals exist between the interior of the containment and the environment. Hence, there are no direct leakage paths between the containment and the environment.

The seismological design bases for the reactor containment are described in Sections 2.5 and 15.2.4.

The containment structural design was in accordance with then current design practices for steel-lined, reinforced-concrete reactor containment structures. The design was based on accepted analytical methods and does not vary in any significant feature from structures then being

Revision 52Updated Online 09/30/20 SPS UFSAR 5.4-2 licensed or approved for construction. Rigid controls were maintained for all the materials, and the construction practices were as indicated in Chapter 15. Subatmospheric pressure operation results in no significant effect on the structural design.

The subatmospheric containment system does not depart in any significant way from good engineering design practices as used for atmospheric containments, yet it provides a substantial increase in public safety.

The following sections provide descriptions of the analytical models used to calculate the containment pressure and temperature responses to the design basis loss of coolant accident.

5.4.1 LOCA Mass and Energy Release Analysis 5.4.1.1 Purpose of Analysis The analysis documented in this section involves calculations of the long term Loss of Coolant Accident (LOCA) mass and energy releases for the double-ended pump suction guillotine (DEPSG) and double-ended hot leg guillotine (DEHLG) break cases with the proposed uprated conditions. This documentation provides the analytical basis with respect to the LOCA containment mass and energy release for the operation of the Surry Power Station Unit 1 and 2 at the described conditions.

Rupture of any of the piping carrying pressurized high temperature reactor coolant, termed a LOCA, will result in release of steam and water into the containment. This, in turn, will result in an increase in the containment pressure and temperature. The mass and energy release rates described in this section are used in further computations to evaluate the containment heat removal systems capability and containment structural integrity following a postulated loss of coolant accident. These analyses are performed to demonstrate compliance with General Design Criteria 38 and 50 of 10 CFR 50, Appendix A. Section 5.4.1.2 presents the long term mass and energy release analysis for containment pressurization evaluations. Section 5.4.1.3 presents the post-blowdown mass and energy releases for use in evaluation of recirculation spray pump available NPSH.

5.4.1.2 System Characteristics and Modeling Assumptions The mass and energy release analysis is sensitive to the assumed characteristics of various plant systems, in addition to other key modeling assumptions. Some of the most critical items are:

RCS initial conditions, core decay heat, safety injection flow, and metal and steam generator heat release modeling. Specific assumptions concerning each of these items are discussed below.

Tables 5.4-1 through 5.4-5 present key data assumed in the analysis.

For the long term mass and energy release calculations, operating temperatures which bound the highest full power average coolant temperature were used as initial conditions. A core rated power of 2587 MWt (adjusted for calorimetric error of +0.38% of power) was assumed. The use of higher temperatures is conservative because the initial fluid energy is based on coolant temperatures which are at the maximum levels attained in steady state operation. Additionally, an

Revision 52Updated Online 09/30/20 SPS UFSAR 5.4-3 allowance of +4.0°F is reflected in the temperatures in order to account for instrument error and deadband. The initial Reactor Coolant System (RCS) pressure in this analysis is based on a nominal value of 2250 psia. Also included is an allowance of +30 psi, which accounts for the measurement uncertainty on pressurizer pressure. The selection of 2250 psia as the limiting pressure is considered to affect the blowdown phase results only, since this represents the initial pressure of the RCS. The RCS rapidly depressurizes from this value until the point at which it equilibrates with containment pressure.

The rate at which the RCS blows down is initially more severe at the higher RCS pressure.

Additionally, the RCS has a higher fluid density at the higher pressure (assuming a constant temperature) and subsequently has a higher RC mass available for releases. Thus, 2280 psia initial pressure was selected as the limiting case for the long term mass and energy release calculations.

A 3% increase in the nominal RCS volume (which is composed of 1.6% allowance for thermal expansion and 1.4% for uncertainty) is also modeled. These assumptions conservatively maximize the mass and energy contained in the RCS.

The selection of the core model for the long term mass and energy calculation is based on the need to conservatively maximize the core stored energy. To maximize the core stored energy used in the analysis, an upper bound value is used which addresses the effect of uncertainties in the fuel temperature models and the material properties. The fuel design features and conditions used in the calculation of fuel temperatures for the core stored energy are selected to be bounding for the Surry reload cores. The core stored energy is calculated as a function of the number of feed assemblies, once-burned assemblies, and twice-burned assemblies in the reload core. To ensure conservatism in the core stored energy calculation, the feed fuel core stored energy is based on the maximum fuel temperatures, which typically occur at or near the beginning of cycle. The core stored energy for the once-burned fuel assemblies is assumed to be the maximum core stored energy over the assembly burnup range covering the lowest assembly burnup for the once-burned fuel to the licensed burnup limit. The core stored energy for the twice-burned fuel assemblies is assumed to be the maximum core stored energy over the assembly burnup range covering the lowest assembly burnup for the twice-burned fuel to the licensed burnup limit. The reload core stored energy is a weighted average of the maximum core stored energies calculated for the feed assemblies, the once-burned assemblies, and the twice burned assemblies.

Regarding safety injection flow, the mass and energy calculation considered configurations to conservatively bound potential alignments. The spectrum of cases included: Minimum SI -

Single Train (conservatively low ECCS flowrates); Maximum SI - Single Train (conservative early depletion of refueling water storage tank); and Maximum SI - Two Train (nominal configuration).

Revision 52Updated Online 09/30/20 SPS UFSAR 5.4-4 The following assumptions were employed to ensure that the mass and energy releases are conservatively calculated, thereby maximizing energy release to containment:

1. Maximum expected operating temperature of the Reactor Coolant System (100% power conditions).

2. An allowance in temperature for instrument error and dead band (+4.0°F).

3. Margin in RCS volume of 3% (which is composed of 1.6% allowance for thermal expansion and 1.4% for uncertainty).

4. Core rated power of 2587 MWt.

5. Allowance for calorimetric error (+0.38% power).

6. Conservative coefficient of heat transfer (i.e., steam generator primary/secondary heat transfer and Reactor Coolant System metal heat transfer).

7. Allowance in core stored energy for effect of fuel densification.

8. An upper bound calculation of core stored energy which addresses the effect of uncertainties in the fuel temperature models and the material properties.

9. An allowance for RCS initial pressure uncertainty (+30 psi).

10. A maximum containment backpressure equal to design pressure.

11. The steam generator (SG) metal mass was modeled to include only the portion of the SG which is in contact with the fluid on the secondary side (i.e., 474,000 lbm/SG out of a possible 691,000 lbm). Portions of the SGs such as the elliptical head, upper shell, and miscellaneous internals have poor heat transfer due to their location with respect to the secondary side water level. The energy stored in these areas available for release to the primary side break flow will not be able to effectively transfer energy to the RCS, thus the energy will be removed at a much slower rate and time period (i.e., >10,000 seconds).

12. A provision for modeling steam flow from the secondary side to the turbine through the turbine isolation valve was conservatively addressed with the control systems and setpoints in the modeling code. The valve isolation time was modeled as occurring at 1.5 seconds after the low pressurizer pressure reactor trip setpoint was reached.

13. As noted in Section 2.4 of Reference 2, the option to provide more specific modeling pertaining to decay heat has been exercised to specifically reflect the Surry Unit 1 and 2 core heat generation, while retaining the two sigma uncertainty to assure conservatism.

14. Steam generator tube plugging level (0% uniform).

• Maximizes reactor coolant volume and fluid release.

• Maximizes heat transfer area across the SG tubes.

• Reduces coolant loop resistance, which reduces p upstream of break and increases break flow.

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15. The main feedwater flow was modeled as a linear coastdown over a duration of ten seconds.

The coastdown was modeled as occurring after the low pressurizer pressure-SI trip setpoint was reached plus a signal processing time.

Use of the above conditions and assumptions result in a bounding analysis of the release of mass and energy from the RCS in the event of a LOCA. This analysis is applicable for operation of Surry Unit 1 and 2 at a core rated power of 2587 MWt.

5.4.1.3 Long Term LOCA Mass and Energy Release Analysis 5.4.1.3.1 Introduction The evaluation model used to the long term LOCA mass and energy release calculations was the March 1979 model described in Reference 2. This evaluation model has been reviewed and approved by the NRC, and has been used in the analysis of other dry containment plants.

These mass and energy releases are used in the containment response analysis described in Section 5.4.2.

5.4.1.3.2 LOCA Mass and Energy Release Phases The containment system receives mass and energy releases following a postulated rupture in the RCS. These releases continue over a time period, which, for the LOCA mass and energy analysis, is typically divided into four phases:

1. Blowdownthe period of time from accident initiation (when the reactor is at steady state operation) to the time that the RCS and containment reach an equilibrium state.

2. Refillthe period of time when the lower plenum is being filled by accumulator and ECCS water. At the end of blowdown, a large amount of water remains in the cold legs, downcomer and lower plenum. To conservatively consider the refill period for the purpose of containment mass and energy releases, it is assumed that this water is instantaneously transferred to the lower plenum along with sufficient accumulator water to completely fill the lower plenum. This allows an uninterrupted release of mass and energy to containment.

Thus, the refill period is conservatively neglected in the mass and energy release calculation.

3. Refloodbegins when the water from the lower plenum enters the core and ends when the core is completely quenched.

4. Post-reflood (GOTHIC)describes the period following the reflood transient. For the pump suction break, a two-phase mixture exits the core, passes through the hot legs and is superheated in the steam generators. After the broken loop steam generator cools, the break flow becomes two phase.

Computer Codes The Reference 2 mass and energy release evaluation model is comprised of mass and energy release versions of the following codes: SATAN VI and WREFLOOD. These codes were used to calculate the long term LOCA mass and energy releases through the end of reflood for the

Revision 52Updated Online 09/30/20 SPS UFSAR 5.4-6 Surry Power Station Unit 1 and 2. GOTHIC calculates the post reflood mass and energy releases in accordance with Topical Report DOM-NAF-3 (Reference 8).

SATAN calculates blowdown, the first portion of the thermal-hydraulic transient following break initiation, including pressure, enthalpy, density, mass and energy flowrates, and energy transfer between primary and secondary systems as a function of time.

The WREFLOOD code addresses the portion of the LOCA transient where the core reflooding phase occurs after the primary coolant system has depressurized (blowdown) due to the loss of water through the break and when water supplied by the Emergency Core Cooling refills the reactor vessel and provides cooling to the core. The most important feature is the steam/water mixing model. (See Section 5.4.1.3.5.2.)

GOTHIC models the post-reflood portion of the transient. The GOTHIC code is used for the transfer of decay heat and the stored energy in the primary and secondary systems to the containment.

5.4.1.3.3 Break Size and Location Generic studies have been performed with respect to the effect of postulated break size on the LOCA mass and energy releases. The double ended guillotine break has been found to be limiting due to larger mass flow rates during the blowdown phase of the transient. During the reflood and froth phases, the break size has little effect on the releases.

Three distinct locations in the reactor coolant system loop can be postulated for pipe rupture:

1. Hot leg (between vessel and steam generator)

2. Cold leg (between pump and vessel)

3. Pump suction (between steam generator and pump)

The break locations analyzed for this program are the double-ended pump suction guillotine, DEPSG (10.48 ft2) and the double-ended hot leg guillotine, DEHLG (9.18 ft2). Break mass and energy releases have been calculated for the blowdown, reflood and post-reflood phases of the LOCA for each case analyzed. The following information provides a discussion on each break location.

The DEHLG has been shown in previous studies to result in the highest blowdown mass and energy release rates. Although the core flooding rate would be the highest for this break location, the amount of energy released from the steam generator secondary is minimal because the majority of the fluid which exits the core bypasses the steam generators venting directly to containment. As a result, the reflood mass and energy releases are reduced significantly as compared to either the pump suction or cold leg break locations where the core exit mixture must pass through the steam generators before venting through the break. For the hot leg break, generic

Revision 52Updated Online 09/30/20 SPS UFSAR 5.4-7 studies have confirmed that there is not reflood peak (i.e., from the end of the blowdown period the containment pressure would continually decrease). The DEHLG reflood and post-reflood phase calculations are not required to determine peak containment pressure, but were calculated for use in the calculation for the recirculation spray pump available NPSH. Further details about the hot leg mass and energy analysis are contained in Section 5.4.1.4. The mass and energy releases for the hot leg break blowdown phase are included in the present section.

The cold leg break location has also been found in previous studies to be much less limiting in terms of the overall containment energy releases. The cold leg blowdown is faster than that of the pump suction break, and more mass is released into the containment. However, the core heat transfer is greatly reduced, and this results in a considerably lower energy release into containment. Studies have determined that the blowdown transient for the cold leg is, in general, less limiting than for the pump suction break. During reflood, the reflooding rate is greatly reduced and the energy release rate into the containment is reduced. Since the DEPSG case provides bounding results, the cold leg break location is not explicitly analyzed.

The pump suction break combines the effects of the relatively high core flooding rate, as in the hot leg break, and a break flow path through which the stored energy in the steam generators can be transferred to the containment. As a result, the pump suction break yields the highest energy flow rates during the post-blowdown period since all of the Reactor Coolant System available energy contributes to the calculated mass and energy releases.

5.4.1.3.4 Assessment of Single Failure Effects An analysis of the effects from various single failures has been performed on the mass and energy release rates for each break analyzed. An inherent assumption in the generation of the mass and energy release is that offsite power is lost. This results in the actuation of the emergency diesel generators, required to power the safety injection system. This is not an issue for the blowdown period which is limited by the DEHLG break.

Three cases have been analyzed for the effects of a single failure. In the case of minimum safeguards, the single failure postulated to occur is the loss of an emergency diesel generator. This results in the loss of one pumped safety injection train. Two variations on the minimum safeguards scenario were addressed. The first case was a maximum safety injection (SI) flow, single train case. This case will result in low flow rates, but an early refueling water storage tank depletion. The second configuration is the minimum SI flow, single train case. As compared to the first case the SI flow would be minimized, although the time of RWST depletion would be later.

Sensitivities indicate that containment depressurization time is more limiting for the minimum SI-single train case. Containment depressurization peak pressure and LHSI pump NPSH are more limiting for the maximum SI-single train scenario. For the case of maximum safeguards, no failure is postulated to occur. Sensitivity cases using mass and energy data for the pump suction, maximum safeguards case indicated that this configuration is not limiting for any analysis acceptance criteria. Therefore, no detailed data or containment analysis results are presented for

Revision 52Updated Online 09/30/20 SPS UFSAR 5.4-8 this case. The analysis of the cases described ensures that the effect of all credible single failures is bounded.

5.4.1.3.5 Mass and Energy Release Data 5.4.1.3.5.1 Blowdown Mass and Energy Release Data A version of the SATAN-VI code, which is the code used for the Emergency Core Cooling System (ECCS) calculation in Reference 3 is used for computing the blowdown transient. The code utilizes the control volume (element) approach with the capability for modeling a large variety of thermal fluid system configurations. The fluid properties are considered uniform and thermodynamic equilibrium is assumed in each element. A point kinetic model is used with weighted feedback effects. The major feedback effects include moderator density, moderator temperature and Doppler broadening. A critical flow calculation for subcooled (modified Zaloudek), two-phase (Moody) or superheated break flow is incorporated into the analysis. The methodology for the use of this model is described in Reference 2.

5.4.1.3.5.2 Reflood Mass and Energy Release Data The WREFLOOD code used for computing the reflood transient is a modified version of that used in the 1981 ECCS evaluation model (Reference 3).

The WREFLOOD code consists of two basic hydraulic modelsone for the contents of the reactor vessel, and one for the coolant loops. The two models are coupled through the interchange of the boundary conditions applied at the vessel outlet nozzles and at the top of the downcomer.

Additional transient phenomena, such as pumped safety injection and accumulators, reactor coolant pump performance and steam generator release, are included as auxiliary equations which interact with the basic models as required. The WREFLOOD code permits the capability to calculate variations during the core reflooding transient of basic parameters, such as core flooding rate, core and downcomer water levels, fluid thermodynamic conditions (pressure, enthalpy, density) throughout the primary system, and mass flow rates through the primary system. The code permits hydraulic modeling of the two flow paths available for discharging steam and entrained water from the core to the break, i.e., the path through the broken loop and the path through the unbroken loops.

A complete thermal equilibrium mixing condition for the steam and emergency core cooling injection water during the reflood phase has been assumed for each loop receiving ECCS water. This is consistent with the usage and application of the Reference 2 mass and energy release evaluation model in recent analyses, e.g., D. C. Cook Docket (Reference 4). Even though the Reference 2 model credits steam/mixing only in the intact loop and not in the broken loop, justification, applicability and NRC approval for using the mixing model in the broken loop has been documented (Reference 4). This assumption is justified and supported by test data, and is summarized as follows:

Revision 52Updated Online 09/30/20 SPS UFSAR 5.4-9 The model assumed a complete mixing condition (i.e., thermal equilibrium) for the steam/water interaction. The complete mixing process, however, is made up of two distinct physical processes. The first is a two phase interaction with condensation of steam by cold ECCS water. The second is a single phase mixing of condensate and ECCS water. Since the steam release is the most important influence to the containment pressure transient, the steam condensation part of the mixing process is the only part that need be considered. (Any spillage directly heats only the sump.)

The most applicable steam/water mixing test data has been reviewed for validation of the containment integrity reflood steam/water mixing model. This data is that generated in 1/3 scale tests (Reference 5), which are the largest scale data available and thus most clearly simulates the flow regimes and gravitational effects that would occur in a PWR. These tests were designed specifically to study the steam/water interaction for PWR reflood conditions.

From the entire series of 1/3 scale tests, a group corresponds almost directly to containment integrity reflood conditions. The injection flowrates for this group cover all phases and mixing conditions calculated during the reflood transient. The data from these tests were reviewed and discussed in detail in Reference 2. For all of these tests, the data clearly indicate the occurrence of very effective mixing with rapid steam condensation. The mixing model used in the containment integrity reflood calculation is therefore wholly supported by the 1/3 scale steam/water mixing data.

Additionally, the following justification is noted. The post-blowdown limiting break for the containment integrity peak pressure analysis is the double-ended pump suction guillotine. For this break, there are two flowpaths available in the RCS by which mass and energy may be released to containment. One is through the outlet of the steam generator, the other via reverse flow through the reactor coolant pump. Steam which is not condensed by ECCS injection in the intact RCS loops passes around the downcomer and through the broken loop cold leg and pump in venting to containment. This steam also encounters ECCS injection water as it passes through the broken loop cold leg, complete mixing occurs and a portion of it is condensed. It is this portion of steam which is condensed that is taken credit for in this analysis. This assumption is justified based upon the postulated break location, and the actual physical presence of the ECCS injection nozzle. A description of the test and test results is contained in References 2 and 5.

The methodology previously discussed in Reference 2 has been utilized and approved on the Dockets for numerous dry containment plants such as Beaver Valley Unit 2, Millstone Unit 3 and Indian Point Unit 2.

The blowdown and reflood mass and energy release data (including the transients of principal parameters during reflood) are provided in Section 5.4.2 as composite tables of data used in the containment response analysis. Section 5.4.2 describes the usage of these data in the GOTHIC code.

Revision 52Updated Online 09/30/20 SPS UFSAR 5.4-10 5.4.1.3.5.3 Post-Reflood Mass and Energy Release Data The GOTHIC code (Reference 8) is used for computing the post-reflood transient.

GOTHIC calculates the transfer of decay heat and the stored energy in the primary and secondary systems to the containment. The mass and energy releases that occur during this phase are typically superheated due to the depressurization and equilibration of the broken loop and intact loop steam generators. During this phase of the transient, the RCS has equilibrated with the containment pressure, but the steam generators contain a secondary inventory at an enthalpy that is much higher than the primary side. Therefore, there is a significant amount of reverse heat transfer that occurs. Steam is produced in the core due to core decay heat. For a pump suction break, a two phase fluid exits the core, flows through the hot legs and becomes superheated as it passes through the steam generator. Once the broken loop cools, the break flow becomes two phase.

The mass and energy release rates calculated by GOTHIC are processed as described in Section 5.4.2.1 for use in the containment response analysis.

5.4.1.3.5.4 Decay Heat Model As part of the Surry Core Uprating effort a detailed DEHLG mass and energy release analysis (Section 5.4.1.4) was completed for use in the evaluation of recirculation spray pump available NPSH. The 1975 mass and energy release evaluation model (Reference 6) was used for this calculation. The decay heat standard available and incorporated into the Reference 6 evaluation model was adopted by the ANS Standards Subcommittee in October 1971.

The NRC staff Safety Evaluation Report (SER) for the March 1979 evaluation model approved use of the November 1979 ANS Standard-5.1 decay heat model for the calculation of mass and energy releases to the containment following a loss-of-coolant accident. Therefore, to more realistically model the RCS, the Reference 7 decay heat model was utilized for this core uprating effort in conjunction with the 1975 evaluation model. The Reference 7 decay heat model is utilized in the GOTHIC containment analysis. This standard was used in the mass and energy release model with the following input specific for the Surry Power Station Unit 1 and 2. The primary assumptions which make this calculation specific for the Surry Power Station are the enrichment factor, minimum/maximum number of the new fuel assemblies per cycle and fuel cycle length. A conservative lower bound for enrichment of 3% was used. Table 5.4-2 lists the decay heat values used in this analysis.

Significant assumptions in the generation of the decay heat values:

1. Decay heat sources considered are fission product decay and heavy element decay of U-239 and Np-239.

2. Decay heat power from fissioning isotopes other than U-235 is assumed to be identical to that of U-235.

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3. Fission rate is constant over the operating history of maximum power level.

4. The factor accounting for neutron capture in fission products has been taken from Equation 11 of Reference 7 up to 10,000 seconds, and Table 10 of Reference 7 beyond 10,000 seconds.

5. The fuel has been assumed to be at full power for 108 seconds.

6. The number of atoms of U-239 produced per second has been assumed to be equal to 70% of the fission rate.

7. The total recoverable energy associated with one fission has been assumed to be 200 MeV/fission.

8. Two sigma uncertainty (two times the standard deviation) has been applied to the fission product decay.

9. End of cycle core average burnup that is less than or equal to 40,000 MWD/MTU.

10. Core fresh fuel loading that is greater than or equal to 72.5 MTU.

11. Core average fuel enrichment that is greater than or equal to 3.0%.

5.4.1.3.6 Sources of Mass and Energy The sources of mass considered in the LOCA mass and energy release analysis are the reactor coolant system, accumulators and pumped safety injection.

The energy inventories considered in the LOCA mass and energy release analysis include:

1. Reactor Coolant System Water

2. Accumulator Water

3. Pumped Injection Water

4. Decay Heat

5. Core Stored Energy

6. Primary Metal

7. Steam Generator Metal (includes transition cone, shell, wrapper, and other internals)

8. Steam Generator Secondary Energy (includes fluid mass and steam mass)

9. Secondary Transfer of Energy (feedwater into and steam out of the steam generator secondary)

10. SG tubes In the mass and energy release data presented, no Zirc-water reaction heat was considered because the clad temperature did not rise high enough for the rate of the Zirc-water reaction heat to be of any significance.

Revision 52Updated Online 09/30/20 SPS UFSAR 5.4-12 The consideration of the various energy sources in the mass and energy release analysis provides assurance that all available sources of energy have been included in this analysis. Thus the review guidelines presented in Standard Review Plan Section 6.2.1.3 have been satisfied.

5.4.1.4 Mass and Energy Releases for Available NPSH Analysis (Hot Leg Double Ended Rupture, Post-Blowdown)

In support of the evaluation of recirculation spray pump and low head safety injection pump available NPSH, a LOCA long term mass and energy release analysis was completed. Mass and energy releases for use in the DEHLG evaluation, Maximum Safety injection - two train case are provided in Table 5.4-6. The DEHLG data is presented for two break paths. Break path 1 represents the mass and energy exiting from the reactor vessel side of the break. Break path 2 is the mass and energy exiting from the SG side of the break.

The large break LOCA mass and energy releases were generated using the evaluation models described in References 2 and 6. The blowdown phase mass and energy releases were calculated using the Reference 2 evaluation model, as described in Section 5.4.1.3.5.1 and provided in Table 5.4-6. The large break LOCA mass and energy releases for the reflood phase was generated using the 1975 mass and energy release evaluation model (Reference 6).

Table 5.4-6 provides the hot leg mass and energy instantaneous releases, plus reflood mass and energy data for the reflood phase. The Reference 6 mass and energy release evaluation model was utilized because of its capability to calculate reflood phase transient mass and energy release data.

The focus of the Reference 2 evaluation model is for the pressure and temperature response of containment. As noted in Section 5.4.1.3.3, generic studies confirm that for the hot leg break, there is no reflood peak, therefore the reflood code applicability of the Reference 2 model was not pursued. The Reference 6 evaluation model still remains a valid analytical tool that has been reviewed and approved by the NRC, although it does not exhibit the benefits of the improved model, i.e., steam water mixing model during reflood. Please note the reflood phase modeling of the Reference 6 evaluation model has been enhanced to incorporate the 1979 decay model as described in Section 5.4.1.3.5.4 for this Surry core uprating program. The DEHLG mass and energy releases in Table 5.4-6 are based on the initial conditions consistent with the design basis analysis of Section 5.4.1.3.

The analysis performed to calculate long term mass and energy releases following a postulated DEHLG is similar to the analysis described in Section 5.4.1.3. The transient is divided into four phases: blowdown, refill, reflood and post-reflood. The characteristics of the phases are also similar except for the hot leg break, where the amount of energy released from the SG is minimal because the majority of the fluid which exits the core bypasses the SG venting directly to containment.

The analysis as noted above utilized both the Reference 2 and 6 mass and energy release evaluation models. The computer models used were comprised of mass and energy release versions of the following codes: SATAN VI model (Reference 2) for blowdown, WREFLOOD model (Reference 6) for reflood phase, and GOTHIC (Reference 8) for the post-reflood phase.

Revision 52Updated Online 09/30/20 SPS UFSAR 5.4-13 These codes were used to calculate the long term LOCA mass and energy releases for the hot leg break. The blowdown releases are calculated with the same version of the SATAN code described in Section 5.4.1.3.2.

The WREFLOOD code addresses the portion of the LOCA transient where the core reflooding phase occurs after the primary coolant system has depressurized (blowdown) due to the loss of water through the break and when water supplied by the Emergency Core Cooling System refills the reactor vessel and provides cooling to the core. The WREFLOOD version of the Reference 6 model does not include the enhanced steam/water mixing model of Reference 2.

5.4.2 LOCA Containment Pressure and Temperature Response The containment pressure and temperature response is analyzed for the primary system breaks which are discussed in Section 5.4.1. Various single failures of the engineered safety features are analyzed to identify the limiting single failures.

There is one pressure peak following a Reactor Coolant System (RCS) hot leg or cold leg rupture. This pressure peak occurs near the end of the initial blowdown of the RCS after a double ended guillotine (DEG) of either a hot or cold leg. This will be referred to as the blowdown peak pressure. Its magnitude is a function of the following parameters:

1. The containment free volume.

2. The mass of air inside the containment structure (a function of initial pressure and temperature).

3. The amount of energy flow out of the break during the initial blowdown of the RCS.

4. The rate of heat removal from the containment atmosphere by the passive heat sinks within the containment structure.

A DEHLG produces the largest blowdown peak pressure. This event releases the most energy to the containment atmosphere during the initial blowdown since the hot leg pipe size is larger than that of a RCS pump discharge and there is no resistance to flow due to a RCS pump as is the case with a DEPSG. The magnitude of the blowdown peak pressure is independent of the active engineered safety feature (ESF) because ESF does not become effective until after the peak pressure occurs. However, the accumulators do have a small effect on the blowdown peak pressure.

Following the core reflooding period, the containment depressurization systems and containment passive heat sinks remove energy from the containment atmosphere at a rate sufficient to reduce the pressure to below atmospheric pressure in less than 60 minutes. The depressurization time is a function of the following parameters:

1. The containment free volume.

2. The mass of air inside the containment structure.

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3. The rate of heat transfer between the containment atmosphere and the passive heat sinks within the containment structure.

4. The rate of heat removal from the containment atmosphere by the containment heat removal systems (this is significantly dependent on the ultimate heat sink temperature).

5. The rate of mass and energy release to the containment from the break following the end of core reflooding.

6. The mass of nitrogen added to the containment from the SI accumulators.

After the containment is depressurized, the depressurization systems continue to remove energy from the containment at a rate sufficient to maintain the containment at subatmospheric pressure. The heated passive heat sinks add energy back to the containment atmosphere following depressurization. The containment experiences a pressure peak less than 1.0 psig after the termination of containment spray associated with emptying the RWST.

5.4.2.1 Containment Response Analytical Model The GOTHIC computer program which is used to model the containment system, the passive heat sinks, and the containment heat removal systems, was developed for the Electric Power Research Institute (EPRI) by Numerical Applications, Inc. A topical report (DOM-NAF-3) described in detail the assumptions used and the mathematical formulations employed. The use of GOTHIC for containment analysis has been approved by the NRC as documented in DOM-NAF-3-0.0-P-A (Reference 8).

GOTHIC solves the conservation equations for mass, momentum, and energy for multi-component, multi-phase flow in lumped parameter and/or multi-dimensional geometries.

The phase balance equations are coupled by mechanistic models for interface mass, energy and momentum transfer that cover the entire flow regime from bubbly flow to film/drop flow, as well as single phase flows. The interface models allow for the possibility of thermal non-equilibrium between phases and unequal phase velocities, including countercurrent flow. GOTHIC includes full treatment of the momentum transport terms in multidimensional models, with optional models for turbulent shear and turbulent mass and energy diffusion. Other phenomena include models for commonly available safety equipment, heat transfer to structures, hydrogen burn and isotope transport.

5.4.2.1.1 Passive Heat Sinks Thermal conductors are the primary heat sink for the blowdown energy. The conductors can be made up of any number of layers of different materials. One-dimensional conduction solutions are used to be consistent with the lumped modeling approach.

The thermal conductor is divided into regions, one for each material layer, with an appropriate thickness and material property for each region. GOTHIC accepts inputs for material density, thermal conductivity and specific heat. These values are obtained from published

Revision 52Updated Online 09/30/20 SPS UFSAR 5.4-15 literature for the materials present in each conductor. Conductors with high heat flux at the surface and low thermal conductivity must have closely spaced nodes near the surface to adequately track the steep temperature profile. The node spacing is set so the node Biot number for each node is less than 0.1. The Biot number is the ratio of external to internal conductance.

It is not practical or necessary to model each individual piece of equipment or structure in the containment with a separate conductor. Smaller conductors of similar material composition can be combined into a single effective conductor. In this combination, the total mass and the total exposed surface area of the conductors is preserved. The thickness controls the response time for the conductors and is of secondary importance. The conductors are grouped by thickness and material type. The effective thickness for a group of wall conductors is calculated by the equation below. The heat sink material types, surface areas, and thickness are derived based on plant-specific inventories. Concrete, carbon steel, and stainless steel are the most common materials.

igroup ti Ai t eff = ----------------------------

Ai igroup If there is a small air gap or a contact resistance between the containment liner and the concrete, it is modeled as a separate material layer at the nominal gap thickness with applicable material properties. This overestimates the contact resistance because convection and radiation effects will be ignored. A maximum gap conductance of 40 Btu/hr-ft2-F is used. The gap width is determined by dividing the gap thermal conductivity by the gap conductance.

All containment passive heat sinks are included in the lumped containment volume. The primary system metal and SG secondary shells are included in the simplified RCS model that is used for the calculation of long-term mass and energy release; however, these conductors are not used for condensation or convection heat transfer with the containment atmosphere.

5.4.2.1.2 Conductor Surface Heat Transfer The Direct heat transfer option with the DLM (Diffusion Layer Model) condensation option is used for all containment passive heat sinks except the sump floor. With the Direct option, all condensate goes directly to the liquid pool at the bottom of the volume. The effects of the condensate film on the heat and mass transfer are incorporated in the formulation of the DLM option. Under the DLM option, the condensation rate is calculated using a heat and mass transfer analogy to account for the presence of non-condensing gases.

For a conductor representing the containment floor or sump walls that will eventually be covered with water from the break and condensate, the Split heat transfer option is used to switch the heat transfer from the vapor phase to the liquid phase as the liquid level in the containment

Revision 52Updated Online 09/30/20 SPS UFSAR 5.4-16 builds. A quicker transition to liquid heat transfer is more conservative for containment analysis.

The Split option is used with lmax , the maximum liquid fraction, set to d-l max = ---

H where d is the transition water depth and H is the volume height. A reasonable value for d of 0.1 inch switches the heat transfer from the vapor phase to the liquid phase as the liquid level in the containment reaches 0.1 inch. Other values may be appropriate depending on the geometry of the floor and sump.

For conductors with both sides exposed to the containment, the Direct option is applied to both sides. Alternatively, if the conductor is symmetric about the centerplane, a half-thickness conductor can be used with the total surface area of the two sides and an insulated back side heat transfer option. The conductor face that is not exposed to the atmosphere is assumed insulated.

The Specified Heat Flux option is used with the nominal heat flux set to zero.

Containment walls above grade and the containment dome have a specified external temperature boundary condition with a heat transfer coefficient of 2.0 Btu/hr-ft2-F to model convective heat transfer to the outside atmosphere. The GOTHIC heat transfer solution scheme allows for accurate initialization of the temperature distribution in the containment wall and dome prior to the transient initiation.

A conservative containment liner response is obtained by adding a small conductor that has the same construction and properties as the liner conductor. A conductor surface area of 1 ft2 is used to minimize impact on the lumped containment pressure and temperature response. The inside heat transfer option is the same as used for the actual liner conductor (Direct with DLM) with a multiplier of 1.2 for conservatism.

5.4.2.1.3 Spray Modeling GOTHIC includes models that calculate the sensible heat transfer between the drops and the vapor and the evaporation or condensation at the drop surface. The efficiencythe actual temperature rise over the difference between the vapor temperature and the drop inlet temperaturecannot be directly specified in GOTHIC. The efficiency is primarily a function of the drop diameter. The GOTHIC models account for the effect of the diameter through the Reynolds number dependent fall velocity and heat transfer coefficients. A heat and mass transfer analogy is used to calculate the effective mass transfer coefficient, which is used to calculate the evaporation or condensation. Containment spray is modeled as described in DOM-NAF-3-0.0-P-A.

5.4.2.1.4 Containment Heat Removal Heat exchangers that remove energy from the containment sump are modeled with the available heat exchanger options in GOTHIC. Use of a GOTHIC heat exchanger option

Revision 52Updated Online 09/30/20 SPS UFSAR 5.4-17 dynamically couples the heat exchanger performance to the predicted primary and secondary fluid conditions. This can provide a small benefit compared to other codes (e.g., LOCTIC) that use bounding UA values to cover the fluid conditions predicted over the entire transient.

The GOTHIC heat exchanger type that closely matches the actual heat exchanger is selected. The inside and outside heat transfer areas are calculated from the heat exchanger geometry details. For tube and shell arrangements, the shell side flow area is set to the open area across the tubes at the mid-plane of the heat exchanger and the shell side hydraulic diameter is set to the tube outer diameter. The GOTHIC option for built-in heat transfer coefficients is used to determine heat transfer coefficients that depend on the primary and secondary side Reynolds and Prandtl numbers. The heat exchanger models in GOTHIC are for basic heat exchanger designs and may not account for the details of a particular heat exchanger (e.g., baffling in a tube-and-shell heat exchanger). A forcing function can be used on the primary and secondary side heat transfer coefficients to tune the heat exchanger performance to manufacturer or measured specifications. Alternatively, the heat transfer area can be adjusted to match the specified performance. Fouling factors and tube plugging are applied when conservative.

5.4.2.1.5 LOCA Mass and Energy Release to Containment During a LOCA event, most of the vessel water will be displaced by the steam generated by flashing. The vessel is then refilled by the accumulators and the high and low pressure injection systems. GOTHIC is not suitable for modeling the refill period because it involves quenching of the fuel rods where film boiling conditions may exist. Current versions of GOTHIC do not have models for quenching and film boiling. Therefore, for the blowdown, refill and reflood stages, the mass and energy release rates are obtained from Westinghouse LOCA analysis. The Westinghouse release data includes the water from the ECCS accumulators, but the nitrogen release to containment is modeled separately in GOTHIC.

The LOCA mass and energy release rates are input to GOTHIC for the blowdown and reflood periods of the design basis LOCAs. The calculation of these release rates is described in Section 5.4.1. The mass and energy release rates used in the containment peak pressure, containment depressurization, and NPSH analyses for the RS and LSHI pumps are provided in this section. The mass and energy release rates for the DEHLG through the end of reflood are tabulated in Table 5.4-6 for maximum two-train safety injection flow. The mass and energy release rates for the reactor coolant DEPSG through the end of reflood are provided in Table 5.4-7 for maximum single-train safety injection flow.

At the end of reflood, the core has been recovered with water and the ECCS continues to supply water to the vessel. Residual stored energy and decay heat comes from the fuel rods.

Stored energy in the vessel and primary system metal will also be gradually released to the injection water and released to the containment via steaming through the core or spillage into the containment sump. In addition, there may be some buoyancy-driven circulation through the intact steam generator loops that will remove stored energy from the steam generator metal and the

Revision 52Updated Online 09/30/20 SPS UFSAR 5.4-18 water on the secondary side. Depending on the location of the break, the two-phase mixture in the vessel may pass through the steam generator on the broken loop and acquire heat from the stored energy in the secondary system. For these conditions, GOTHIC is capable of calculating the mass and energy release from the break into containment.

The GOTHIC long-term mass and energy release accounts for the transfer of the decay heat and the stored energy in the primary and secondary systems to the containment after the end of reflood. The energy for each source term is acquired at the end of reflood from the Westinghouse mass and energy release analysis. The rate of energy release is determined by a simplified GOTHIC RCS model that is coupled to the containment volume. Thus, the flow from the vessel to the containment is dependent on the GOTHIC-calculated containment pressure.

Lumped volumes are used for the vessel, downcomer, cold legs, steam generator secondary side, up-flow steam generator tubes and down-flow steam generator tubes. Separate sets of loop and secondary system volumes are used for the intact and broken loops with the connections between the broken loop and containment as necessary for the modeled break location. The Westinghouse calculated mass and energy inventory at the end of reflood establishes the liquid volume fractions and the fluid temperatures in the primary and secondary systems.

The primary and secondary system geometries, including primary system resistances, are consistent with the models used for non-LOCA accident analyses. In order to predict the natural circulation through the intact loops and the correct water level in the vessel and downcomer, the volumes are modeled with the correct elevations and heights. The vessel height may be adjusted so that the water and steam inventory at the end of reflood matches the vendors boundary conditions, but this correction does not affect the hydraulic analysis.

Safety injection fluid is added to the downcomer volume (for the intact cold legs) and the broken loop cold leg. In both locations, the SI fluid mixes with the resident fluid and any vapor from the intact SGs. The SI flow is taken from the RWST until a low-low level is reached, at which time the SI fluid is taken from the containment sump.

A thermal conductor is used to model the transfer of energy stored in the shell side of the steam generator to the SG secondary fluid. The initial temperature is set to match the available stored energy specified at the end of reflood by the fuel vendor analysis. The up flow and down flow tubes on the steam generators are modeled separately with thermal conductors. This allows for the possibility of boiling in the up flow tubes and superheating of the steam in the down flow tubes. The heat transfer from the secondary side to the primary side is modeled using conductors with the inside connected to the primary system tube volumes. The Film heat transfer option is used on both sides of the tube. This option automatically accounts for heat transfer to the liquid or vapor phase as appropriate and includes boiling heat transfer modes.

Revision 52Updated Online 09/30/20 SPS UFSAR 5.4-19 5.4.2.1.6 NPSH Available NPSHa (net positive suction head available) is the difference between the fluid stagnation pressure and the saturation pressure at the pump intake. To calculate NPSHa for a given pump, the GOTHIC containment model includes a separate small volume for the pump suction. The volume elevation and height are set so that the mid-elevation of the volume is at the elevation of the pump first-stage impeller centerline. The volume pressure (with some adjustments for sump depth) can then be used in the NPSHa calculation. The temperature in the suction volume provides the saturation pressure. The junction representing piping between the sump and the suction volume reflects the friction and form pressure drop between the sump and the pump suction. The pump suction volume also allows accurate modeling of the mixing of cold water that is injected into the suction of the RS pumps.

The single volume GOTHIC model does not account for geometry details of the sump or the liquid that is held up in other parts of the containment. GOTHIC does calculate the total amount of liquid in the containment. A correlation is used to define the sump depth or liquid level as a function of the water volume in the containment. The correlation accounts for the sump geometry variation with water depth and accounts for the holdup of water in other parts of the containment.

Worst case conditions for NPSHa depend on the time that the pumps take suction from the sump. Therefore, the parameter settings that minimize NPSHa may vary depending on the timing for the operation of the pumps. In general, settings that reduce containment pressure and increase the sump water temperature reduce the NPSHa.

The water in the sump comes from three sources: direct deposit of mass from the break, condensate from the conductors, and spray drops. The drops from the blowdown will be very small and at the saturation temperature at the containment steam partial pressure when they enter the sump. After the blowdown, the spillage water from the vessel is directly put in the sump with no heat transfer to the atmosphere or walls and equipment in the containment. This is a conservative approach for NPSH analysis. The condensate is generated at the saturation temperature at the steam partial pressure and added directly to the sump. The heat transfer between the conductors and the condensate on the way to the sump is conservatively neglected.

Heat and mass transfer at the sump surface is allowed. GOTHICs model for heat and mass transfer at a pool is in good agreement with experimental data. For NPSH analysis, the liquid temperature is greater than the vapor temperature for most of the event, so a minimum pool area is specified to minimize evaporation. With this overall approach, the predicted sump temperature is conservatively high for the duration of the simulation.

Revision 52Updated Online 09/30/20 SPS UFSAR 5.4-20 The following adjustments are made to ensure a conservative calculation of NPSHa:

1. The heat and mass transfer to the containment heat sinks are expected to be under-predicted using the Direct heat transfer model. This is non-conservative for NPSH analysis. A multiplier of 1.2 applied to the heat transfer coefficient was shown to provide adequate conservatism in the calculation.

2. All of the spray water is injected as droplets into the containment atmosphere (nozzle spray flow fraction of 1). Analyses are performed using the largest Sauter droplet size. A confirmatory analysis is performed by reducing the Sauter diameter by 2, which sufficiently covers code and spray performance uncertainty (i.e., variation in nozzle design and orientation, nozzle flow rate and different header elevations) without creating drops too small that may cause excess droplet holdup in the atmosphere. NPSH analyses are relatively insensitive over this range of droplet size, and the two cases together confirm that the effect of sprays on reducing containment pressure is maximized. The minimum NPSHa is reported from the case that provides the smaller NPSHa.

3. A conservative water holdup volume is subtracted from the containment liquid volume to reduce the sump water height.

4. The upper limit on containment free volume is used.

5. The minimum containment air pressure is used.

6. Conservative assumptions for spray and other system parameters are used in accordance with plant-specific sensitivity studies.

A modification of the NPSH methodology used for developing component design inputs was submitted to the NRC in Reference 10. This alternate methodology can be used for NPSH and LOCA analyses that develop design inputs for component design, such as determination of margin for sump strainer design. The NRC approved the alternate methodology in Reference 11, thus confirming that it can be used for the intended application stated in Reference 10.

5.4.2.1.7 LOCA - Containment Pressure and Temperature Results The containment LOCA analysis is performed for the two limiting pipe break locations (DEPSG and DEHLG) discussed in Section 5.4.1. The DEPSG is most limiting for long-term containment temperature and pressure response. Table 5.4-7 provides the DEPSG mass and energy instantaneous releases as well as the releases for the reflood phase. The DEPSG data is presented for two break paths. Break path 1 represents the mass and energy exiting from the SG side of the break. Break path 2 is the mass and energy exiting from the pump side of the break.

Containment analysis parameters are listed in Table 5.4-17. The RS pumps start with individual delay times on 60% RWST level coincident with a CLS High High containment pressure signal.

Revision 52Updated Online 09/30/20 SPS UFSAR 5.4-21 The results of the containment pressure analysis are tabulated in Table 5.4-11. The initial containment conditions which yield the highest peak calculated containment pressure are the maximum pressure, temperature, and relative humidity, and are given in Table 5.4-10. The containment pressure and temperature transients for the hot leg double-ended guillotine are given on Figures 5.4-1 and 5.4-2, respectively.

The maximum peak containment pressure occurs after a DEHLG. As shown in Table 5.4-11, the calculated containment pressure is below the containment design pressure of 45 psig. The DEHLG is the design basis accident (DBA) for the containment structure (containment integrity DBA).

A single failure analysis is not necessary for the peak containment pressure evaluation since the peak pressure for each break case analyzed occurs early in the transient before any of the engineered safety feature (ESF) systems start.

The results of the containment depressurization analysis are tabulated in Table 5.4-12. Only a DEPSG is considered for the containment depressurization analysis since, as described earlier, this break produces the highest energy flow rates during the post-blowdown period. The containment pressure is less than 1.0 psig within one hour and less than 0 psig within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> as shown in Table 5.4-12. The SI flow is based on a minimum estimate. This minimizes the credit for steam condensation due to steam/water mixing.

The initial conditions which result in the maximum depressurization time are as follows:

1. Initial containment pressure of 12.52 psia.

2. Initial containment temperature of 125°F.

3. Initial containment relative humidity of 100%.

4. Service water (ultimate heat sink) temperature of 100°F.

5. Refueling water storage tank temperature of 45°F.

The initial conditions which result in the maximum time to approach the four hour subatmospheric requirement are as follows:

1. Initial containment pressure of 10.97 psia

2. Initial containment temperature of 75°F

3. Initial containment humidity of 100%

4. Service water (ultimate heat sink) temperature of 100°F

5. Refueling water storage tank temperature of 45°F These limiting values are consistent with the Technical Specifications. Instrumentation uncertainties for these parameters have been included in the safety analysis.

Revision 52Updated Online 09/30/20 SPS UFSAR 5.4-22 The highest depressurization peak pressure may result from service water temperatures less than the TS maximum.

A chronology of events for this DEPSG with minimum ESF for both sets of initial conditions described in this section is given in Table 5.4-13.

Representative results for containment pressure, vapor temperature, sump water temperature and RS heat exchanger duty are shown in Figures 5.4-3, 5.4-4 (illustrates containment vapor and liquid temperature) and 5.4-5, respectively. These results are based on the initial conditions which result in the maximum time to approach the four hour subatmospheric requirement as described above.

For the depressurization analysis, only the diesel generator failure is considered since all other single failures result in increased containment heat removal capability as compared to this single failure.

The results of the LOCA analysis are reported in Tables 5.4-11 and 5.4-12.

5.4.3 MSLB Containment Pressure and Temperature Response Surry did not previously have an explicit MSLB containment response analysis. The containment response was bounded by the Beaver Valley Unit 1 MSLB analysis. Reference 9 made the comparison and described the conservatism in the Beaver Valley Unit 1 MSLB containment pressurization analysis versus Surry. MSLB analysis has been performed for Surry using GOTHIC. The GOTHIC model is as described in Section 5.4 for LOCA analysis with the exception of the mass and energy release data. GOTHIC analysis inputs are provided in Table 5.4-17.

5.4.3.1 MSLB Mass and Energy Release to Containment For MSLB, the mass and energy release data is obtained from Westinghouse using NRC-approved methods. Surry does not have explicit mass and energy release data from Westinghouse. The North Anna MSLB mass and energy release data from Westinghouse was confirmed to be conservative for Surry and was applied for this analysis. The break junction uses 100-micron droplets for entrained liquid release. A range of break sizes from small split breaks to the largest double-ended break size is analyzed over the range of 0% to 114.3% of rated thermal power of 2587 MWt. Analysis of this range ensures that the most conservative results are predicted for containment pressure and temperature.

5.4.3.2 MSLB Pressure and Temperature Analysis 5.4.3.2.1 MSLB Peak Pressure Analysis The maximum containment peak pressure occurs for the 1.4 ft2 break at 0% power because it has the highest SG liquid mass and results in the largest mass release to the containment before the faulted SG dries out. Table 5.4-14 shows the results in peak pressure are less than the design limit of 59.7 psia. Table 5.4-15 shows the time sequence of events for the case with the proposed

Revision 52Updated Online 09/30/20 SPS UFSAR 5.4-23 TS air partial pressure limit of 11.3 psia. The atmosphere remains superheated for a very short time, returning to saturation within 10 seconds from the time of the break. The containment temperature and pressure peaks occur about 20 seconds before SG dryout, when condensation and the CS system overcome the steam release rate. Containment pressure drops rapidly once operator action terminates AFW to the faulted SG at 30 minutes, which stops the steam release to the containment.

SPS has cavitating venturis in the AFW lines leading to each SG that limit the flow rate to about 350 gpm. For the MSLB analyses, the mass release is 400 gpm after the faulted SG reaches dryout. This assumption provides a conservative, but reasonable long-term containment pressure and temperature response for SPS but does not affect the containment peak pressure and temperature, which occur earlier in the event.

The maximum initial air partial pressure is independent of SW temperature, because the RS system is not assumed to operate. Therefore, the maximum allowable TS air partial pressure is a constant line until the containment depressurization analyses limit the curve. In summary, a maximum operating containment air partial pressure of 11.3 psia ensures that the MSLB peak containment pressure will be less than the design limit of 59.7 psia.

5.4.3.2.2 MSLB Peak Temperature Analysis The maximum peak temperature occurs for the 0.6 ft2 break at 114.3% of 2587 MWt core power. This break has a saturated steam release at an enthalpy of about 1200 Btu/lbm for the entire accident. Minimum air partial pressure, maximum containment air temperature, and 0%

humidity are conservative. Table 5.4-16 compares the analysis results. The increase in air pressure causes an increase in containment peak pressure but reduces the containment peak temperature.

Figures 5.4-8 and 5.4-9 show the containment pressure and vapor temperature. The containment temperature peaks at 31 seconds when the break flow is reduced suddenly by the isolation of the non-faulted SGs from the steam line header. The vapor temperature decrease starting at 101 seconds is driven by the delivery of containment spray to the atmosphere. Containment pressure drops rapidly once operator action terminates AFW to the faulted SG at 30 minutes, which stops the steam release to the containment.

The analyses included an additional 1 ft2 thermal conductor to determine a conservative containment liner temperature response in accordance with Section 3.3.3 of DOM-NAF-3A. The conductor used a 1.2 multiplier on the Direct/DLM heat transfer coefficient. The peak liner temperature for the proposed configuration was 251.1°F at 490 seconds, so the sustained superheat does not adversely affect the containment liner.

Revision 52Updated Online 09/30/20 SPS UFSAR 5.4-24

5.4 REFERENCES

1. USAEC, Division of Reactor Licensing 1972. Safety Evaluation Report for Virginia Electric Power Company, Surry Power Station Units 1 and 2. Docket 50-280 and 50-281.

2. Westinghouse LOCA Mass and Energy Release Model for Containment Design - March 1979 Version, WCAP-10325-P-A, May 1983 (Proprietary), WCAP-10326-A (Non-Proprietary).

3. Westinghouse ECCS Evaluation Model - 1981 Version, WCAP-9220-P-A, Rev. 1, February 1982 (Proprietary), WCAP-9221-A, Rev. 1 (Non-Proprietary).

4. Docket No. 50-315, Amendment No. 126, Facility Operating License No. DPR-58 (TAC No. 7106), for D. C. Cook Nuclear Plant Unit 1, June 9, 1989.

5. EPRI 294-2, Mixing of Emergency Core Cooling Water with Steam; 1/3 Scale Test and Summary, (WCAP-8423), Final Report June 1975.

6. Westinghouse Mass and Energy Release Data For Containment Design, WCAP-8264-P-A, Rev. 1, August 1975 (Proprietary), WCAP-8312-A (Non-Proprietary).

7. ANSI/ANS-5.1-1979, American National Standard for Decay Heat Power in Light Water Reactors, August 1979.

8. Topical Report DOM-NAF-3, Rev. 0.0-P-A, GOTHIC Methodology For Analyzing the Response to Postulated Pipe Ruptures Inside Containment, September 2006.

9. Letter from W.L. Stewart to Harold R. Denton (NRC), Supplement to An Amendment to Operating Licenses DPR-32 and DPR Proposed Reduction in Boron Concentrations -

Surry Power Station Units 1 and 2, Serial No. 521B, November 30, 1983.

10. Letter from G.T. Bischof of Virginia Electric and Power Company to USNRC Document Control Desk, Virginia Electric and Power Company, Surry Power Station Units 1 and 2, License Amendment Request, Alternative Containment Analysis Methodology, Serial No. 07-0693, October 22, 2007.

11. Letter from USNRC to Virginia Electric and Power Company, Safety Evaluation Report approving the Alternative Containment Analysis Methodology, November 15, 2007.

12. Westinghouse Letter, Surry LOCA Mass and Energy Reanalysis Report to Address Analysis Issues, VRA-07-57, December 11, 2007.

Revision 52Updated Online 09/30/20 SPS UFSAR 5.4-25 Table 5.4-1 LOCA MASS & ENERGY RELEASE ANALYSIS SYSTEM PARAMETERS INITIAL CONDITIONS Parameters Value Core Thermal Power (100.38% of 2587 MWt) 2597 MWt Reactor Coolant System Flowrate, per Loop 88,500 gpm Vessel Outlet Temperature 605.6°Fa Core Inlet Temperature 540.4°Fa Vessel Average Temperature 573.0°Fa Initial Steam Generator Steam Pressure 785 psia Steam Generator Design Model 51F Steam Generator Tube Plugging 0%

Total SG Dry Weight 691,000 lbm SG Weight in Contact with Secondary Water 474,000 lbm Initial SG Secondary-Side fluid 113,740 lbm Assumed Maximum Containment Backpressure 59.7 psia Accumulator Water Volume 1000 ft3 N2 Cover Gas Pressure 600 psia Temperature 105°F Safety Injection Delay 27.0 sec (includes time to reach pressure setpoint)

a. These are nominal values; analysis value includes +4.0°F allow-ance for instrument error and deadband

Revision 52Updated Online 09/30/20 SPS UFSAR 5.4-26 Table 5.4-2 LOCA MASS & ENERGY RELEASE ANALYSIS SYSTEM PARAMETERS CORE DECAY HEAT FRACTION Time Decay Heat (sec) (Btu/Btu) 10 0.052168 15 0.048917 20 0.047448 40 0.041405 60 0.038402 80 0.036324 100 0.03476 150 0.032104 200 0.03036 400 0.0266 600 0.024426 800 0.022885 1000 0.021666 1500 0.019429 2000 0.017851 4000 0.014334 6000 0.01263 8000 0.011588 10000 0.010856 15000 0.01013 20000 0.009368 40000 0.007784 60000 0.006976 80000 0.006439 100000 0.006034 150000 0.005336 200000 0.004859 400000 0.003781 600000 0.003212 800000 0.002844 1000000 0.002589 1500000 0.002175 2000000 0.001915 4000000 0.001356 6000000 0.00109 8000000 0.000924 10000000 0.000804

Revision 52Updated Online 09/30/20 SPS UFSAR 5.4-27 Table 5.4-3 LOCA MASS & ENERGY RELEASE ANALYSIS SAFETY INJECTION FLOW MAXIMUM SI - SINGLE TRAIN INJECTION MODE (REFLOOD PHASE)

RCS Pressure (psig) Total Flow (gpm) 0 3978.7 40 3975.1 80 3649.0 120 2975.6 160 1513.2 175 753.9 200 515.2 INJECTION MODE (POST-REFLOOD PHASE)

RCS Pressure (psig) Total Flow (gpm) 45 3839 RECIRCULATION MODE RCS Pressure (psig) Total Flow (gpm) 0 3330

Revision 52Updated Online 09/30/20 SPS UFSAR 5.4-28 Table 5.4-4 LOCA MASS & ENERGY RELEASE ANALYSIS SAFETY INJECTION FLOW MINIMUM SI - SINGLE TRAIN INJECTION MODE (REFLOOD PHASE)

RCS Pressure (psig) Total Flow (gpm) 0 3303.6 40 3300.7 80 2771.0 120 1881.5 160 501.2 165 376.4 200 394.5 INJECTION MODE (POST-REFLOOD PHASE)

RCS Pressure (psig) Total Flow (gpm) 45 3280 RECIRCULATION MODE RCS Pressure (psig) Total Flow (gpm) 0 2900.4

Revision 52Updated Online 09/30/20 SPS UFSAR 5.4-29 Table 5.4-5 LOCA MASS & ENERGY RELEASE ANALYSIS SAFETY INJECTION FLOW MAXIMUM SI - TWO TRAIN INJECTION MODE (REFLOOD PHASE)

RCS Pressure (psig) Total Flow (gpm) 0 4784.2 40 4778.1 80 4772.1 120 4008.2 160 2439.0 175 1414.4 200 788.6 INJECTION MODE (POST-REFLOOD PHASE)

RCS Pressure (psig) Total Flow (gpm) 45 4563 RECIRCULATION MODE RCS Pressure (psig) Total Flow (gpm) 0 4100

Revision 52Updated Online 09/30/20 SPS UFSAR 5.4-30 Table 5.4-6 DEHLG, MAXIMUM SI TWO TRAIN MASS AND ENERGY RELEASES FOR CONTAINMENT ANALYSIS Break Path No. 1a Break Path No. 2b Time Flow Energy Enthalpy Flow Energy Enthalpy Thousands Thousands Seconds LBM/Sec BTU/LBM LBM/Sec BTU/LBM BTU/Sec BTU/Sec 0.00 97796.90 61319.60 627.01 97796.90 61319.60 627.01 0.05 44488.90 28002.40 629.42 26374.60 16393.20 621.55 0.10 44302.10 27906.20 629.91 25174.90 15676.20 622.69 0.15 39440.50 25074.10 635.74 24625.80 15322.90 622.23 0.20 32457.30 20820.80 641.48 23499.50 14590.50 620.89 0.25 31268.70 20074.80 642.01 21949.60 13577.10 618.56 0.30 31745.90 20358.30 641.29 20801.30 12808.40 615.75 0.35 31640.20 20286.90 641.17 20047.60 12272.10 612.15 0.40 31270.70 20055.50 641.35 19456.50 11831.30 608.09 0.45 31022.50 19902.20 641.54 19012.10 11481.90 603.93 0.50 30853.90 19801.80 641.79 18621.70 11167.20 599.69 0.55 30730.90 19734.60 642.17 18272.20 10883.10 595.61 0.60 30513.40 19612.70 642.76 18012.40 10658.60 591.74 0.65 30308.90 19502.00 643.44 17727.50 10424.00 588.01 0.70 30155.40 19424.10 644.13 17504.70 10233.10 584.59 0.75 30039.70 19372.90 644.91 17307.50 10061.50 581.34 0.80 29916.00 19320.10 645.81 17120.00 9899.30 578.23 0.85 29771.70 19256.40 646.80 16928.90 9740.80 575.39 0.90 29607.90 19183.10 647.90 16757.40 9597.00 572.70 0.95 29421.00 19096.60 649.08 16615.10 9474.00 570.20 1.00 29205.90 18994.20 650.35 16489.00 9362.90 567.83 1.05 28914.90 18843.50 651.69 16384.00 9266.80 565.60 1.10 28819.70 18825.50 653.22 16252.20 9158.20 563.51 1.15 28596.10 18729.30 654.96 16138.10 9062.30 561.55 1.20 28397.80 18648.90 656.70 16054.50 8985.80 559.71 1.25 28238.70 18594.40 658.47 15973.10 8912.10 557.94 1.30 28027.80 18502.30 660.14 15917.50 8855.00 556.31 1.35 27815.50 18402.70 661.60 15885.40 8812.20 554.74 1.40 27606.20 18300.00 662.89 15868.80 8778.70 553.21 1.45 27409.20 18201.70 664.07 15863.30 8752.80 551.76

Revision 52Updated Online 09/30/20 SPS UFSAR 5.4-31 Table 5.4-6 (CONTINUED)

DEHLG, MAXIMUM SI TWO TRAIN MASS AND ENERGY RELEASES FOR CONTAINMENT ANALYSIS Break Path No. 1a Break Path No. 2b Time Flow Energy Enthalpy Flow Energy Enthalpy Thousands Thousands Seconds LBM/Sec BTU/LBM LBM/Sec BTU/LBM BTU/Sec BTU/Sec 1.50 27222.10 18108.40 665.21 15867.70 8733.20 550.38 1.55 27061.10 18032.90 666.38 15879.90 8718.60 549.03 1.60 26892.10 17952.70 667.58 15898.70 8708.40 547.74 1.65 26712.80 17866.40 668.83 15922.70 8701.30 546.47 1.70 26525.00 17773.60 670.07 15950.60 8697.60 545.28 1.75 26323.70 17670.70 671.28 15981.50 8696.00 544.13 1.80 26098.50 17550.30 672.46 16014.30 8696.40 543.04 1.85 25858.10 17417.60 673.58 16047.60 8697.70 541.99 1.90 25599.30 17270.50 674.65 16082.80 8700.60 540.99 1.95 25337.50 17119.20 675.65 16119.30 8705.10 540.04 2.00 25081.60 16970.10 676.60 16156.60 8711.20 539.17 2.05 24829.40 16823.20 677.55 16196.00 8718.80 538.33 2.10 24591.10 16685.20 678.51 16234.90 8727.20 537.56 2.15 24365.00 16555.70 679.49 16273.30 8736.10 536.84 2.20 24153.70 16436.40 680.49 16310.30 8745.00 536.16 2.25 23947.60 16320.40 681.50 16345.40 8753.70 535.55 2.30 23744.10 16204.80 682.48 16376.50 8761.00 534.97 2.35 23533.50 16082.00 683.37 16405.10 8767.80 534.46 2.40 23310.60 15947.70 684.14 16430.70 8773.60 0.00 2.45 23085.40 15808.00 0.00 16451.70 8777.80 0.00 2.50 22856.50 15663.00 685.28 16469.00 8780.40 533.15 2.55 22640.50 15524.60 685.70 16482.40 8781.90 532.80 2.60 22426.90 15387.00 686.10 16492.50 8781.90 532.48 2.65 22223.30 15255.20 686.45 16498.60 8780.50 532.20 2.70 22024.80 15125.80 686.76 16501.30 8777.60 531.93 2.75 21832.80 14999.40 687.01 16500.00 8773.30 531.72 2.80 21648.60 14877.10 687.21 16495.30 8767.60 531.52 2.85 21471.40 14758.60 687.36 16487.10 8760.30 531.34 2.90 21305.30 14646.70 687.47 16475.80 8751.90 531.20 2.95 21141.40 14535.40 687.53 16461.00 8741.90 531.07

Revision 52Updated Online 09/30/20 SPS UFSAR 5.4-32 Table 5.4-6 (CONTINUED)

DEHLG, MAXIMUM SI TWO TRAIN MASS AND ENERGY RELEASES FOR CONTAINMENT ANALYSIS Break Path No. 1a Break Path No. 2b Time Flow Energy Enthalpy Flow Energy Enthalpy Thousands Thousands Seconds LBM/Sec BTU/LBM LBM/Sec BTU/LBM BTU/Sec BTU/Sec 3.00 20982.90 14425.80 687.50 16443.30 8730.80 530.96 3.05 20821.10 14311.70 687.37 16422.30 8718.30 530.88 3.10 20663.70 14198.20 687.11 16398.70 8704.70 530.82 3.15 20509.00 14084.80 686.76 16371.40 8689.30 530.76 3.20 20369.20 13981.00 686.38 16342.60 8673.50 530.73 3.25 20234.70 13880.30 685.97 16311.00 8656.50 530.72 3.30 20106.20 13782.60 685.49 16277.00 8638.40 530.71 3.35 19976.50 13682.60 684.93 16240.20 8619.10 530.73 3.40 19854.30 13586.20 684.30 16202.30 8599.40 530.75 3.45 19736.10 13491.10 683.57 16161.20 8578.20 530.79 3.50 19626.80 13401.30 682.81 16117.90 8556.10 530.84 3.55 19525.90 13317.20 682.03 16072.70 8533.10 530.91 3.60 19430.00 13236.00 681.21 16026.00 8509.60 530.99 3.65 19334.10 13153.10 680.31 15976.20 8484.50 531.07 3.70 19235.50 13067.10 679.32 15924.20 8458.50 531.17 3.75 19149.10 12987.80 678.25 15869.80 8431.40 531.29 3.80 19063.00 12908.60 677.15 15814.40 8403.80 531.40 3.85 18989.40 12837.20 676.02 15755.40 8374.60 531.54 3.90 18920.00 12768.20 674.85 15695.10 8344.70 531.68 3.95 18855.00 12700.80 673.60 15631.50 8313.40 531.84 4.00 18795.50 12636.50 672.32 15566.10 8281.10 532.00 4.10 18695.90 12518.70 669.60 15426.70 8212.70 532.37 4.20 18624.40 12418.70 666.80 15276.10 8138.60 532.77 4.30 18580.50 12333.00 663.76 15107.90 8055.50 533.20 4.40 18582.10 12274.00 660.53 14936.50 7971.10 533.67 4.50 18647.10 12257.60 657.35 14755.90 7882.20 534.17 4.60 18784.30 12281.60 653.82 14573.10 7792.10 534.69 4.70 19026.60 12349.10 649.04 14388.80 7701.50 535.24 4.80 13607.30 9894.10 727.12 14222.70 7621.00 535.83 4.90 14252.30 10254.80 719.52 14062.40 7543.50 536.43

Revision 52Updated Online 09/30/20 SPS UFSAR 5.4-33 Table 5.4-6 (CONTINUED)

DEHLG, MAXIMUM SI TWO TRAIN MASS AND ENERGY RELEASES FOR CONTAINMENT ANALYSIS Break Path No. 1a Break Path No. 2b Time Flow Energy Enthalpy Flow Energy Enthalpy Thousands Thousands Seconds LBM/Sec BTU/LBM LBM/Sec BTU/LBM BTU/Sec BTU/Sec 5.00 14424.20 10285.60 713.08 13866.70 7446.60 537.01 5.10 14605.20 10262.10 702.63 13630.90 7327.90 537.59 5.20 14886.20 10353.60 695.52 13372.30 7196.90 538.19 5.30 14977.50 10337.20 690.18 13123.60 7071.30 538.82 5.40 15118.00 10351.70 684.73 12888.60 6953.50 539.51 5.50 15253.20 10361.80 679.32 12662.90 6840.80 540.22 5.60 15343.10 10337.90 673.78 12444.40 6732.20 540.98 5.70 15482.50 10371.60 669.89 12228.00 6624.50 541.75 5.80 15582.80 10359.70 664.82 11999.20 6510.30 542.56 5.90 15691.40 10369.20 660.82 11784.20 6402.90 543.35 6.00 15806.00 10384.30 656.98 11561.40 6291.20 544.16 6.10 15930.80 10377.40 651.40 11345.20 6182.40 544.94 6.20 16035.90 10422.40 649.94 11130.10 6074.10 545.74 6.30 15700.70 10207.60 650.14 10923.10 5969.50 546.50 6.40 15884.50 10265.40 646.25 10720.30 5866.80 547.26 6.50 16018.80 10297.30 642.83 10519.20 5764.50 548.00 6.60 16143.60 10328.00 639.76 10322.40 5664.20 548.73 6.70 16262.90 10359.10 636.98 10129.00 5565.20 549.43 6.80 16383.70 10392.70 634.33 9941.40 5469.10 550.13 6.90 16511.10 10431.50 631.79 9763.80 5378.00 550.81 7.00 16622.80 10462.20 629.39 9589.30 5288.30 551.48 7.10 16736.10 10497.00 627.21 9423.80 5203.20 552.13 7.20 16915.30 10567.20 624.71 9258.20 5117.80 552.79 7.30 17036.60 10605.50 622.51 9099.70 5036.10 553.44 7.40 17221.10 10679.40 620.13 8943.40 4955.50 554.10 7.50 17464.70 10781.90 617.35 8792.40 4877.80 554.77 7.60 17776.00 10916.80 614.13 8647.20 4803.10 555.45 7.70 17969.50 11012.40 612.84 8507.70 4731.20 556.11 7.80 17881.00 10938.90 611.76 8367.00 4658.80 556.81 7.90 17784.70 10860.80 610.68 8232.10 4589.40 557.50

Revision 52Updated Online 09/30/20 SPS UFSAR 5.4-34 Table 5.4-6 (CONTINUED)

DEHLG, MAXIMUM SI TWO TRAIN MASS AND ENERGY RELEASES FOR CONTAINMENT ANALYSIS Break Path No. 1a Break Path No. 2b Time Flow Energy Enthalpy Flow Energy Enthalpy Thousands Thousands Seconds LBM/Sec BTU/LBM LBM/Sec BTU/LBM BTU/Sec BTU/Sec 8.00 17644.80 10756.00 609.58 8098.80 4520.80 558.21 8.10 17477.70 10640.30 608.79 7967.80 4453.40 558.92 8.20 16639.10 10196.20 612.79 7837.90 4386.80 559.69 8.30 15243.20 9447.20 619.76 7708.80 4320.50 560.46 8.40 14863.10 9234.20 621.28 7585.40 4257.20 561.24 8.50 14843.50 9205.90 620.20 7463.80 4195.20 562.07 8.60 14873.80 9206.00 618.94 7342.50 4133.30 562.93 8.70 14912.50 9214.00 617.87 7225.00 4073.60 563.82 8.80 14950.90 9223.80 616.94 7110.90 4015.90 564.75 8.90 14938.20 9205.50 616.24 7002.30 3961.30 565.71 9.00 14894.00 9171.30 615.77 6897.30 3908.80 566.71 9.10 14834.20 9128.90 615.40 6795.30 3857.70 567.70 9.20 14756.40 9075.70 615.03 6692.10 3805.90 568.72 9.30 14654.70 9009.00 614.75 6591.70 3755.50 569.73 9.40 14514.20 8921.30 614.66 6491.40 3705.30 570.80 9.50 14327.20 8809.30 614.87 6393.60 3656.60 571.92 9.60 14104.00 8679.20 615.37 6298.00 3608.90 573.02 9.70 13867.90 8542.90 616.02 6204.40 3562.40 574.17 9.80 13636.30 8408.90 616.66 6108.40 3514.70 575.39 9.90 13438.50 8293.70 617.16 6017.10 3469.80 576.66 10.00 13263.90 8190.60 617.51 5925.70 3424.90 577.97 10.10 13098.80 8092.70 617.82 5833.40 3379.80 579.39 10.20 12945.80 8002.30 618.14 5745.60 3336.90 580.77 10.30 12786.10 7908.90 618.55 5656.40 3293.70 582.30 10.40 12630.00 7818.40 619.03 5571.70 3252.90 583.83 10.50 12467.60 7725.10 619.61 5487.20 3212.20 585.40 10.60 12302.20 7630.70 620.27 5405.40 3173.00 587.01 10.70 12132.30 7534.50 621.03 5323.40 3133.80 588.68 10.80 11955.90 7435.50 621.91 5241.50 3094.80 590.44 10.90 11782.80 7338.60 622.82 5162.20 3057.10 592.21

Revision 52Updated Online 09/30/20 SPS UFSAR 5.4-35 Table 5.4-6 (CONTINUED)

DEHLG, MAXIMUM SI TWO TRAIN MASS AND ENERGY RELEASES FOR CONTAINMENT ANALYSIS Break Path No. 1a Break Path No. 2b Time Flow Energy Enthalpy Flow Energy Enthalpy Thousands Thousands Seconds LBM/Sec BTU/LBM LBM/Sec BTU/LBM BTU/Sec BTU/Sec 11.00 11608.10 7241.60 623.84 5083.80 3020.00 594.04 11.10 11439.50 7148.50 624.90 5008.00 2984.30 595.91 11.20 11269.20 7054.80 626.02 4932.70 2949.00 597.85 11.30 11096.90 6960.60 627.26 4856.50 2913.20 599.86 11.40 10932.60 6871.60 628.54 4784.80 2879.80 601.86 11.50 10762.70 6780.00 629.95 4710.20 2845.10 604.03 11.60 10602.30 6694.60 631.43 4641.40 2813.40 606.15 11.70 10434.20 6605.20 633.03 4570.70 2780.80 608.40 11.80 10268.60 6517.80 634.73 4501.80 2749.10 610.67 11.90 10091.70 6424.80 636.64 4431.50 2716.70 613.04 12.00 9909.40 6329.80 638.77 4362.90 2685.20 615.46 12.10 9719.60 6232.00 641.18 4293.90 2653.50 617.97 12.20 9515.50 6128.40 644.04 4221.10 2620.00 620.69 12.30 9305.20 6022.70 647.24 4146.90 2585.80 623.55 12.40 9085.40 5914.00 650.93 4069.60 2550.40 626.70 12.50 8858.80 5803.80 655.15 3988.50 2513.50 630.19 12.60 8626.30 5693.00 659.96 3903.40 2475.40 634.17 12.70 8396.50 5585.60 665.23 3815.90 2436.60 638.54 12.80 8164.40 5480.20 671.23 3723.80 2396.70 643.62 12.90 7933.60 5379.20 678.03 3626.70 2355.40 649.46 13.00 7705.40 5282.00 685.49 3525.70 2313.00 656.04 13.10 7480.20 5190.00 693.83 3421.60 2270.20 663.49 13.20 7258.50 5102.00 702.90 3316.20 2227.30 671.64 13.30 7038.40 5016.60 712.75 3210.30 2184.70 680.53 13.40 6816.70 4932.30 723.56 3103.20 2141.80 690.19 13.50 6594.80 4850.30 735.47 2995.10 2098.60 700.68 13.60 6379.70 4769.30 747.57 2892.90 2056.90 711.02 13.70 6162.20 4687.90 760.75 2791.30 2015.00 721.89 13.80 5946.00 4605.30 774.52 2695.10 1974.50 732.63 13.90 5735.40 4523.10 788.63 2605.50 1935.50 742.85

Revision 52Updated Online 09/30/20 SPS UFSAR 5.4-36 Table 5.4-6 (CONTINUED)

DEHLG, MAXIMUM SI TWO TRAIN MASS AND ENERGY RELEASES FOR CONTAINMENT ANALYSIS Break Path No. 1a Break Path No. 2b Time Flow Energy Enthalpy Flow Energy Enthalpy Thousands Thousands Seconds LBM/Sec BTU/LBM LBM/Sec BTU/LBM BTU/Sec BTU/Sec 14.00 5525.20 4441.00 803.77 2522.00 1898.40 752.74 14.10 5320.70 4362.40 819.89 2443.20 1861.90 762.07 14.20 5120.90 4281.60 836.10 2371.90 1827.70 770.56 14.30 4918.70 4196.00 853.07 2307.30 1795.60 778.23 14.40 4720.00 4111.70 871.12 2249.40 1765.80 785.01 14.50 4518.50 4027.30 891.29 2195.60 1737.00 791.13 14.60 4302.80 3942.10 916.17 2147.00 1710.20 796.55 14.70 4045.90 3823.60 945.06 2102.80 1685.30 801.46 14.80 3784.80 3662.80 967.77 2061.60 1661.70 806.02 14.90 3563.70 3503.90 983.22 2023.00 1639.80 810.58 15.00 3396.50 3366.60 991.20 1986.50 1619.00 815.00 15.10 3258.30 3253.70 998.59 1951.70 1599.20 819.39 15.20 3114.70 3144.50 1009.57 1918.70 1580.60 823.79 15.30 2973.40 3041.90 1023.04 1885.20 1561.60 828.35 15.40 2827.40 2941.90 1040.50 1853.10 1543.80 833.09 15.50 2678.30 2841.30 1060.86 1820.20 1525.90 838.31 15.60 2526.10 2742.80 1085.78 1785.90 1508.10 844.45 15.70 2371.10 2640.80 1113.74 1750.90 1491.40 851.79 15.80 2225.40 2538.60 1140.74 1714.80 1475.60 860.51 15.90 2099.00 2439.30 1162.12 1676.30 1459.80 870.85 16.00 1995.10 2350.60 1178.19 1636.20 1444.50 882.84 16.10 1912.90 2277.70 1190.71 1594.40 1429.20 896.39 16.20 1879.50 2251.60 1197.98 1551.00 1413.50 911.35 16.30 1810.70 2182.20 1205.17 1506.20 1396.80 927.37 16.40 1754.80 2119.90 1208.06 1464.40 1382.50 944.07 16.50 1708.20 2071.90 1212.91 1422.20 1370.30 963.51 16.60 1651.50 2007.10 1215.32 1378.70 1355.10 982.88 16.70 1593.00 1943.30 1219.90 1339.40 1339.90 1000.37 16.80 1530.90 1874.30 1224.31 1302.50 1325.30 1017.50 16.90 1469.10 1804.50 1228.30 1269.20 1314.60 1035.77

Revision 52Updated Online 09/30/20 SPS UFSAR 5.4-37 Table 5.4-6 (CONTINUED)

DEHLG, MAXIMUM SI TWO TRAIN MASS AND ENERGY RELEASES FOR CONTAINMENT ANALYSIS Break Path No. 1a Break Path No. 2b Time Flow Energy Enthalpy Flow Energy Enthalpy Thousands Thousands Seconds LBM/Sec BTU/LBM LBM/Sec BTU/LBM BTU/Sec BTU/Sec 17.00 1410.30 1737.20 1231.79 1234.70 1301.60 1054.18 17.10 1356.40 1674.70 1234.67 1204.80 1288.00 1069.06 17.20 1317.00 1628.20 1236.29 1178.30 1274.50 1081.64 17.30 1279.60 1580.40 1235.07 1155.00 1261.40 1092.12 17.40 1238.60 1529.30 1234.70 1134.30 1248.40 1100.59 17.50 1199.50 1483.00 1236.35 1115.20 1235.40 1107.78 17.60 1142.00 1413.10 1237.39 1097.30 1223.30 1114.83 17.70 1097.10 1365.60 1244.74 1080.30 1213.20 1123.02 17.80 1057.20 1318.60 1247.26 1058.30 1203.10 1136.82 17.90 1015.60 1266.50 1247.05 1031.90 1190.80 1153.99 18.00 982.80 1224.60 1246.03 1003.60 1178.30 1174.07 18.10 946.40 1178.90 1245.67 953.50 1142.00 1197.69 18.20 912.70 1142.10 1251.34 900.50 1092.20 1212.88 18.30 871.20 1093.10 1254.71 848.70 1036.90 1221.75 18.40 826.20 1034.60 1252.24 801.50 982.70 1226.08 18.50 800.00 1003.00 1253.75 756.20 929.20 1228.78 18.60 757.70 950.60 1254.59 716.90 882.00 1230.30 18.70 722.70 907.50 1255.71 672.50 828.00 1231.23 18.80 696.40 874.70 1256.03 601.20 740.40 1231.54 18.90 663.10 832.90 1256.07 514.50 634.40 1233.04 19.00 637.70 801.00 1256.08 453.80 560.90 1236.01 19.10 618.20 776.40 1255.90 380.30 470.50 1237.18 19.20 598.20 751.20 1255.77 322.40 399.90 1240.38 19.30 580.30 728.50 1255.39 281.40 349.90 1243.43 19.40 565.30 709.40 1254.91 247.90 308.90 1246.07 19.50 551.20 691.10 1253.81 231.90 289.50 1248.38 19.60 536.20 672.20 1253.64 232.80 291.10 1250.43 19.70 510.70 639.50 1252.20 198.30 247.80 1249.62 19.80 494.60 617.00 1247.47 105.20 131.80 1252.85 19.90 499.80 622.30 1245.10 0.00 0.00 0.00

Revision 52Updated Online 09/30/20 SPS UFSAR 5.4-38 Table 5.4-6 (CONTINUED)

DEHLG, MAXIMUM SI TWO TRAIN MASS AND ENERGY RELEASES FOR CONTAINMENT ANALYSIS Break Path No. 1a Break Path No. 2b Time Flow Energy Enthalpy Flow Energy Enthalpy Thousands Thousands Seconds LBM/Sec BTU/LBM LBM/Sec BTU/LBM BTU/Sec BTU/Sec 20.00 298.80 376.20 1259.04 0.00 0.00 0.00 20.10 66.40 85.30 1284.64 0.00 0.00 0.00 20.20 0.00 0.00 0.00 71.80 93.50 1302.23 20.30 0.00 0.00 0.00 0.00 0.00 0.00 20.5 0 0.00 0.00 0.00 0.00 0.00 20.6 0 0.00 0.00 0.00 0.00 0.00 20.8 574.6 151.30 263.31 0.00 0.00 0.00 20.9 349.2 197.30 565.01 0.00 0.00 0.00 23.4 1094.8 354.20 323.53 0.00 0.00 0.00 26.6 1692 463.51 273.94 0.00 0.00 0.00 29.3 1918.1 502.29 261.87 1799.7 136.90 76.07 33.4 1852.1 471.80 254.74 2894.5 192.60 66.54 45.3 1650.1 430.00 260.59 2011 129.41 64.35 46 1641.3 363.60 221.53 0.00 0.00 0.00 50 1376 311.80 226.6 0.00 0.00 0.00 57.7 866 264.10 304.97 0.00 0.00 0.00 88 420.8 218.80 519.96 0.00 0.00 0.00 100 407 214.60 527.27 0.00 0.00 0.00 115.3 395.6 210.60 532.36 0.00 0.00 0.00

a. Break Path No. 1 refers to the mass and energy exiting from the reactor vessel side of the break.

b. Break Path No. 2 refers to the mass and energy exiting from the SG side of the break.

Revision 52Updated Online 09/30/20 SPS UFSAR 5.4-39 Table 5.4-7 DEPSG, MINIMUM SI SINGLE TRAIN MASS AND ENERGY RELEASES FOR CONTAINMENT ANALYSIS Break Path No. 1a Break Path No. 2b Time Flow Energy Enthalpy Flow Energy Enthalpy Thousands Thousands Seconds LBM/Sec BTU/LBM LBM/Sec BTU/LBM BTU/Sec BTU/Sec 0.00 147896.10 79856.5 539.95 147896.10 79856.5 539.95 0.10 39713.70 21300.9 536.36 20013.90 10695.1 534.38 0.20 40315.50 21753.3 539.58 21993.10 11761.8 534.80 0.30 41093.10 22346.5 543.80 22177.50 11872.6 535.34 0.40 43232.40 23723.9 548.75 21755.60 11658.1 535.87 0.50 42782.70 23719.9 554.43 20914.60 11215.7 536.26 0.70 43347.40 24527.3 565.83 19421.00 10422.6 536.67 0.90 42430.10 24426.6 575.69 18386.20 9871.2 536.88 1.40 38047.10 22833.5 600.14 17469.40 9384 537.17 1.90 32587.80 20573.9 631.34 17203.40 9237.7 536.97 2.40 26788.20 17811.7 664.91 16740.80 8986.7 536.81 2.60 21346.80 14405.3 674.82 16469.90 8841.1 536.80 2.70 19921.30 13561.6 680.76 16137.70 8662.5 536.79 2.80 19578.40 13416.1 685.25 15901.80 8536.6 536.83 3.00 17642.40 12167.1 689.65 15474.90 8308.9 536.93 3.30 15505.90 10780.7 695.26 14881.20 7994.1 537.19 3.60 13835.20 9688.4 700.27 14377.10 7728.2 537.54 4.20 11874.60 8402.7 707.62 13449.60 7239.7 538.28 5.00 10422.10 7381.8 708.28 12320.00 6642.2 539.14 5.40 10019.40 7044.7 703.11 13061.30 7047.3 539.56 5.80 10327.70 7242.9 701.31 12624.70 6816.1 539.90 6.20 8542.20 6757.4 791.06 12130.10 6553.3 540.25 6.40 8097.20 6512 804.23 11957.80 6462.8 540.47 7.00 8215.30 6324.4 769.83 11673.70 6316.4 541.08 7.80 9041.50 6384.4 706.12 11048.00 5972.3 540.58 8.40 9227.10 6239 676.16 10626.70 5741.2 540.26 10.60 6898.70 5030.9 729.25 9319.50 5030.2 539.75 11.40 6208.00 4696.1 756.46 8803.20 4750.7 539.66 13.60 4872.30 3881.1 796.56 7388.60 3988.2 539.78 14.20 4488.80 3608 803.78 6848.50 3597.5 525.30 14.40 4360.70 3513 805.60 7342.20 3776.3 514.33 14.80 4151.20 3349.2 806.80 6083.70 3003.3 493.66 15.20 3944.60 3212.1 814.30 11223.70 5460 486.47 15.40 3774.30 3120.5 826.78 9121.60 4461.5 489.11

Revision 52Updated Online 09/30/20 SPS UFSAR 5.4-40 Table 5.4-7 (CONTINUED)

DEPSG, MINIMUM SI SINGLE TRAIN MASS AND ENERGY RELEASES FOR CONTAINMENT ANALYSIS Break Path No. 1a Break Path No. 2b Time Flow Energy Enthalpy Flow Energy Enthalpy Thousands Thousands Seconds LBM/Sec BTU/LBM LBM/Sec BTU/LBM BTU/Sec BTU/Sec 15.60 3710.30 3136 845.21 5303.40 2586.9 487.78 15.80 3726.10 3184.7 854.70 5139.50 2397.4 466.47 16.00 3574.00 3135.3 877.25 11125.80 5090.8 457.57 16.40 3252.00 3084.5 948.49 4818.10 2276.2 472.43 16.60 3205.30 3119.5 973.23 4110.80 1872.6 455.53 16.80 3034.10 3078.5 1014.63 6391.90 2715.1 424.77 17.00 2718.90 2921.3 1074.44 8144.90 3472.7 426.36 17.20 2410.40 2781.5 1153.96 5522.90 2371.9 429.47 17.40 2200.70 2659.1 1208.30 4366.60 1882.7 431.16 17.80 1844.60 2273.6 1232.57 3174.80 1316.6 414.70 18.20 1485.10 1841.7 1240.12 4472.10 1677.7 375.15 19.20 866.70 1084.4 1251.18 3306.50 1137.1 343.90 20.20 460.20 578.2 1256.41 1917.70 585.1 305.11 20.60 395.50 497.6 1258.15 1276.70 372.3 291.61 21.20 282.90 356.4 1259.81 0.00 0 0.00 22.60 0.00 0 0.00 0.00 0 0.00 23.50 0.00 0 0.00 0.00 0 0.00 23.60 44.50 52.4 1177.53 0.00 0 0.00 23.70 25.30 29.8 1177.87 0.00 0 0.00 24.20 51.70 60.9 1177.95 0.00 0 0.00 25.00 79.50 93.6 1177.36 0.00 0 0.00 26.60 118.80 139.9 1177.61 0.00 0 0.00 27.60 137.90 162.5 1178.39 0.00 0 0.00 28.60 211.40 249.5 1180.23 1667.90 178.9 107.26 29.10 388.60 460 1183.74 3898.50 440.2 112.92 29.70 424.10 502.4 1184.63 4232.10 493.8 116.68 30.70 446.00 528.5 1184.98 4460.00 498.2 111.70 32.70 427.50 506.4 1184.56 4282.20 481.9 112.54 34.70 410.10 485.6 1184.10 4111.90 465.6 113.23 35.70 401.90 475.9 1184.13 4030.80 457.9 113.60 37.70 386.70 457.7 1183.60 3877.20 443.1 114.28 39.70 372.80 441.1 1183.21 3734.30 429.4 114.99 41.70 360.00 425.9 1183.06 3601.20 416.7 115.71 43.70 348.20 411.9 1182.94 3476.90 404.8 116.43

Revision 52Updated Online 09/30/20 SPS UFSAR 5.4-41 Table 5.4-7 (CONTINUED)

DEPSG, MINIMUM SI SINGLE TRAIN MASS AND ENERGY RELEASES FOR CONTAINMENT ANALYSIS Break Path No. 1a Break Path No. 2b Time Flow Energy Enthalpy Flow Energy Enthalpy Thousands Thousands Seconds LBM/Sec BTU/LBM LBM/Sec BTU/LBM BTU/Sec BTU/Sec 44.70 342.70 405.3 1182.67 3417.70 399.1 116.77 46.70 332.20 392.8 1182.42 3304.70 388.3 117.50 47.80 243.70 287.7 1180.55 236.70 109.6 463.03 48.80 251.20 296.6 1180.73 239.60 113.4 473.29 50.80 244.70 288.8 1180.22 237.10 110.2 464.78 56.80 226.50 267.3 1180.13 230.40 101.4 440.10 60.80 215.20 253.9 1179.83 226.20 96 424.40 68.80 195.50 230.6 1179.54 219.10 86.8 396.17 69.80 193.30 227.9 1179.00 218.30 85.7 392.58 73.80 184.60 217.7 1179.31 215.30 81.8 379.93 81.80 169.30 199.6 1178.97 210.00 75 357.14 89.80 156.20 184.1 1178.62 205.60 69.4 337.55 90.30 155.50 183.2 1178.14 205.30 69.1 336.58 97.80 145.30 171.3 1178.94 202.00 64.8 320.79 105.80 136.40 160.7 1178.15 199.20 61.2 307.23 113.80 129.30 152.3 1177.88 196.90 58.4 296.60 121.80 123.70 145.8 1178.66 195.20 56.2 287.91 129.80 119.60 140.9 1178.09 193.90 54.5 281.07 137.80 116.50 137.3 1178.54 192.90 53.3 276.31 139.80 116.10 136.7 1177.43 192.80 53.1 275.41 145.80 115.10 135.7 1178.97 195.00 53.3 273.33 149.80 114.60 135 1178.01 198.90 54 271.49 153.80 113.80 134.1 1178.38 204.60 55.1 269.31 157.80 112.90 133 1178.03 211.70 56.5 266.89 165.80 109.90 129.4 1177.43 229.00 59.8 261.14 173.80 109.40 128.9 1178.24 244.20 62.3 255.12 175.80 109.20 128.6 1177.66 247.60 62.8 253.63 183.80 107.90 127.1 1177.94 259.40 63.9 246.34 191.80 106.20 125.1 1177.97 268.70 64.1 238.56 199.80 104.10 122.7 1178.67 276.00 63.7 230.80 200.70 103.90 122.4 1178.06 276.70 63.6 229.85

a. Mass and energy exiting the SG side of the break.

b. Mass and energy exiting the pump side of the break.

Revision 52Updated Online 09/30/20 SPS UFSAR 5.4-42 Table 5.4-8 THERMOPHYSICAL PROPERTIES OF PASSIVE HEAT SINK MATERIALS Thermal Specific Heat Material Conductivity Capacity Density (lbm/ft3)

(Btu/hr/ft/°F) (Btu/lbm/°F)

Concrete 1.0 0.156 142 Stainless Steel 9.4 0.12 501 Carbon Steel 27 0.10 490 Paint 0.125 0.10 110

Revision 52Updated Online 09/30/20 SPS UFSAR 5.4-43 Table 5.4-9 PASSIVE HEAT SINKS TC # Description Surface Area, ft2 Thickness, inch 1 Interior Concrete Wall 1 7740 6.0 a 2 Interior Concrete Wall 2 57,435 12.0 a 3 Interior Concrete Wall 3 51,064 18.0 a 4 Interior Concrete Wall 4 10,691 24.0 a 5 Interior Concrete Wall 5 8673 27.0 a 6 Interior Concrete Wall 6 3353 36.0 a 7 Cont Wall Below Grade b 20,108 54.375 a 8 Cont Wall Above Grade b 24,576 54.375 a 9 Containment Dome b 24,656 30.5 a 10 Containment Floor 11,757 146.65 a 11 Stainless Steel Group 1 7180 0.25 12 Stainless Steel Group 2 11,290 0.42 13 Stainless Steel Group 3 488 1.53 14 Galvanized Metal 86,459 0.066 a 15 Carbon Steel Group 1 7192 0.236 a 16 Carbon Steel Group 2 66,345 0.439 a 17 Carbon Steel Group 3 7454 0.906 a 18 Carbon Steel Group 4 c 2414 1.70 a 19 Carbon Steel Group 5 7000 2.90 a 29 Accumulators c 1276 1.0

a. Includes 0.006-inch paint layer.

b. The containment walls and dome include a liner-concrete gap conductance. The wall above grade and the dome use a constant HTC of 2.0 Btu/hr-ft2-F and a specified temperature of 95°F on the outside.

c. The MSLB model accounts for the water in the accumulators as TC #29, different than the LOCA model (empty shell grouped with other carbon steel). TC #18 surface area of 2414 ft2 is reduced by the accumulator surface area of 2073 ft2 to 341 ft2.

Revision 52Updated Online 09/30/20 SPS UFSAR 5.4-44 Table 5.4-10 CONTAINMENT LOCA ANALYSIS INITIAL CONDITIONSa Air Partial Pressure 10.1 to 11.3 psia Temperature 75 to 125°F Relative Humidity 0 to 100 percent RWST Temperature 45°F (maximum)

Service Water Temperature 25 to 100°F

a. Instrumentation uncertainties for these parameters have been included in the safety analysis.

Table 5.4-11 CONTAINMENT LOCA ANALYSIS PEAK PRESSURE RESULTS Break Location Hot Leg Break Type DEG Peak Pressure 43.95 psig Time of Peak Pressure 19.48 sec Peak Vapor Temperature 273.3°F Table 5.4-12 CONTAINMENT DEPRESSURIZATION RESULTS DEPSG Depressurization Depressurization Time Peak Pressure Single Failure ESF Train ESF Train Initial Containment Conditionsa Total Pressure 12.52 psia 10.97 psia Temperature 125.0°F 75.0°F Relative Humidity 100% 100%

Service Water Temperature 100.0°F 100.0°F Depressurization Time (< 15.7 psia) 3110 sec 3038 sec Depressurization Peak Pressure 0.34 psig 0.70 psigb Depressurization Peak Pressure Time 5206 sec 5162 sec Remains Subatmospheric Time 7732 sec 10,800 sec

a. Instrumentation uncertainties for these parameters have been included in the safety analysis.

b. Highest analysis value obtained for depressurization peak pressure.

Revision 52Updated Online 09/30/20 SPS UFSAR 5.4-45 Table 5.4-13 ACCIDENT CHRONOLOGY DEPSG, MINIMUM ESF Time (sec)a Depressurization Depressurization Peak Event 2.1 2.3 CLS High High containment pressure 99.1 99.3 Containment spray delivers to containment 1861 1757 IRS spray delivered to containment 2012 1889 ORS spray delivered to containment 3110 3038 Containment pressure reaches 15.7 psia 3775 3734 Switchover to SI recirculation mode 4343 4304 Containment spray pumps stopped 5206 5162 Depressurization peak pressure occurs 7732 10,800 Containment pressure < 14.7 psia permanently

a. Times are analysis values obtained for initial conditions given in Table 5.4-12. These time values are approximate.

Revision 52Updated Online 09/30/20 SPS UFSAR 5.4-46 Table 5.4-14 MSLB CONTAINMENT PEAK PRESSURE ANALYSIS Initial Conditions TS Containment Air Partial Pressure, psia 11.3 Initial Containment Pressure, psia a 13.52 Initial Air Temperature, F 125.5 Relative Humidity, % 100 Results Peak containment pressure, psia 59.48 Time of peak containment pressure, sec 215.7 Peak containment temperature, F 274.4 Time of peak containment temperature, sec 213.7

a. GOTHIC total pressure is TS air pressure + 0.25 psi uncertainty + 1.97 psia vapor pressure.

Table 5.4-15 ACCIDENT CHRONOLOGY MSLB PEAK PRESSURE ANALYSIS Event Time (sec)

Accident start 0.0 CLS High High containment pressure 4.2 Start SI 27.9 CS delivered to containment 101.2 Containment peak pressure 215.7 Faulted SG dryout 235.0 AFW terminated 1800 Transient Termination 7200

Revision 52Updated Online 09/30/20 SPS UFSAR 5.4-47 Table 5.4-16 MSLB CONTAINMENT PEAK TEMPERATURE ANALYSIS Initial Conditions TS Containment Air Partial Pressure, psia 10.1 Initial Containment Pressure, psia a 9.85 Initial Air Temperature, F 125.5 Relative Humidity, % 0 Results Peak containment temperature, F 318.9 Time of peak containment temperature, sec 31.1 Peak containment pressure, psia 47.4 Time of peak containment pressure, sec 412.1

a. GOTHIC containment pressure = TS air pressure - 0.25 psi uncertainty (no vapor pressure).

Revision 52Updated Online 09/30/20 SPS UFSAR 5.4-48 Table 5.4-17 KEY PARAMETERS IN THE CONTAINMENT ANALYSIS Parameter Value Maximum Core Power (100.38% x 2587 rated thermal power), MWt 2597 Containment Air Partial Pressure Uncertainty, psi +/- 0.25 Containment Temperature, °F (includes 0.5°F uncertainty) 74.5-125.5 Containment Relative Humidity, % 0-100 SW Temperature, °F 24-101 RWST Temperature, °F (includes 1.6°F uncertainty) a 32-46.6 Accumulator Pressure, psia (includes uncertainty) 578-706 Accumulator Temperature, °F 75-105 Accumulator Water Volume, ft3 975-1025 Accumulator Nitrogen Volume, ft3 (includes uncertainty) 416-484 Minimum Service Water Flow Rate with 10% RSHX tube plugging, gpm 7789 at Accident Start b Maximum Service Water Flow Rate with 0% RSHX tube plugging, gpm 10,000 b ORS Pump Flow Rate, gpm 2900-3300 IRS Pump Flow Rate, gpm 3100-3650 LHSI Injection Mode Flow Rate (Single-Train), gpm 2844-3264 Maximum LHSI Recirculation Mode Flow Rate (Single-Train), gpm 3330 HHSI Injection Mode Flow Rate (Single-Train), gpm 435-528 Minimum CS Bleed Flow Rate to ORS Pump Suction, gpm See Note c Minimum IRS Recirculation Flow Rate to Pump Suction, gpm 300 CS Flow Rate, gpm See Note d IRS Piping Fill Volumes, ft3 358-421.3 ORS Piping Fill Volumes, ft3 456.5-558.1

a. Minimum RWST temperature of 32°F is assumed for evaluation of the inadvertent CS actuation event.

Normal operating range for RWST temperature is 40-45°F.

b. SW minimum flow rate decreases as the intake canal level decreases during the accident. The initial value is specified for a canal level of 23 ft. For maximum flow, a constant 10,000 gpm is assumed throughout the accident (ORS pump NPSHa analyses are not very sensitive to this input).

c. For the RS pump NPSH analyses, the CS bleed flow is input as a function of differential pressure between the containment and the RWST (C-L in psid). The flow rate varies from 294 gpm (26.8 psid) to 325.6 gpm (-8.6 psid and maintained constant for more negative differential pressures).

d. The CS flow rate varies as a function of differential pressure between the containment and RWST (C-L in psig). The minimum single-pump flow rate varies from 2006 gpm (26.9 psid) to 2708 gpm (-4.0 psid and maintained constant for more negative differential pressures). The maximum single-pump flow rate varies from 2409 gpm (26.9 psid) to 3024 gpm (-10.0 psid and maintained constant for more negative differential pressures).

Revision 52Updated Online 09/30/20 SPS UFSAR 5.4-49 Table 5.4-17 (CONTINUED)

KEY PARAMETERS IN THE CONTAINMENT ANALYSIS Parameter Value CS Spray Delivery Delay from CLS signal, sec 59-97 LHSI Pump Suction Friction Loss at maximum 1-pump flow, ft 6.8 ORS Pump Suction Friction Loss at maximum flow, ft 6.8 IRS Pump Suction Friction Loss at maximum flow, ft 2.0 CLS High High Containment Pressure, psia 27 RWST WR Level for RS Pump Start (60% +/- 2.5% uncertainty) 57.5%-62.5%

ORS Pump Start Time Delay, seconds (+/-12 second timer uncertainty and 108-142 0 or 10 seconds for ramp to full flow depending on which is conservative)

RWST WR Level Setpoint for RMT (13.5% +/- 2.5% uncertainty) 11.0-16.0%

Time to complete RMT function, minutes 2-3 Minimum RWST volume at accident initiation, gallons 384,000 (95%

NR)

Minimum containment free volume, ft3 1,730,000 Maximum containment free volume for NPSHa Analysis, ft3 1,819,000

Revision 52Updated Online 09/30/20 SPS UFSAR 5.4-50 Figure 5.4-1 CONTAINMENT PRESSURE DEHLG PEAK PRESSURE ANALYSIS 60 55 50 45 40 Pressure (psia) 35 30 25 20 15 10 0.1 1 10 100 Time (sec)

Figure 5.4-2 CONTAINMENT VAPOR TEMPERATURE DEHLG PEAK PRESSURE ANALYSIS y

300 250 Temperature (deg F) 200 150 100 0.1 1 10 100 Time (sec)

Revision 52Updated Online 09/30/20 SPS UFSAR 5.4-51 Figure 5.4-3 CONTAINMENT PRESSURE DEPSG DEPRESSURIZATION ANALYSIS 60 55 50 45 Pressure(psia) 40 35 30 25 20 15 10 0.1 1 10 100 1000 10000 Time(sec)

Figure 5.4-4 CONTAINMENT TEMPERATURES DEPSG DEPRESSURIZATION ANALYSIS 300 250 vapor Temperature(°F) sumpliquid 200 150 100 0.1 1 10 100 1000 10000 Time(sec)

Revision 52Updated Online 09/30/20 SPS UFSAR 5.4-52 Figure 5.4-5 TOTAL RS HEAT EXCHANGER HEAT RATE DEPSG DEPRESSURIZATION ANALYSIS 250 200 HeatRate(MBTU/hr) 150 100 50 0

1000 2000 3000 4000 5000 6000 Time(sec)

Figure 5.4-6 CONTAINMENT PRESSURE 1.4 FT2 MSLB PEAK PRESSURE ANALYSIS 70 60 50 Pressure (psia) 40 30 20 10 0

0.1 1 10 100 1000 10000 Time (sec)

Revision 52Updated Online 09/30/20 SPS UFSAR 5.4-53 Figure 5.4-7 CONTAINMENT TEMPERATURE 1.4 FT2 MSLB PEAK PRESSURE ANALYSIS 300 250 200 Temperature (F) 150 100 Saturation Temperature Vapor Temperature 50 0

0.1 1 10 100 1000 10000 Time (sec)

Figure 5.4-8 CONTAINMENT PRESSURE 0.6 FT2 MSLB PEAK TEMPERATURE ANALYSIS 50 40 30 Pressure (psia) 20 10 0

0.1 1 10 100 1000 10000 Time (sec)

Revision 52Updated Online 09/30/20 SPS UFSAR 5.4-54 Figure 5.4-9 CONTAINMENT TEMPERATURE 0.6 FT2 MSLB PEAK TEMPERATURE ANALYSIS 350 300 250 Temperature (F) 200 150 100 Vapor Temperature 50 Saturation Temperature 0

1 10 100 1000 10000 Time (sec)

Revision 52Updated Online 09/30/20 SPS UFSAR 5.5-1 5.5 CONTAINMENT TESTS AND INSPECTIONS The following information is HISTORICAL and is not intended or expected to be updated for the life of the plant.

5.5.1 Initial Containment Testing A testing and surveillance program was in effect during construction and continues in effect during operation (See 5.5.3) to confirm that the containment can perform its intended function. The initial program consisted of construction testing, and an initial leakage rate test.

Materials and fabrication inspections and tests are described in Section 15.4.

Construction testing included provisions for testing the leaktightness of all penetrations and liner welds during construction and for an air pressure test when the containment was completed to ensure the structural integrity of the containment. Electrical penetrations were assembled and tested as a unit for leaktightness following installation in the containment (Section 15.5).

Leaktightness testing of all liner welds during construction was performed by welding a structural steel gas test channel over each weld seam. For the bottom and the straight shell, the test channels were placed on the inside of the liner. For the dome, the test channels were on the outside (concrete side) of the liner.

These channels formed a space into which pure Freon at 50 psig was injected. The weld seams were then tested for leakage using a halogen leak detector. The test channels did not form a single continuous channel but were segmented for convenience in testing. Test gas was introduced through threaded connections after evacuating the channels to ensure a homogeneous test gas throughout the channel section. If a leak rate of greater than 1.8 x 10-5 cm3/sec was found, the defective test channel seam or liner weld seam was ground out and the weld remade and retested. After testing, the gas was purged from the channels with air and the threaded connections plugged.

The air pressure test to ensure the structural integrity of the containment was performed after the liner was completed, the last concrete pour cured, and all penetration sleeves installed.

The containment pressure was raised to 52 psig (115.5% of the 45 psig design pressure) and held for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, thus ensuring the structural integrity of the containment.

The initial leakage rate tests were performed at 39.2 psig and 25 psig, after the completion of construction and the installation of all systems penetrating the containment boundary. On Unit 1, an additional leakage rate test was performed at 9 psia. These testes were performed using the reference volume method as described under the leakage-monitoring system (Section 5.3.2).

5.5.2 Continuing Containment Leakage Testing A testing program is in place to measure primary containment leakage periodically throughout the plants operating life. The testing program includes performance of Type A tests to

Revision 52Updated Online 09/30/20 SPS UFSAR 5.5-2 measure the overall integrated leakage rate, Type B tests to detect and measure local leakage across pressure-containing or leakage-limiting boundaries other than valves, and Type C tests to measure containment isolation valve leakage rates.

The leakage tests are performed in accordance with the requirements of 10 CFR 50 Appendix J, Primary Reactor Containment Leakage Testing for Water Cooled Power Reactors.

5.5.3 Containment Integrated Leakage Rate Test (Type A)

A Type A test program has been developed and is scheduled and conducted in accordance with 10 CFR 50 Appendix J, Option B and NEI 94-01, Revision 3-A, Industry Guidelines for Implementing Performance-Based Option of 10 CFR 50, Appendix J, Option B. Each Type A test establishes the measured containment leakage rate, Lam, which verifies that the maximum allowable leakage rate, La, used in the accident analysis is not exceeded.

Type A tests are conducted at periodic intervals based on historical performance of the overall containment system in accordance with 10 CFR 50, Appendix J, Option B, and NEI 94-01, Revision 3-A. The leakage rate must not exceed the allowable leakage rate (La) with margin as specified in the Technical Specifications.

A general inspection of the accessible interior and exterior surfaces of the containment is performed, prior to each Type A test and at periodic intervals between tests based on performance, to detect structural deterioration. Defects are resolved prior to conducting the test.

Test instrumentation includes an absolute manometer, temperature detectors, and dewpoint sensors.

The containment isolation valves are closed by their normal mode of operation. Where possible, lines subjected to containment atmosphere following a LOCA are drained and vented during the Type A test. Systems that are normally filled with water and operating under post-accident conditions are not drained and vented nor are their Type C test results included in the Type A test leakage rate. For those systems that should be vented and drained but are not, the Type C test result is added to the Type A leakage rate.

After completion of the procedural prerequisites, the containment is pressurized to slightly greater than containment design pressure (45 psig) and allowed to stabilize for a minimum of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. The Type A test period normally commences when the rate of change of the containment air temperature for the latest hour does not deviate by more than 0.5°F/hr from the average rate of change over the last 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. A test computer is normally used for data acquisition and/or leakage rate calculations. The data acquisition package reads the inputs (pressure, temperature, and dewpoint temperature), converts these readings into engineering units, and stores/prints these values to be used for leakage rate calculations. The leakage rate is calculated using the absolute method of mass point analysis. The absolute method of mass point analysis consists of periodically calculating air masses within the containment structure over a period of time from pressure, temperatures, and dew point observations during the test. The air masses are computed

Revision 52Updated Online 09/30/20 SPS UFSAR 5.5-3 using the ideal gas law. The leakage rate is then determined by plotting the air mass as a function of time, using a least-squares fit to determine the slope. The leak rate is expressed as a percentage of containment air mass lost in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. A 95% confidence level is calculated using a T distribution. The sum of the leakage rate at a 95% confidence level must be less than 0.75 La.

A verification test is performed following each Type A test. This test provides a method of assuring that systematic error or bias is given adequate consideration. The verification test consists of a superimposed leakage rate equal to 75% to 125% of L a , which is measured independently from Type A test instrumentation. This air change and that which is measured by the containment leakage Type A instrumentation must agree within +/-25% La.

5.5.4 Containment Penetration Leakage Rate Test (Type B)

Type B tests measure local leakage across containment pressure boundaries that are either atypically large and/or whose design incorporates resilient seals, gaskets, or sealant compounds, and piping penetrations fitted with expansion bellows.

Type B tests, except airlocks, are performed at periodic intervals based on the historical performance at each penetration in accordance with 10 CFR 50, Appendix J, Option B, and NEI 94-01, Revision 3-A, Industry Guidelines for Implementing Performance-Based Option of 10 CFR 50, Appendix J, Option B. Air locks are tested at the frequency described in Section 5.5.6.

Type B tests are performed by either one of two methods.

The first method is to pressurize between the double o-ring seals of the cover used as the containment pressure barrier (i.e. air lock doors, electrical penetration flanges, fuel transfer blank flange). The makeup air method is used to determine the penetration leakage by applying a test pressure equal to or greater than containment design pressure (45 psig) between the o-ring seals.

The second method also uses the makeup air method when performing the full blown air lock test. The space between the inner and outer air lock door is pressurized to a test pressure equal to or greater than containment design pressure (45 psig).

The acceptance criterion for the combined leakage rate of the penetrations and valves subject to Types B and C tests shall be equal to or less than 60% of the maximum allowable leakage rate of the containment.

5.5.5 Containment Isolation Valve Leakage Rate Test (Type C)

There are two methods used in Type C tests. With either method, each valve to be tested is closed by normal operation without any preliminary exercise or adjustment.

In Method 1, the section of piping with the containment isolation valves is isolated from the remainder of the fluid system by using valves or blank flanges as necessary, and the piping is drained. The inside and outside containment isolation valves are tested individually with air at a pressure equal to or greater than containment design pressure. Test air is applied at a test connection on the inboard side (toward the inside of the containment structure) of the valve to be

Revision 52Updated Online 09/30/20 SPS UFSAR 5.5-4 tested, and the leakage air is vented through a test vent on the outboard side of the valve. A flowmeter, connected in line with the pressure source, is used to measure leakage through the containment isolation valve as a function of time. In this procedure, the test airflow is directed across the valve seat from the inside-to-the-outside containment structure direction.

In Method 2, the section of piping with the containment isolation valves is isolated from the remainder of the fluid system, using valves or blank flanges as necessary, and the piping is drained. The inside and outside containment isolation valves are tested simultaneously with air at a pressure equal to or greater than containment design pressure. Test air is applied at a test connection between the two valves, and leakage air is vented through a test vent on the outboard side of the penetration. A flowmeter connected in line with pressure source is used to measure leakage through the containment isolation valves as a function of time. The containment isolation valves are typically diaphragm or symmetric butterfly type valves. The outside containment isolation valves are tested in the outward direction. The inside containment isolation valves are tested in the reverse direction. This test is equally effective for diaphragm and symmetric butterfly valves. Penetrations 90, 91, and 103 are tested using this method.

The acceptance criterion for the combined leakage rate of the penetrations and valves subject to Type B and C tests shall be equal to or less than 60% of the maximum allowable leakage rate of the containment.

Tables 5.2-1 and 5.2-2, for Units 1 and 2 respectively, identify those valves that are required to be tested in accordance with 10 CFR 50 Appendix J, Option B to ensure containment integrity during LOCA conditions. The basis for valves which are: (1) not Type C tested; (2) Type C tested but the leakage penalty is not included in the overall Type B and C total leakage; or (3) Type C tested but the leakage penalty is not included in the overall Type A leakage is provided below:

Main Steam and Feedwater Penetrations Reason for Type C Testing Exemption 39, 40, 41, 73, 74, 75, 76, 77, 78, 87, 88, These penetrations are directly connected to the 102 steam generator secondary side and, therefore, are considered a closed system (an extension of the primary containment). In addition, the S/G remains at a pressure greater than Pa for at least the first hour and is not considered a credible leakage path from containment.

Reference:

T.S.

Amendment 72/73 dated September 29, 1981

Revision 52Updated Online 09/30/20 SPS UFSAR 5.5-5 Component Cooling Penetrations Reason for Type B & C Leakage Exclusion 1, 2, 4, 5, 9, 10, 11, 12, 13, 14, 16, 17, 18, These penetrations are a closed system.

25, 26, 27, 110 Containment penetration check valves and trip valves are leakage tested, but the leakage is not included in the 10 CFR 50 Appendix J Type B and C total leakage. During the associated check valve leakage test, the containment penetration manual isolation valve is leak tested in the reverse direction. The valve is tested with system pressure on the upstream side and the downstream side vented.

Reference:

T. S.

Amendment 72/73 dated September 29, 1981 Safety Injection Penetrations Type C Tested but Leakage Penalty not Included in Overall Type A Leakage 7, 15, 21, 23, 46, 60, 61, 62, 66, 67, 68, 69, These penetrations are water filled and/or 113 normally operating under accident conditions at a pressure greater than Pa. Therefore, these penetrations are not considered credible leakage paths from containment.

Reference:

NRC SER dated November 21, 1988.

RCP Seal Water Penetrations Not Type C Tested 35, 36, 37 Needle valves are throttled open and administratively controlled. These lines remain open after a safety injection signal and contribute to the total injection flow while cooling the RCP seals. The three incoming lines have a check valve inside containment and a local manual valve (throttle valve) outside containment, combined with both a closed system and a continuous water seal at a pressure sufficient to preclude containment atmospheric leakage.

Service Water to RSHX Penetration Reason for Type B & C Leakage Exclusion 79, 80, 81, 82, 83, 84, 85, 86 These systems are closed systems. Each train is leak tested but the leakage is not included in the 10 CFR 50 Appendix J Type B and C total leakage. The valves in these lines remain open during a design basis accident.

Reference:

T. S.

Amendment 72/73 dated September 29, 1981.

Revision 52Updated Online 09/30/20 SPS UFSAR 5.5-6 5.5.6 Scheduling and Recordkeeping of Periodic Tests Primary containment integrated leakage rate tests are conducted at periodic intervals based on historical performance of the overall containment system. Type A tests are only conducted while the plant is in the shutdown condition.

If any periodic Type A test fails to meet the acceptance criteria, the schedule for subsequent Type A test is determined in accordance with 10 CFR 50, Appendix J, Option B, and NEI 94-01, Revision 3-A, Implementation Guidelines for Implementing Performance-Based Option of 10 CFR 50, Appendix J.

Containment resilient seal penetration tests (Type B tests) were performed prior to initial criticality and are performed periodically thereafter during shutdown for refuelings, in accordance with 10 CFR 50 Appendix J, Option B, and NEI 94-01, Revision 3-A.

The personnel air lock full volume test was performed prior to initial fuel load and is performed periodically thereafter, in accordance with 10 CFR 50 Appendix J, Option B and NEI 94-01. If the air lock is open during periods when containment integrity is not required, the lock is tested only at the end of those periods. If the air lock is opened when containment integrity is required, the air lock is tested within 7 days after such opening. If the air lock door is opened more frequently then once every 7 days, the air lock seals are tested at least once every 30 days during the period of frequent opening. The personnel air lock and personnel escape hatch have testable seals and testing of the seals fulfills the 10 CFR 50 Appendix J after each use requirement. Seal tests are not substituted for the air lock full volume test.

Containment isolation valve testing (Type C tests) was performed prior to initial criticality and is performed periodically thereafter during each reactor shutdown for refueling, in accordance with 10 CFR 50 Appendix J, Option B.

The results of periodic Type A, B, and C tests are documented to show that performance criteria have been met. The comparison to previous results of the performance of the overall containment and of individual components are documented to provide a basis for the established test intervals for the containment and individual components.

5.5.7 Special Testing Requirements Type A, B, and C tests, as applicable, are conducted following containment structure modifications in accordance with 10 CFR 50 Appendix J, Option B, and NEI 94-01, Revision 3-A, Implementation Guidelines for Implementing Performance-Based Option of 10 CFR 50, Appendix J.