ML16256A244
| ML16256A244 | |
| Person / Time | |
|---|---|
| Site: | Waterford  |
| Issue date: | 08/25/2016 |
| From: | Entergy Operations |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML16256A115 | List:
|
| References | |
| W3F1-2016-0053 | |
| Download: ML16256A244 (46) | |
Text
WSES-FSAR-UNIT-35.4-15.4COMPONENT AND SUBSYSTEM DESIGN5.4.1REACTOR COOLANT PUMPS5.4.1.1Design BasesIn addition to the design transients in Subsection 3.9.1.1, the reactor coolant pumps are designed to:
a)Withstand 4000 cycles (pressure) of reactor coolant pump startup and stopping withoutexceeding code allowable stress limits. This transient is classified as a normal condition event.b)Provide sufficient moment of inertia to reduce the flow decay through the core uponloss of pump power, ensuring that fuel design limits are not exceeded.c)Prevent reverse rotation of the pump upon loss of pump power with the other pumpsoperating.d)Operate without cooling water for periods up to three minutes without incurring sealdamage.5.4.1.2DescriptionThe reactor coolant is circulated by four vertical, single bottom suction, horizontal discharge centrifugal,motor driven pumps. (See Figures 5.4-1 and 5.4-2). The piping and instrumentation diagram for the reactor coolant pump is shown in Figure 5.1-3. Design parameters for the pumps are listed in Table 5.4-1.The predicted pump performance curve is shown in Figure 5.4-3.5.4.1.2.1Reactor Coolant Pump Assembly The reactor coolant pump assembly consists of the following:
a)Pump Case b)Rotating Assembly(Containing the impeller which is welded to the shaft)c)Mechanical Seal Cartridge Assembly d)Motor Mount e)Motor Assembly5.4.1.2.1.1Pump Case and Motor MountThe reactor coolant pump motor is connected to and supported by the pump case through the motormount. There are two openings on opposite sides of the motor mounts that provide access for assembly of the flanged rigid coupling between the motor and pump and for seal WSES-FSAR-UNIT-3 5.4-2 Revision 304 (06/10) cartridge replacement. The pump case includes the seal heat exchanger, which cools the seal cartridge assembly.
5.4.1.2.1.2 Rotating Assembly
The pump rotating assembly consists of the impeller, a hydrostatic radial bearing, and the rotating
elements of the seal cartridge assemb ly. The hydrostatic radial bearing maintains radial alignment of the shaft and is located just above the impeller. The upper radial bearings and an axial thrust bearing are located in the motor.
5.4.1.2.1.3 Mechanical S eal Cartridge Assembly (EC-6256, R302; EC-18520, R304)
The seal cartridge consists of four multiface type me chanical seals; three full-pressure seals mounted in series, and a fourth low pressure seal designed to withstand system operating pressure when the pump is not operating. A controlled bleedoff flow through t he seals is used to cool the seals and to equalize the pressure drop across each seal. The seals are capabl e of operation without cooling water for up to ten minutes without incurring damage. The controlled bleedoff flow is collected in the volume control tank of the Chemical and Volume Control System (CVCS). Leakage past the vapor seal is routed to the
Containment sump via the Floor Drain System. (EC-6256, R302; EC-18520, R304)
The seal cartridge concept reduces the time requir ed for seal maintenance, thereby lowering personnel radiation exposure. The seal cartridge can be remov ed without draining the pump case. Details of the seal are shown in Figure 5.4-2.
5.4.1.2.1.4 Motor Assembly
The motor assembly includes the following:
a) Air Cooler
b) Motor Bearing Lubrication
c) Oil Lift Pumps
d) Motor Shaft
e) Upper and Lower Radial Guide Bearings
f) Axial Thrust Bearing
g) Flywheel
h) Anti-Reverse Rotation Device
i) Motor
Cooling water for all motor heat exchangers is supp lied from the Component Cooling Water System. Heat exchangers supplied include the upper bearing lube oil cooler, lower bearing oil cooler and motor air
cooler.
WSES-FSAR-UNIT-3 5.4-3 Revision 307 (07/13)
(EC-8458, R307)
The oil lubrication system consists of two high pressure oil lift pumps with attached low pressure oil
pumps, a lube oil cooler, motor lower bearing oil c ooler, and slinger assembly all within the motor housing. During normal pump operation, the coil slinger maintains oil flow through the motor upper bearings. During pump startups and shutdowns, both the selected high pressure oil lift pump and its
associated low pressure oil pumps are manually started to maintain sufficient oil pressure to the bearing and anti-reverse rotation device. If the oil pumps have not been started manually when the motor is stopped, the selected oil pump starts automatically, on a reactor coolant pump breaker trip. After the reactor coolant pump is at rated speed or has stopped, the operator must stop the lube oil pumps manually (see Subsection 5.4.1.5.2.3). (EC-8458, R307)
The motor bearing support system consists of a Kingsbury double acting thrust bearing and two radial
bearings located above and below the motor rotor.
(EC-8458, R307)
To improve pump coastdown characteristics in order to meet system requirements during a loss of pump power condition, the flywheel and motor-pump rotating a ssembly has a total moment of inertia of 106,500 lbm-ft 2. (EC-8458, R307)
The reactor coolant pumps are designed with an anti-re verse rotation device (see Figure 5.4-4) to prevent reverse rotation of the pump. This device prevent s reverse rotation after the pump stops from normal speed (1183 rpm) while the other pumps continue to operate.
The pump assembly is sized for continuous operati on while pumping primary coolant with a specific gravity of 1.0. The motor is designed for 500 heatup cycles, during which the horsepower demand decreases from 9700 to 7800 over a period of approxim ately seven hours. The motors are also designed to start and accelerate to speed under full load at 80 percent of rated voltage. The pump and motor
assembly is designed to operate in a radiation environment of 10.5 x 10 6 rads (integrated) over a period of 40 years with nominal maintenance. Section 12.3 contains a further discussion of radiological considerations for these pumps.
5.4.1.3 Evaluation (EC-38245, R307)
The reactor coolant pumps are sized to deliver flow that equals or exc eeds the design flow rate used in the thermal hydraulic analysis of the reactor coolant system. Anal ysis of steady state and anticipated transients is performed assuming the minimum design flow rate. Tests were performed to evaluate reactor coolant pump performance during the post core load hot functional testing to verify adequate flow. (EC-38245, R307)
(DRN 03-2059, R14; EC-8458, R307) The actual nominal RCS flowrate is approximately equal to 110% of the original design value of 99,000
gpm per pump. This results in a mass flowrate of approximately 163.5 Mlb/hr under normal operating conditions. (DRN 03-2059, R14; EC-8458, R307) (EC-30077, R 305; EC-38245, R307)
Leakage from the pump past the pump shaft is controlled by the shaft seal assembly. Coolant entering the seal chambers is cooled and collected in clos ed systems so that reactor coolant leakage to containment is essentially zero. The normal vapor stage leakoff is routed to the Containment Sump. In the event of a seal malfunction, inst rumentation in the form of pressure transducers, a flow meter, and a temperature detector are provided to al ert the operator to a potential problem. (EC-30077, R 305; EC-38245, R307)
(DRN 05-150, R14)
Component cooling water to the reactor coolant pumps is not required to ensure (1) the integrity of the reactor coolant pressure boundary, (2) the capability to shutdown the reactor and maintain it in a safe shutdown condition, or (3) the capability to prevent or mitigate the consequences of accidents that could result in potential offsite exposures comparable to the guideline exposures of 10CFR50.67. Low cooling flow to each pump is alarmed (DRN 05-150, R14)
WSES-FSAR-UNIT-3 5.4-4 Revision 307 (07/13) in the main control room. Upon sounding of four alarms showing that flow is not available to all pumps, the operator trips the reactor and then trips the pumps allowing the system to be cooled down by natural circulation flow. (DRN 00-1059, R11-A; EC-38245, R307)
The reactor coolant pumps, by design and field experience, are not susceptible to a loss-of-coolant accident (LOCA) due to seal failure resulting from loss of seal cooling water. The pumps are equipped with four series-arranged face seals, a ll of which are designed for 2500 psid. The P across any one of the three main seals during normal operation is 750 psi.
The loss of any single seal would result in a P of approximately 1100 psi. A seal leakage chamber structurally designed for 2500 psia is provided to collect controlled seal leakage and conduct it to a closed system. The fourth face seal is provided as an integral part of the seal leakage chamber to prev ent liquid or gaseous leakage from escaping to the atmosphere. The seal is designed to operate norma lly against a backpressure of 25 to 250 psia and is capable of holding against 2500 psia in the static c ondition and during coastdown following failure of the three series-arranged main seals. When holding agains t 2500 psia in the static condition, seal leakage should not exceed the normal operating seal leakage. (DRN 00-1059, R11-A)
Four reactor coolant pumps with seals of similar design have been operated for up to 40 minutes with no
component cooling water flow. While there was some increase in controlled seal leakage (to the closed system), the mechanical seals were subsequently dismantled and refurbished without finding major damage. Therefore, a loss of component cooling wate r flow to the reactor coolant pumps for up to 40 minutes is not expected to result in a complete seal failure. (EC-38245, R307)
The component cooling water supply to the RCP's is isolated upon a Containment Spray Actuation Signal (CSAS). The RCP's are then stopped, resulting in no requirement for component cooling water. The CSAS is provided by a high-high containment pressu re signal. Such a signal can be produced by only two accidents: (1) a LOCA, or (2) a major steam line break inside containment. In either case, the Plant Protective System trips the reac tor and the operators trip the pumps.
In the event of a break in the reactor coolant pum p suction piping, a high reverse flow through the pump is prevented by the anti-reverse rota tion device, as described in Subsecti on 5.4.1.2.1.4. In the event of a discharge line break, increased flow through the pum p tends to accelerate the pump impeller and flywheel in the forward direction. A detailed evaluati on of this incident relating to the integrity of the flywheel is presented in Subsection 5.4.1.4.
5.4.1.4 Reactor Coolant Pump Flywheel Integrity
The design, fabrication and testing of the reactor c oolant pump flywheel in relation to recommendations of Regulatory Guide 1.14 (formerly Safety Guide 14) and the results of an analysis of pump and flywheel performance during a LOCA are discussed in this section.
5.4.1.4.1 Regulatory Guide 1.14, Rev. 00 (10/27/71)
The following design conditions and material specifications for the flywheels are consistent with the
recommendations of Regulatory Guide 1.14, Rev.
00, "Reactor Coolant Pump Flywheel Integrity".
WSES-FSAR-UNIT-3 5.4-5 Revision 14-A (03/06)5.4.1.4.1.1 Flywheel Material Specifications The material used to manufacture the flywheel is pressure vessel quality, prepared by the vacuum melt and degassing process, ASTM-A-543 Grade B, Class I steel plate. Pump flywheel material is prepared
and tested such that: a) The nil-ductility transition temperature of the flywheel is no greater than +10F, as obtained from drop-weight tests performed in accordance with specification ASTM-E-208. b) The Cv upper shelf energy level, in the Wr direction, as obtained per ASTM-A-370, is no less than 50 ft lb. A minimum of three Cv specimens are tested from each plate. (DRN 00-1059, R11-A; 05-1392, R14-A)c) The minimum fracture toughness of the material at the normal operating temperature of the flywheel is equivalent to a dynamic stress intensity factor (K IC dynamic) and is equal to or greater than 100 Ksi in. Compliance is demonstrated by one of three methods listed in NRC Regulatory Guide 1.14, Rev. 00, paragraph C.1.c. (DRN 00-1059, R11-A; 05-1392, R14-A)d) Each finished flywheel is subjected to a 100 percent volumetric, ultrasonic examination from the flat surface using ASME Section III, Class I criteria. This inspection is performed on the flywheel after final machining and overspeed test. e) After flame cutting, at least 1/2 in. of stock is left on the outer and bore radii for machining to final dimensions. f) Finished machined bores, keyways and drilled holes in the flywheel are subjected to amagnetic particle or liquid penetrant examination before final assembly. (DRN 03-2059, R14) 5.4.1.4.1.2 Flywheel Design Criteria The flywheels are designed to withstand normal operating conditions, anticipated transients, and the original design basis LOCA 1 combined with the safe shutdown earthquake. Thus the following criteria are met: (DRN 03-2059, R14) a) The combined primary stresses at normal operating speed do not exceed one-third of the minimum specified yield strength or one-third of the measured yield strength in the weak
direction of the material. b) The combined primary stresses at design overspeed do not exceed two thirds of the minimum specified yield strength or two-thirds of the measured yield strength in the weak direction of the material. Design overspeed is defined as 125 percent of normal operating speed.
See Subsection 5.4.1.4.2 for further discussion of pump overspeed following a discharge line
break.Each flywheel is tested for one minute at the design overspeed of 1500 rpm.
The flywheel is accessible for 100 percent in-place volumetric ultrasonic inspection. The motor assembly is designed to allow such inspection with a minimum of motor disassembly. (DRN 03-2059, R14) 1The effects of the original design basis LOCA (i.e., MCLBs, which have been eliminated via LBB, see Section 3.6.3) envelop the effects of the next limiting set of pipe breaks (BLPBs) on RCS flywheel design. (DRN 03-2059, R14)
WSES-FSAR-UNIT-3 5.4-6 Revision 14 (12/05
)5.4.1.4.2 Additional Data and Analysis (DRN 03-2059, R14) As mentioned previously, a LOCA in the pump discharge leg of the reactor coolant piping tends to accelerate the pump in the forward direction during blowdown. A maximum speed of 1585 rpm has been calculated using the CE FLASH computer program, assuming a mechanistic break 1 located at the pump discharge with a simultaneous loss of electrical power to the pump motor. This is about 134 percent of normal operating speed. Even at pump speeds greater than 1585 rpm, no other failures of sufficient magnitude to increase the consequences of the initiating LOCA can be postulated except by assuming failure of the flywheel. Stress calculations have demonstrated that even at speeds up to 270 percent of normal operating speed with a sharp edged crack located to the base of the keyway, failure cannot be predicted. It is therefore concluded that adequate margin has been provided to preclude the development of a flywheel rupture by reactor coolant pump overspeed. (DRN 03-2059, R14) 5.4.1.5 Reactor Coolant Pump InstrumentationThe reactor coolant pumps and motors are equipped with the instrumentation necessary for proper operation and to warn of incipient failures. A description of the major channels follows. See Figure 5.1-3.
Measurement channels are typical for each reactor coolant pump.
5.4.1.5.1 Temperature 5.4.1.5.1.1 Motor Stator Temperature Each reactor coolant pump motor is provided with six thermocouples embedded in the stator windings.Indication of stator temperature is provided by the plant monitoring computer. During initial pump testing, the highest reading thermocouple will be selected for this temperature measurement channel. Should
stator temperature exceed a predetermined limit, a high temperature alarm will be annunciated by the plant monitoring computer. High temperature is detrimental to motor winding insulation life, and may be caused by high ambient temperature, reduction in the cooling airflow to the stator, or inadequate time delay
between successive starts of the motor. 5.4.1.5.1.2 Motor Bearing Temperatures High temperature alarms for the oil lubricated bearings are annunciated by the plant monitoring computer at the main control board. These include: a) Motor upper guide bearing b) Upper and lower thrust bearings c) Motor lower guide bearing High temperature is indicative of bearing or oil supply problems. The motor thrust bearing temperature is continuously monitored at the main control board. Temperatures of the motor guide bearings are fed to the
plant monitoring computer. (DRN 03-2059, R14) 1The effects of the original design basis LOCA (i.e., MCLBs, which have been eliminated via LBB, see Section 3.6.3) envelop the effects of the next limiting set of pipe breaks (BLPBs) on RCS flywheel design. (DRN 03-2059, R14)
WSES-FSAR-UNIT-35.4-7Revision 8 (5/96)5.4.1.5.1.3Pump Controlled Bleedoff TemperatureThe temperature of the controlled bleedoff flow is displayed in the main control room. An alarm signal isprovided at the main control panel should the controlled bleedoff temperature exceed a high limit. A high temperature condition is an indication that the seal assembly or seal water cooler is not operating properly.5.4.1.5.1.4Lube Oil Cooler Inlet TemperatureThe inlet temperature of the lube oil cooler is alarmed by the plant monitoring computer.5.4.1.5.1.5Lube Oil Cooler Outlet TemperatureThe outlet temperature of the lube oil cooler is alarmed by the plant monitoring computer.5.4.1.5.1.6Anti-Reverse Rotation Device Bearing TemperatureIndication of the anti-reverse rotation device temperature is provided by the plant monitoring computer. Ahigh temperature condition in the anti-reverse rotation device is alarmed by the plant monitoring computerand is an indication of lubrication problems.5.4.1.5.1.7Seal Water Cooler Component Cooling Water Outlet TemperatureThe temperature of component cooling water is monitored at the seal water cooler outlet of each ReactorCoolant Pump (1A, 1B, 2A, 2B) with respective temperature elements. The temperature parameter is displayed in the main control room.Two independent high temperature setpoints are used to annunciate an alarm in the main control roomand to initiate an automatic isolation of the affected seal cooler by closing the component coolant water inlet and outlet valves. The high temperature alarm is an indication that leakage from the primary loop to the component coolant water has developed inside the seal water cooler or that the cooling water flow has decreased.To countermand an isolation of seal coolers on a spurious signal, a control switch is provided in the maincontrol room to allow the operator to override the high temperature signal and to reopen the component coolant water inlet and outlet valves to purge the seal water cooler for a preset time span. In that case, if the high temperature signal persists, the seal water cooler will be again automatically isolated at the end ofthe preset time span.5.4.1.5.2Pressure5.4.1.5.2.1Pump Seal PressuresThe middle, upper and vapor seal cavities in each pump are provided with pressure elements thatgenerate a signal proportional to the pressure within the cavity. The pressure in the seals is displayed in the main control room and are monitored by the plant monitoring computer. High and low pressure alarms for the middle and upper seal cavities and a high pressure alarm for the vapor seal cavity are annunciated by the plant monitoring computer.Abnormally high or low pressure is an indication of seal malfunction.
WSES-FSAR-UNIT-35.4-8Revision 8 (5/96)5.4.1.5.2.2Lube Oil Cooler PressureA common low pressure alarm at the inlet to the lube oil cooler and anti-reverse rotation device isannunciated in the main control room.5.4.1.5.2.3High Pressure Oil Lift Pump Discharge Pressures Pressure switches at each high pressure oil lift pump discharge actuate an alarm on low pressure in themain control room. In the event of a failure of one of the oil lift pumps, the second oil lift pump must be started. A separate measurement channel provides a control signal to the respective reactor coolant pump circuit, which prevents the starting of the reactor coolant pump if insufficient oil lift pressure exists.
Another separate measurement channel provides local indication of pressure in the oil lift pump dischargeheader.5.4.1.5.2.4High Pressure Oil Lift Pump and Low Pressure Oil Pump FiltersFilters are used during motor start-up and shutdown at the suction and discharges of the high pressure oillift and low pressure oil pumps. A high differential pressure across the filter indicates a clogged filter and is alarmed by the plant monitoring computer.5.4.1.5.2.5Reactor Coolant Pump Differential PressureTwo independent differential pressure transmitters are provided on each reactor coolant pump. Thedifferential pressure signal is indicated in the main control room. A calibration curve is used to relate pump differential pressure to pump flow.5.4.1.5.2.6Lube Oil Pressure to Anti-Rotation DeviceLocal pressure gages are provided in the oil line to the anti-rotation device for testing purposes only.
5.4.1.5.2.7Reverse Rotation Indicator SwitchA differential pressure switch in the lube oil system mounted near the upper bearing bracket provides anindication that the reactor coolant pump motor is turning in the reverse direction. This switch alarms via the plant monitoring computer.5.4.1.5.2.8Seal Water Cooler Component Cooling Water Outlet PressureThe component cooling water pressure is monitored at each of the reactor coolant pumps (RCP 1A, 1B,2A, and 2B) seal water outlet. A high pressure signal annunciates an alarm in the main control room.
WSES-FSAR-UNIT-35.4-95.4.1.5.3Flow5.4.1.5.3.1Pump Controlled Bleedoff Flow A flowmeter is used to measure the controlled bleedoff flow from the bleedoff seal cavity to the CVCS.This instrument provides an indication of the flow rate and annunciates high and low flow alarms via the plant monitoring computer. Abnormal flow is an indication of improper seal performance.5.4.1.5.3.2Motor Circulating Oil System FlowA lube oil flow switch is provided at the inlet to the lube oil cooler. Should the lube oil flow to the cooler fallbelow a predetermined setpoint, a low flow alarm is actuated via the plant monitoring computer.5.4.1.5.3.3Motor Low Pressure Oil Pump FlowThe lube oil flow to the anti-reverse rotation device is sensed by a lube oil flow switch. This switchactuates a low flow alarm via the plant monitoring computer when flow drops below the predeterminedsetpoint.5.4.1.5.3.4Motor High Pressure Oil Lift Pump FlowFlow in the discharge header of the two high pressure oil lift pumps is sensed by a lube oil flow switch.This switch actuates a low flow alarm via the plant monitoring computer when flow drops below the predetermined setpoint.5.4.1.5.4LevelA level sensing transmitter in each oil reservoir transmits a signal for level indication and high and lowalarms via the plant monitoring computer.5.4.1.5.5VibrationMotor vibration is sensed by a vibration switch attached to the pump motor casing. Excessive vibration isalarmed by the plant monitoring computer.Review of Generic Letter 89-15 prompted the addition of new RC pump and motor monitoring.Though not required by that letter, additional "monitoring" devices have been added to correlate data to provide input and trending information on the status of motor, shaft, bearings and seals. The addition consists of x and y direction radial proximity probes and one axial thrust probe at the motor thrust bearing, x and y direction radial proximity probes at the bottom of the motor and x and y direction radial proximityprobes at the pump shaft. The output of these devices does not interact with any plant controls, but provides additional "monitoring" capability.Review of Generic Letter 89-15 also prompted the addition of two (2) keyphasors at the RC pump couplingspacer. These provide additional monitoring of shaft rotative speed and vibration phase angle. These devices do not interact with any plant controls, but provide additional "monitoring" capability.
WSES-FSAR-UNIT-3 5.4-10 Revision 307 (07/13) 5.4.1.5.6 Speed
5.4.1.5.6.3 Reactor Cool ant Pump Speed Sensors
Reactor coolant pump shaft speed is transmitted to the Core Operating Limit Supervisory System (COLSS) and the Plant Protective System (PPS) for continuous monitoring and protection. See Sections 7.2 and 7.7 for a further description.
5.4.1.6 Tests and Inspections (DRN 00-1059, R11-A)
The reactor coolant pump pressure boundary is nondestr uctively inspected as required by the ASME Code,Section III (see Table 5.2-1) for Class 1 components. The pump casing inspections include complete radiography and liquid penetrant testing.
The pump casing, cover and seal cooling heat exchanger are subjected to a hydrostatic test at the manufacturing facility. The pump is hydrostatically pressure tested along with the RCS. In-service inspection will be performed during plant life in accordance with the ASME Code,Section XI as discussed in Subsection 5.2.4. (DRN 00-1059, R11-A)
All rotating parts of the pump are statically and dy namically balanced in two planes. Where possible, balancing is done for the entire assembly.
The pump assembly is performance tested in the vendor's shop in accordance with the Standards of the
Hydraulic Institute to verify hydr aulic performance, as well as the ab ility of the pumps to function as required by the specifications. The vibration levels are monitored during this test. Evidence of the pumps operating near a critical speed would be noted as excessive vibration.
Full scale seal testing is performed at rated pressure, temperature, water chemistry, and speed to
demonstrate the capability of the seals to sati sfactorily perform their design function.
5.4.2 STEAM GENERATORS
5.4.2.1 Design Basis (DRN 03-2059, R14; 06-1060, R15; EC-8458, R307)
Replacement Steam Generators, implemented in Cy cle 18, are designed to transfer 3735 MWt from the RCS to the secondary system with 0% t ube plugging, producing approximately 16.6 x 10 6 lbm/hr of 837.2 psia saturated steam. At 10% tube pluggi ng, the RSGs produce approximately 16.6 x 10 6 lbm/hr of 818.0 psia saturated steam. Moisture separators and st eam driers in the shell side of the steam (DRN 03-2059, R14; 06-1060, R15; EC-8458, R307)
WSES-FSAR-UNIT-3 5.4-11 Revision 307 (07/13)
(DRN 03-2059, R14; EC-8458, R307) generator limit the moisture content of the steam to 0.10% during normal operation at full power. The steam generator design parameters are listed in Tabl e 5.4-2. The steam generators, including the tubes, are designed for the RCS transients listed in Subsecti on 3.9.1.1 so that the c ode allowable stress limits are not exceeded for the specified number of cycles. All transients have been established based on
conservative assumptions of oper ating conditions in considerati on of supportive system design capabilities. The steam generators are capable of sustaining the following additional design transients
without exceeding code a llowable stress limits: (DRN 03-2059, R14)
a) Ten secondary side hydrostatic tests with secondary side pressurized to 1-1/4 times the design pressure and the primary side pressuriz ed so that the tube differential pressure does not exceed 1375 psia psi (test condition).
b) Secondary side tube leak test Cycles Shell Side Pressure (PSIG) Shell Side Temperature (°F) 400 200 60-250 200 400 60-250 120 600 60-250 80 840 60-250
c) 5.700 cycles of adding feedwater to the steam generators through the main feedwater nozzle when in hot standby conditions (normal operations
). The following conditions and cycles are based on actual operating conditions: Cycles Feedwater temp. (°F)
Flow change (gpm) Rate for flow change 600 40 50 to 1100 1 second 948 40 0 to 450 1 second 1302 40 0 to 50 70 seconds 600 70 50 to 1100 1 second 948 70 0 to 450 1 second 1302 70 0 to 50 70 seconds (DRN 00-1059, R11-A) d) 600 cycles of feeding during a three hour low power transient. Cycles Power change (%) Feedwater temp change (°F) Flow change (lb/hr) 300 0 to 15 40 to 260 25,000 to 1.04x10 6 300 15 to 0 260 to 40 1.04x10 6 to 25,000 (DRN 00-1059, R11-A) e) Four thousand pressure transients of 113 psi across the primary divider plate in either direction caused by starting and stopping reactor coolant pumps (normal condition).
f) Twenty cycles of a Loss of Main Feedwater Transient.
g) One cycle for a Design Bases Earthquake transient with a Steam Line Break.
h) One cycle for a Design Bases Earthquake transient with a Feedwater Line Break. (EC-8458, R307)
The operating pressure and temperature limits for t he steam generator primary side were determined in accordance with the ASME Code,Section III, Appendix G. (EC-8458, R307)
The method of fastening tubes to the tube sheet confor m with the requirements of Sections III and IX of the ASME Code. Tube expansion into the tube sheet is total with no voids or crevices occurring along the length of the tube in the tube sheet. Eight tube support plates of the flat-contact broached trifoil tube hole design are used in the RSG. The broached tube support plate is designed to reduce the tube-to-
tube support plate crevice area while providing for ma ximum steam/water flow in the open areas adjacent to the tube. Flat tube contact geometry in the RSG provides additional dryout margin. (EC-8458, R307)
WSES-FSAR-UNIT-3 5.4-12 Revision 307 (07/13)
(EC-8458, R307) The steam generator is designed to ensure that critical vibration frequencies are well out of the range expected during normal operation and during abnormal conditions. The tubing and tubing supports are designed and fabricated with considerations given to both secondary side flow induced vibration and
reactor coolant pump induced vibrations. In additi on, the steam generator assemblies are designed to withstand the blowdown forces resulting from severance at the steam nozzle. (EC-8458, R307)
Discussion of the techniques used to maintain cleanliness during final assembly and shipment are
discussed in Subsection 5.2.3.
Onsite cleaning and cleanliness control procedures for the steam generator were consistent with the recommendations of Regulatory Guide 1.37, "Qualit y Assurance Requirements for Cleaning of Fluid Systems and Associated Components of Water-Cooled Nuclear Power Plants", (3/16/73) and ANSI N45.2-1-1973, "Cleaning of Fluid Systems and A ssociated Components For Nuclear Power Plants".
5.4.2.2 Description (DRN 03-2059, R14; EC-8458, R307)
The steam generator is illustrated in Figure 5.4-5. Mo isture-separating equipment in the shell side of the steam generators limits moisture content of the ex it steam to a maximum of 0.10%. Manways and handholes are provided for access to the steam generator internals. (DRN 03-2059, R14; EC-8458, R307)
Reactor coolant enters at the bottom of each steam generator through the single inlet nozzle, flows
upward through the U-tubes, and leaves through the two out let nozzles. A vertical divider plate separates the inlet and outlet plenums in the lower head.
(DRN 02-674, R12;03-130, R14;06-290, R 14-B; EC-8458, R307)
Feedwater enters the steam generator through the feedwater nozzle where it is distributed via a
feedwater distribution ring. The feedwater ring is constructed wi th eccentric reducer fittings and perforated discharge nozzles which discharge at the t op of the feedring to avoid regions where steam pockets can be formed. The feedwater ring is locat ed above the elevation of t he feed nozzle to minimize the time required to fill the feed nozzle during a cold water addition transient. (DRN 02-674, R12;03-130, R14;06-290, R 14-B; EC-8458, R307)
The downcomer in the steam generator is an annular passage formed by the inner surface of the steam
generator shell and the cylindrical shell that encloses the vertical U-tubes. Upon exiting from the bottom of the downcomer, the secondary flow is directed upward over the vertical U-tubes. Heat transferred from the primary side converts a portion of the secondary flow into steam.
(EC-8458, R307)
Upon leaving the vertical U-tube heat transfer surfac e, the steam-water mixtur e enters the centrifugal-type separators. These impart a centrifugal motion to the mixture and separate the water particles from the steam. The water exits from the perforated s eparator housing and combines with the feedwater to repeat the cycle. Final drying of the steam is accomplished by passage of the steam through the single-tier banked dryers. Drain pipes located at the ends of each secondary separator dryer bank carry captured water downward into the recirculation pool. The pressure drop from the steam generator feedwater nozzles to the steam outlet nozzle is appr oximately 31.6 psi when operating at full power. The steam generator supports are described in Subsection 5.4.14. Secondary side overpressure protection is provided by 12 spring-loaded ASME Code safety valves mounted on the main steam lines as described in Subsection 5.4.13.
The RSGs have a 3.00 in NPS blowdown nozzle connected to the steam generator secondary water
through a series of holes in the tubesheet secondar y surface that intersect an internal blowdown passage. (EC-8458, R307)
WSES-FSAR-UNIT-3 5.4-13 Revision 307 (07/13)
(EC-8458, R307)
The RSGs contain an integral flow limiting device, cons isting of seven 8.53 inch I.D. venturis installed in the steam outlet nozzle. This flow restrictor r educes energy release to containment and loads on the steam generator internals during a postulated steamline break. (EC-8458, R307)
5.4.2.3 Evaluation
5.4.2.3.1 Steam Generator Tubes
5.4.2.3.1.1 Chemistry Compatibility (EC-8458, R307)
The steam generators, tubed with Alloy 690 TT, 0.75 in. OD by 0.044 for Rows 1&2 and 0.043 for Rows 3 to 138 wall tubing, incorporates a general corrosion allowance that will provide for reliable operation over the plant design lifetime. (EC-8458, R307)
Localized corrosion has led to steam generator tube l eakage in some operating plants. Examination of tube defects that have resulted in leakage has shown that two mechanisms are primarily responsible.
These localized corrosion mechanisms are referred to as (1) stress assisted caustic cracking, and (2) wastage or beavering. Both of these types of co rrosion have been related to steam generators that have operated on phosphate chemistry. The caustic stress co rrosion type of failure is precluded by controlling bulk water chemistry to the specification limits show n in Subsection 10.3.5. Removal of solids from the secondary side of the steam generator is discussed in Subsection 10.4.8.
Localized wastage has been eliminated by removing phosphates from the chemistry control program.
Volatile chemistry (discussed in Subsection 10.3.5) has been successfully used in all C-E steam
generators that have gone into operation since 1972.
5.4.2.3.1.2 Mechanical Considerations
Because the reactor coolant pumps have a rotati onal speed of 1180 rpm, the imposition of exciting frequencies of 19 to 20 Hz and 95 to 100 Hz was cons idered in the design. The low frequency range is defined as a mechanical vibration resulting from the transmission of a mechanical impulse at the frequency of pump rotation. The upper frequency range is defined as a sinusoidal pressure vibration of +
six psi in the reactor coolant piping that contains the pump. The pressure variation results from the impeller vanes interacting with the cut-water vane at the volute outlet during each revolution of the impeller.
5.4.2.3.1.3 Hydraulic Stab ility of Feedwater System
Pressure pulses have been observed in some feedwater lines which were initiated by steam-water interaction causing ripple formation at a steam-water in terface in the feedwater piping. This resulted in the formation of a water slug whic h isolated the steam in the feedwat er line. As the isolated steam bubble condensed, the decreasing pressure would accelerate the water slug in the direction of the void.
The kinetic energy in the slug would increase unt il the steam bubble collapsed. When the water slug impacted with the water filling the upstream side of the pipe, pressure pulses were generated. With a design incorporating only a small amount of entr apped steam, the steam bubble will collapse before the water slug can gain significant kinetic energy, and the intensity of the pressure pulses is reduced to
negligible levels.
WSES-FSAR-UNIT-3 5.4-14 Revision 307 (07/13)
(EC-18652, R304)
In Waterford 3, the potential for a significant water hammer transient in the Feedwater System is reduced by the construction of the feedwater ring, described in Subsection 5.4.2.
- 2. Additional assurance that the feedwater ring will remain full of water is provided by automatic initiation of the Emergency Feedwater System (see Section 7.3). In addition, a downwar d sloping elbow at the steam generator feedwater nozzle connects to the feedwater piping. This design minimizes the drainable volume of the feedwater
piping. Hence, when the feedwater ring and nozzle ar e drained and steam enters this region, the surface area of subcooled water exposed to saturated steam is minimized. As a result, only a small amount of steam can be trapped in the elbow and any water sl ug picked up during steam-water interaction is incapable of gaining significant kinetic energy before the steam bubble is collapsed. (EC-18652, R304)
The effectiveness of the similar modifications has already been demonstrated at Indian Point 2, Trojan and St. Lucie I. Tests at a European plant (Doel I in Be lgium) have also confirmed the effectiveness of a short horizontal run in reducing wa terhammer to negligible levels.
(EC-18652, R304)
The feedring is subject to damage from pressure transients that can occur when refilling a steam generator whose water level has fallen below the feedri ng. Removal of the thermal liner o-ring and clamp on both steam generators and the acceptance of missing vent piping on Steam Generator #1 allows the feedring to drain after a loss of feedwater. These changes were evaluated on the basis of the Steam
Generators remaining capable of performing t heir safety function with a damaged feedring. (EC-18652, R304) 5.4.2.3.1.4 Tube Wall Thinning (EC-8458, R307)
Tube wall thinning acceptance criteria is specified in the Technical Specifications. The Replacement Steam Generator tube structural integrity analysis is provided in WCAP-17263-P (Reference 1). (EC-8458, R307)
5.4.2.3.2 Potential E ffects of Tube Rupture (DRN 01-1283, R12)
The steam generator tube rupture accident is a penetra tion of the barrier between the RCS and the Main Steam System. The integrity of this barrier is significant from the standpoint of radiological safety in that a
leaking steam generator tube allows the transfer of reactor coolant into the Main Steam System.
Radioactivity contained in the reactor coolant would mi x with water in the shell side of the affected steam generator. This radioactivity woul d be transported by steam to the turbine and then to the condenser or directly to the condenser via the Steam Bypass Sy stem. Non-condensible r adioactive gases in the condenser are removed by the Main Condenser Evac uation System and discharged to the atmosphere.
Analysis of a steam generator tube rupture acciden t, assuming complete severance of a tube, is presented in Section 15.6. (DRN 01-1283, R12)
Experience with nuclear steam generators indicates t hat the probability of complete severance of a tube is remote. The material used to fabricate the vert ical U-tube is a Ni-Cr-Fe alloy. A double-ended rupture has never occurred in a steam generator of this design.
The more probable modes of failure, which result in smaller penetrations, are those involving the occurrence of pinholes or small cracks in the tubes, and of
cracks in the seal welds between the tubes and tube sheet. Detection and control of steam generator tube leakage is described in Subsection 5.2.5.
5.4.2.3.3 Composition of Secondary Fluid (DRN 00-1059, R11-A;02-218, R11-A)
The concentration of radioactivity in the sec ondary side of the steam generators is dependent upon the concentration of radionuclides in t he reactor coolant, the primary-to-s econdary leak rate, and the rate of steam generator blowdown. The expected specific activities in the secondary side of the steam generators during periods of normal operation are gi ven in Section 11.1. Activities are based on operations with 0.1 percent failed fuel cladding, a total primary-to-secondary leakage of 100 lbm./day, and 60 gpm blowdown (DRN 00-1059, R11-A;02-218, R11-A)
WSES-FSAR-UNIT-3 5.4-15 Revision 307 (07/13)
(DRN 00-1059, R11-A;02-218, R11-A) rate to the Steam Generator Blowdown System. An ev aluation of the shell side radioactivity is presented in Section 11.2. Limits for radioactivity levels in the secondary side of the steam generators and the bases for these limits are provided in Technical Specifications, Chapter 16. (DRN 00-1059, R11-A;02-218, R11-A)
The recirculation water within the steam generators will contain volatile additives necessary for proper chemistry control. These and other chemistry considerations of the Main Steam System are discussed in
Subsection 10.3.5.
5.4.2.4 Tests and Inspections
5.4.2.4.1 Fabrication Tests and Inspections
The steam generator is tested in accordance with ASME Boiler and Pressure Vessel Code,Section III.
The nondestructive tests, some of which are not r equired by the code, performed during fabrication are given in Table 5.4-3.
During design and fabrication of the steam generator, additional operat ions beyond the requirements of the ASME Boiler and Pressure Vessel Code,Section III, were performed by the vendor. These included
ultrasonic testing for defects in tube sheet clad and ultrasonic testing of weld clad for bond integrity.
Initial hydrostatic tests of the primary and sec ondary sides of the steam generator are conducted in accordance with ASME Code,Section III. Leak test s are also performed. Following satisfactory performance of the hydrostatic tests, magnetic-parti cle inspections are made on all accessible welds.
Steam generator performance is further verified during the initial startup tests. Provisions for onsite
cleaning and cleanliness control are described in Subsection 5.2.3.
In-service Inspection will be performed during plant life in accordance with the ASME Code,Section XI as discussed in Subsection 5.2.4. (EC-8458, R307)
Inservice inspection of steam generator tubing w ill comply with the Technical Specifications.
5.
4.2 REFERENCES
- 1. WCAP-17263-P, Revision 0, November 2010, Regul atory Guide 1.121 Analysis and Structural Integrity Performance Criterion Applications for the Waterford Unit 3 Model Delta 110 Replacement Steam Generators for a NSSS Power of 1869.6 MWt/SG. (EC-8458, R307) 5.4.3 REACTOR COOLANT PIPING 5.4.3.1 Design Basis (DRN 06-546, R15)
The reactor coolant loop piping is designed and anal yzed for normal operation including all transients discussed in Subsection 3.9.1. Loading combinations and stress criteria associated with faulted conditions are presented in Subsection 3.9.3.1. In addition, certain nozzles are subjected to local transients that are included in the design analysis of the areas affect ed. Thermal sleeves are installed in the (DRN 06-546, R15)
WSES-FSAR-UNIT-3 5.4-16 Revision 305 (11/11) surge nozzle, safety injection nozzles, and chargi ng nozzle to accommodate these additional transients.
Principal parameters are listed in Table 5.4-
- 4. The ASME code and addenda to which the piping is designed is specified in Subsection 5.2.1.
In addition to being specified as seismic Category I, the following additional vibratory requirement is specified in the engineering specification. The va rious piping assemblies are designed so that no damage to the equipment is caused by the frequency ranges of 19 to 20 Hz and 95 to 100 Hz. The reasons for selecting these frequencies are the same as descr ibed in Subsection 5.4.
2.3.1.2 for the steam generators. Additional presentations relating to seis mic and dynamic analysis and criteria for the reactor coolant piping are contained in Subsec tions 3.7.2 and 3.9.2, respectively.
5.4.3.2 Description
Each of the two heat transfer loops contains five sections of pipe: one 42 in. ID pipe between the reactor vessel outlet nozzle and steam generator inlet nozzle, two 30 in. ID pipes from the steam generator's two
outlet nozzles to the reactor coolant pumps sucti on nozzle, and two 30 in. ID pipes from the pumps discharge nozzle to the reactor vessel inlet nozzles.
These pipes are referred to as leg hot leg, the suction legs, and the cold legs, respectively. The ot her major pieces of reactor coolant piping are the surge line, a 12 in. schedule 160 pipe between the pre ssurizer and the hot leg in Loop 1, and the spray line, a 4 in. schedule 120 pipe at the pressurize r end reduced to a 3 in. schedule 160 pipe between the pressurizer and the cold legs of Loops 1A and 1B.
To minimize the possiblity of stress corrosion cracki ng, the reactor coolant piping is fabricated from SA 516 GR 70 base material mill clad with type 304L stainle ss steel. Lines such as the surge lines, spray lines, and other small lines are totally made of stainl ess steel. Nozzles are shop fabricated with safe ends to preclude dissimilar-metal field welds. (EC-19087, R305)
Where stainless steel or Ni-Cr-Fe nozzle or safe end material is used, the safe ends are welded to the assembly after final stress relief to prevent furnace sensitization. Other prec autions used in the shop and during field assembly of the piping are described in Subsection 5.2.3. Pressurizer and hot leg nozzle welds that contain NI-CR-Fe materials susceptible to primary water stress corrosion cracking (PWSCC) have received structural weld overlays using Alloy 52M weld material (EC1830). (EC-19087, R305)
(DRN 03-1268, R13)
The 42 in. and 30 in. pipe diameters are selected to obt ain coolant velocities that provide a reasonable balance between erosion-corrosion, pressure drop, and sy stem volume. The surge line is sized to limit the frictional pressure loss through it during the maximum in-surge so that the maximum allowable pressure of the RCS is not exceeded. The spray li ne is sized to ensure a minimum spray flow of 375 gpm with both spray valves open. Pressuri zer parameters are listed in Table 5.4-6. (DRN 03-1268, R13)
To reduce the amount of field welding during plant f abrication, the 42 in. and 30 in. pipes are supplied in major pieces, complete with shop-installed inst rumentation nozzles and connecting nozzles to the auxiliary systems. Where required, the nozzles are supplied with safe ends to facilitate field welding of the connecting piping. To reduce thermal shock, thermal sleeves are provided for all nozzles two in. or
greater in diameter where fluid enters the main pi ping from an auxiliary system during normal operation.
WSES-FSAR-UNIT-3 5.4-17 Revision 307 (07/13)
Flow restricting orifices (7/32 in. diameter by one in. long) are provided in the nozzles for the RCS sampling lines, the reactor coolant pipe hot leg pre ssure measurement nozzles, the pressurizer level and pressure instrument lines, and the reactor coolant pump di fferential pressure instrument lines to limit flow in the event of a break downstream of a nozzle.
(DRN 04-292, R13;05-892, R14)
(DRN 04-292, R13;05-892, R14)
5.4.3.3 Evaluation
It is demonstrated by analysis that the reactor coolant piping is adequate for all normal operating and transient conditions of Subsection 3.9.1.1. In additi on, the fully assembled RCS is subjected to the required hydrostatic tests. Fracture toughness of t he reactor coolant piping is discussed in Subsection 5.2.3.
During the design phase, every effort is made to displace the frequency of the RCS piping from driving frequencies of concern by proper location of piping spring characteristics. The dy namic effects of system operation are also considered and piping restraints sized accordingly.
A discussion of the radiological considerations for the reactor coolant piping is provided in Section 12.3.
5.4.3.4 Tests and Inspections
Prior to and during fabrication of the reactor c oolant piping, nondestructive testing, based on the requirements of the ASME Code (see Table 5.2-1) is applied. Table 5.4-5 summarizes the component inspection program during fabricati on and construction. Tests for RCS integrity following normal opening, modification, or repair are specified in the Technical Specifications. To ascert ain the integrity of the piping during plant life, necessary in-service inspec tions required by Section XI of the ASME Boiler and Pressure Vessel are performed where required on the reactor coolant piping. To facilitate such inspections, longitudinal weld seams have been oriented at the 90 and 270 degree locations where feasible. Removable insulation is installed to assure access to the welds. Inservice inspection of the RCS piping is further discussed in Subsection 5.2.4.
5.4.4 MAIN STEAM LINE FLOW RESTRICTIONS (EC-8458, R307)
Main steam line flow restrictors are not required in the main steam line between the steam generator and the main steam isolation valves. A flow venturi (24 in.
throat diameter) acts as a flow restriction for main steam line breaks downstream of the venturi. Refer to Figure 10.2-4 for location of the flow venturi.
Note: See Section 5.4.2.2 for a descrip tion of integral flow restrictors. (EC-8458, R307) 5.4.5 MAIN STEAM LINE ISOLATION SYSTEM
The Main Steam Line Isolation System is discussed in Sections 7.3 and 10.3.
5.4.6 REACTOR CORE ISOLATION COOLING SYSTEM
This Subsection is not applicable to Waterford 3.
WSES-FSAR-UNIT-3 5.4-18 Revision 15 (03/07) 5.4.7 RESIDUAL HEAT REMOVAL SYSTEM
See Subsection 9.3.6.
5.4.8 REACTOR WATER CLEANUP SYSTEM
This Subsection is not applicable to Waterford 3.
5.4.9 MAINSTEAM LINE AND FEEDWATER PIPING
The Main Steam System piping is discussed in Subsection 10.3.6; Feedwater System piping is described
in Subsection 10.4.7.
5.4.10 PRESSURIZER
5.4.10.1 Design Bases (DRN 06-546, R15)
The pressurizer is designed and analyzed for the transients specified in Subsection 3.9.1.1 and the following additional requirements. During heatup and cooldown of the plant, the allowable rate of temperature change for the pressurizer is increased to 200F/hr. (DRN 06-546, R15)
The pressurizer is designed to:
a) Maintain RCS operating pressure
b) Compensate for changes in coolant volume during load changes
c) Contain sufficient volume to prevent draining the pressurizer as a result of a reactor trip
d) Limit the water volume to minimize the energy release during LOCA
e) Prevent uncovering of the heaters by the out-surge of water following load decreases; 10 percent step decrease and five percent per minute ramp decrease
f) Provide sufficient volume to accept the reactor coolant insurge resulting from a loss of load without the water level reaching the safety valve nozzles g) Provide sufficient volume to yield acceptable pressure response to normal system volume changes during load change transients (DRN 00-1059, R11-A) h) Achieve a total coolant volume change and associated charging and letdown flows which are as small as practical and are compatible with the capacities of the volume control tank, charging pumps and letdown control valves during load-following transients (DRN 00-1059, R11-A) i) Ensure that the minimum pressure observed during transients is above the setpoint of the safety injection actuation signal, and that the maximum pressure is below the high-pressure
trip.
WSES-FSAR-UNIT-35.4-19The heater capacity is selected to provide an adequate pressurizer heatup rate during plant startup.In addition to being specified as seismic Category I, the following additional vibratory requirement isspecified in the engineering specification. The pressurizer vessel, including the heaters, baffles, and supports shall be designed such that no damage to the equipment is caused by the frequency ranges of 19-20 Hz and 95-100 Hz. The lower frequency is defined as a mechanical vibration. The design basis for the higher frequency consists of a pressure pulse of five psi which diminishes internally within the vessel.5.4.10.2System DescriptionThe pressurizer is shown in Figure 5.4-6 and the design parameters are given in Table 5.4-6The pressurizer is a cylindrical carbon steel vessel with stainless steel clad internal surfaces. A spraynozzle on the top head is used in conjunction with heaters in the bottom head to provide level and pressure control. Overpressure protection is provided by two safety valves. The pressurizer is supported by a cylindrical skirt welded to the bottom head.The pressurizer is designed and fabricated in accordance with the ASME Code requirements listed inTable 5.2-1. The interior surface of the cylindrical shell and upper head is clad with weld deposited stainless steel. The lower head is clad with a Ni-Cr-Fe alloy to facilitate welding of the Ni-Cr-Fe alloy heater sleeves to the shell. A stainless steel safe end is provided on the pressurizer nozzles, after vessel final stress relief, to facilitate field welds to the stainless steel piping.The total volume of the pressurizer is established by consideration of the factors given in Subsection5.4.10.1. To account for these factors and to provide adequate margin at all power levels, the water volume in the pressurizer is programmed as a function of average coolant temperature as shown in Figure 5.4-7 in conjunction with Figure 5.4-8. High or low water level error signals result in the control actions shown in Figure 5.4-9.Pressure is maintained by controlling the temperature of the saturated liquid volume in the pressurizer. Atfull load conditions, slightly more than one-half of the pressurizer volume is occupied by saturated water, and the remainder by saturated steam. In order to maintain the programmed pressure, the corresponding saturation temperature must be maintained. To maintain this temperature, heaters are energized to compensate for heat losses through the vessel and to raise the continuous subcooled pressurizer spray flow to the saturation temperature.During load changes, the pressurizer limits pressure variations caused by expansion or contraction of thereactor coolant. The average reactor coolant temperature is programmed to vary as a function of load as shown in Figure 5.4-8. A reduction in load is followed by a decrease in the average reactor coolant temperature to the programmed value for the lower power level. The resulting contraction of the coolant lowers the pressurizer water level, causing the Reactor Coolant System pressure to decrease. This pressure reduction is partially compensated by flashing of pressurizer water into steam. All pressurizer heaters are automatically energized on low system pressure, generating steam and further limiting pressure decrease. Should the water level in the pressurizer drop sufficiently below its WSES-FSAR-UNIT-3 5.4-20 Revision 308A (10/15) setpoint, the letdown control valves close to a minimum value, and additional charging pumps in the Chemical and Volume Control System (CVCS) are autom atically started to add c oolant to the system and restore pressurizer level.
When load is increased, the average reactor coolant tem perature is raised in accordance with the coolant temperature program. The expanding coolant from the reactor coolant piping hot leg enters the bottom of the pressurizer through the surge line, compressi ng the steam and raising system pressure. The increase in pressure is moderated by the c ondensation of steam during compression and by the decrease in bulk temperature in the liquid phase.
Should the pressure increase be large enough, the pressurizer spray valves open, spraying coolant from the reactor coolant pump discharge (cold leg) into the pressurizer steam space. The relatively cold spray water condenses some of the steam in the steam space, limiting the system pressure increase.
The programmed pressurizer water level is a power dependent function. A high level error signal, produced by an insurge, causes the letdown control valves to open, releasing coolant to the CVCS and restoring the pressurizer to the programmed level. Small pressure and primary coolant volume variations ar e accommodated by the steam volume that absorbs flow into the pressurizer and by the water volume that allows flow out of the pressurizer.
The pressurizer heaters are single unit, direct imme rsion heaters that protrude vertically into the pressurizer through sleeves welded in the lower head.
Each heater is internally restrained from high amplitude vibrations and can be individually re moved for maintenance during plant shutdown.
Approximately one-fifth of the heaters are connected to proportional controllers that adjust the heat input as required to compensate for steady-state losses and to maintain the desired steam pressure in the pressurizer. The remaining backup heaters are connect ed to on-off controllers. These heaters, normally deenergized, are turned on by either a low-pressurizer pressure signal or high-level error signal. This latter feature is provided since load increases result in an in-surge of relatively cold coolant into the pressurizer, thereby decreasing the bul k water temperature. The CVCS acts to restore level, resulting in a transient pressure below normal operating pressure.
To minimize the extent of this transient, the backup heaters are energized, contributing more heat to the water. An interlock prevents operation of the backup heaters in the event of concurrent high level error and high-pressurizer pressure signals. A low-low pressurizer level signal deenergizes all heater s to protect the heaters should they become uncovered. (LBDCR 15-028, R308A)
In order to determine the pressurizer heater capacity r equired to maintain natural circulation in the hot standby condition after a loss of offsite power, it was conservatively assumed that the ambient heat loss rate through the pressurizer was 400,000 BTU/hr. T he measured heat loss from startup testing was only 356,000 BTU/hr. With an assumed 400,000 BTU/hr heat loss and a safety valve leakage of up to 0.5 gpm, single phase natural circulation can be ma intained at hot standby conditions with a 50 F subcooled margin indefinitely by energizing 150kW of heater capacit y thirty minutes after the loss of offsite power.
Loss of subcooling, however, does not imply loss of nat ural circulation. The nat ural circulation cooldown analysis (refer to FSAR Section 9.3.6.3.3.1), perform ed to comply with Branch Technical Position 5-4, Design Requirements of the Residual Heat Remova l System, does not credit the operation of any pressurizer heaters. Therefore, t he operator action to energize the Pressurizer Heaters is not a time critical operator action
) (LBDCR 15-028, R308A)
A redundant group of pressurizer proportional heat ers and three redundant groups of backup heaters have been made available to be placed manually on the em ergency diesel generator after a loss of offsite power. Each bank of heaters has access to only one Class 1E division power supply. (DRN 00-1673)
WSES-FSAR-UNIT-3 5.4-21 Revision 308A (10/15)
(LBDCR 15-028, R308A)
Part of the closing circuitry to the breakers that provide power to the 480V non-safety switchgear buses 3A32 and 3B32 (that power the Pressurizer Heater s) share a specific common circuit breaker, CVCEBKR014AB-13. CVCEBKR014AB-13 powers t he interlock 52z relay, SSDEREL2348-D (SSDEREL2398-D). The interlock 52z relay checks fo r completion of load stripping on the respective 480V non-safety switchgear buses 3A32 (3B32) at the onset of a Loss of Offsite Power. If the load stripping is complete, the interlock 52z relay closes a contact in the closing circ uitry to the breakers that provide power to the 480V non-safety switchgear bus es 3A32 and 3B32 to allow the breakers to close automatically when the sequencer load block cont act in the closing circuitry is closed.
Alternatively, if the specific common circuit break er, CVCEBKR014AB-13, is Open, then the breakers that provide that power to the 480V non-safety switc hgear buses 3A32 and 3B32 will not close automatically at the onset of a Loss of Offsite Power. To cl ose the breakers that power each Pressurizer Heater electrical switchgear 3A32 (3B32), local manual action in the respec tive train Switchgear room is necessary.
Reenergization of the necessary heaters from the emergency onsite power can be accomplished manually from the control room. At the onset of a Lo ss of Offsite Power concurrent with the specific common circuit breaker, CVCEBKR014AB-13, being O pen, the reenergization of the 480V non-safety switchgear buses 3A32 and 3B32 (that power the Pressu rizer Heaters) will require action to be performed outside of the Control Room. To close the break ers that power the 480V nonsafety switchgear buses 3A32 and 3B32, local manual operator action in the respec tive train Switchgear room is necessary. Once each 32 switchgear bus is reenergized, the necessary Pressurizer Heaters powered from that bus can be reenergized from the Control Room.
The natural circulation cooldown analysis (refer to F SAR Section 9.3.6.3.3.1), performed to comply with Branch Technical Position 5-4, Design Requirements of the Residual Heat Removal System, does not credit the operation of any pressurize r heaters. Therefore, the operator action to close the breakers that power each Pressurizer Heater electr ical switchgear 3A32 (3B32), locat ed outside of the control room, is not a time critical operator action.
Procedures ensure that the addition of these loads after a loss of offsite power will not exceed the rating of the emergency di esel generator. The heaters are powered from the 480V non-safety switchgear buses 3A32 and 3B32. T he safety-related Class 1E breakers providing power to these buses from the 4.
16kV ESF buses (3A3-S and 3B3-S) w ill trip upon LOOP or SIAS. In this manner, the Class 1E interfaces for main pow er and control power to the pressurizer heaters are protected by safety-grade circuit br eakers. This scheme also ensures that in case of an SIAS the non-Class 1E pressurizer heaters are shed from thei r emergency power sources via Class 1E circuit breakers. (LBDCR 15-028, R308A)
Pressurizer spray is supplied from each of the r eactor coolant pump cold legs in loop one to the pressurizer spray nozzle. Automatic spray control va lves control the amount of spray as a function of pressurizer pressure; both of the spray control va lves function in response to the signal from the controller. These components are sized to use t he differential pressure between the pump discharge and the pressurizer to pass the amount of spray required to maintain the pressurizer steam pressure during normal load following transients. A small continuous flow is maintained through the spray lines at all times to keep the spray lines and the surge line warm to reduce thermal shock during plant transients.
This continuous flow also serves to keep the chem istry and boric acid concentration of the pressurizer water the same as that of the coolant in the heat transfer loops.
An auxiliary spray line is provided from the chargi ng pumps to permit pressurizer spray during plant heatup, or to allow depressurization and cooling if the reactor coolant pumps are shut down. The capability to depressurize using auxiliary spray with one charging isolation valve failed open was demonstrated by testing. With a charging to cold leg isolation valve open, some of the charging pump flow is diverted from the auxiliary sprayflow path to the cold legs. With two charging pumps running (88 gpm) an auxiliary spray flow rate of 37 gpm was achiev ed. This resulted in a depressurization rate of 24 psi/min. This is sufficient to depressurize t he RCS during any design basis accident where auxiliary spray is required WSES-FSAR-UNIT-3 5.4-22 Revision 307 (07/13)
In the event of an abnormal transient that causes a su stained increase in pressurizer pressure at a rate exceeding the control capacity of the spray, a high-pr essurizer pressure reactor trip will be initiated.
5.4.10.3 Safety Evaluation
It is shown by analysis made in a ccordance with the requirements of the ASME Code,Section III that the pressurizer is adequate for all normal operating and trans ient conditions expected during the life of the plant. Following fabrication, the pressurizer was hydrostatically and nondestructively tested in accordance with the ASME Code,Section III.
During hot functional testing, the transient performanc e of the pressurizer is checked by determining its normal heat losses and maximum pressurization and depressurization rates. This information is used in
setting the pressure controllers.
Overpressure protection of the RCS is provi ded by two ASME Code spring loaded safety valves.
A discussion of the radiological considerations fo r the pressurizer is provided in Section 12.3.
5.4.10.4 Inspection Testing and Requirements
Prior to and during fabrication of the pressurizer, nondestructive testing is performed in accordance with the requirements of the ASME Code Section III. The pr essurizer inspection program is summarized in Table 5.4-7.
Further assurance of the structural integrity of the pressurizer duri ng plant life will, be obtained from the inservice inspection performed in accordance with the ASME Code,Section XI and described in Section 5.2.
5.4.11 QUENCH TANK (PR ESSURIZER RELIEF TANK)
5.4.11.1 Design Basis
The quench tank is designed to receive and condens e the normal discharges from the primary (pressurizer) safety valves and to prevent t he discharge from being released to containment.
(DRN 03-2059;05-316, R 14; EC-8458, R307)
The tank is sized to receive and condense a total steam release of 1232 lbm. The maximum normal discharge that the quench tank must withstand occu rs during the loss of condenser vacuum event (described in Section 15.2.1), which is approximately 1200 lbm. (DRN 03-2059;05-316, R 14; EC-8458, R307)
The quench tank is mounted on structural steel fr aming supported from the secondary shield wall.
5.4.11.2 System Description
The quench tank, shown in Figure 5.4-10, is an aust enitic stainless steel vessel suitable for prolonged contact with borated, demineralized water. Nozzles are provided for the safety valve discharge line, vents, drains, instrumentation, makeup water, nitr ogen addition, and the rupture di sc. The tank prevents the steam released from the primary (pressurizer) safe ty valves from being released to the containment atmosphere. The steam is discharged under water by the sparger and condensed. Demineralized water is manually added from the Reactor Auxiliary Building to cool the tank water after a steam discharge. A rupture disc venting to the cont ainment atmosphere is provided fo r overpressure protection.
Noncondensible gases within the quench tank are v ented to the containment vent header through an air-operated valve.
WSES-FSAR-UNIT-3 5.4-23 The tank is designed and fabricated in accordance with the ASME Code,Section VIII. The design parameters am given in Table 5.4-8.
The sparger, spray header, nozzles and rupture disc fittings are stainless steel.
5.4.11.3 Safety Evaluation The quench tank and associated blowdown system are sized such that the maximum safety valve backpressure, 500 psig, is not reached during any antic ipated transient steam release. The quench tank rupture disc has a relief capacity greater than the co mbined relief capacities of the primary (pressurizer) safety valves.
The quench tank is classified as non-nuclear safety cla ss as defined in Section 3.2. The failure of any of these components in no way compromises the integrit y of the RCS pressure boundary or safe shutdown of the plant, nor does it in any way jeopardize the safety of the public.
5.4.11.4 Instrument ation Requirements The quench tank is equipped with level, pre ssure, and temperature instrumentation.
A brief explanation of each instrument is made below.
The location of each of the sensors and functional requirements of the inst ruments are indicated on Figure 5.1-3.
5.4.11.4.1 Temperature The temperature measurement channel consists of a precision resistanc e temperature detector (RTD), a temperature transmitter and associated indicators and al arms. The resistance of the RTD varies as a function of the temperature of the RTD environment.
This change in RTD resistance is sensed by the temperature transmitter and converted to a change in a low dc current, which is used as a signal to the remote temperature indicators and alarms.
Quench tank temperature indication and high temperatur e alarm are provided in the main control room.
A high temperature alarm may be indicative of primary safety valve leakage to the tank.
5.4.11.4.2 Level
Level instrumentation consists of a level trans mitter and associated indicating and alarming equipment.
The level transmitter measures t he pressure difference between a reference column of water and the vessel water level. This pressure difference is conver ted to a small dc current which is proportional to the level of water in the vessel. This dc current output is used as a signal to the remote level indicators and alarms. Quench tank level indication and high and low level alarms are provided in the main control room. A low level alarm is indicative of the sparger being uncover ed or of insufficient tank fluid volume to quench the design basis accident steam release. A high level al arm alerts the operator of insufficient volume in the quench tank to accept the pressurizer steam releas e without becoming overpressurized and causing the rupture disc to burst.
5.4.11.4.3 Pressure
A pressure measurement instrument consis ts of an electric force balance pressure transmitter and associated indicators and alarms. The transmitter produces dc current output that is proportional to the pressure sensed by the instrument. This dc current output is used as a signal to the remote pressure indicators and alarms.
Quench tank pressure indication and high pressure alarm are provided in the main control room. The high pressure alarm alerts the operator to the sit uation prior to the bursti ng of the rupture disc.
WSES-FSAR-UNIT-35.4-24Revision 11 (05/01)5.4.11.5Inspection and Testing RequirementsThe quench, tank is designed to handle the design basis steam releases as described in Subsection5.4.11.2. However, should the rupture disc burst, the primary coolant released to the containment would be minimal when compared to the LOCA for whichthe engineered safety features are designed to accommodate. Therefore, no inspection or testing requirements are imposed on the quench tank.Since, the entire system is located within the containment structure, the quench tank must be vented tocontainment atmosphere, during the containment leakage test, to avoid collapse of the tank.5.4.12VALVES5.4.12.1Design BasisThe safety-related functions of valves within the reactor coolant pressure boundary are to act as pressureretaining vessels and leaktight barriers during normal plant operation, accidents and seismic disturbances.These valves are designed in accordance with the applicable ASME Code,Section III, or the Draft ASMECode for Pumps and Valves, Class I requirements and must withstand the effects of the system design transients (see Subsection 3.9.1.1) plus other transients associated with the valves location or service requirements. The valves meet seismic Category I requirements. Backseats are specified on manualand motor-operated gate and globe valves to further minimize potential leakage. Functional requirements foreach valve are detailed in the individual valve specifications.Materials of construction are specified to assure compatibility with the environment and contained fluids.
5.4.12.2Design Description(DRN 99-811)All valves in the Reactor Coolant System are constructed primarily of stainless steel. Other materials incontact with the coolant, such as hard facing and packing, are compatible materials. Fasteners, packing gland assemblies, and yoke fasteners are also constructed of stainless steel to eliminate corrosion problems. Valve packing glands have provisions to adjust packing compression to reduce or eliminate leakage. These features keep uncontrolled leakage from these valves at essentially zero.(DRN 99-811)5.4.12.3Design EvaluationAll valves within the reactor coolant pressure boundary are stress analyzed in accordance with theapplicable ASME Code, taking into consideration cyclic loadings. Refer to Section 3.9.
WSES-FSAR-UNIT-35.4-25 Revision 11-A (02/02) 5.4.12.4 Tests and Inspections The valves are hydrostatically tested and leak tested across the seats and across the packing in-accordance with individual valve specifications and the applicable ASME Code. Valves greater than 1 inch
inner diameter are dimensionally checked, including measurements to determine wall thickness.
In-service inspection will be performed during the plant life in accordance with ASME Code Section XI, as discussed in Subsection 5.2.4.5.4.13SAFETY AND RELIEF VALVES5.4.13.1Design BasisThere are no power-operated relief valves in the RCS. The primary safety valves on the pressurizer are designed to protect the system, as required by the ASME Code,Section III.
The design basis for establishing the relieving capacity of the primary safety valves is presented in Appendix5.2A. For the postulated transients presented in Chapter 15, the results demonstrate that the pressurizer
will not go "solid" and that the relieving capacity of the safety valves is sufficient to provide overpressure
protection in accordance with Section III of the ASME Code.
Safety valves on the steam side of each steam generator are designed to protect the steam system, asrequired by the ASME Code,Section III. They are conservatively sized to pass a steady flow equivalent to
the maximum expected power level at the design pressure of the Main Steam System.
5.4.13.2 DescriptionThe RCS has two safety valves to provide overpressure protection. The safety valveis illustrated in Figure5.4-11. Thedesign parameters are given in Table 5.4-9. They are direct acting, spring-loaded safety valvesmeeting ASME Code requirements. They have an enclosed bonnet and have a balanced bellows for superimposedbackpressure. The safety valves pass sufficient pressurizer steam to limit the RCS pressureto 2750 psia (110 percent of design) following a complete loss of turbine generator load without simultaneous reactor trip. A delayed reactor trip is assumed on a high-pressurizer pressure signal. To determine maximum steam flow through the primary safety valves, the main steam safety valves are assumed to be operational. Values for the system parameters, delay times, and core moderator coefficient
are given in Chapter 15.Overpressure protection for the shell side of the steam generators and the main steam tine up to theinlet ofthe turbine stop valve isprovided by the secondarysafety valves.5.4.13.3E valuation Overpressure protection is discussed in Subsection 5.2.2. The ASME Code Report on Overpressure Protection, is included as Appendix 5.2A.(DRN 00-1059; 02-88)(DRN 00-1059; 02-88)
WSES-FSAR-UNIT-3 5.4-26 Revision 14 (12/05
)(DRN 00-1059, R11-A; 02-88, R11-A)5.4.13.4 Tests and Inspections(DRN 00-1059, R11-A; 02-88, R11-A)The valves are inspected during fabrication in accordance with ASME Code,Section III requirements. The inlet is hydrostatically tested and the outlet is pneumatically tested. Seat leakage is checked by a steam
test at 93 percent of set pressure. Set pressures are adjusted using steam. In-service inspection will be performed during plant life in accordance with the ASME Code Section XI as discussed in Subsection 5.2.4. 5.4.14 COMPONENTS SUPPORT (DRN 03-2059, R14) 5.4.14.1 Design BasesThe criteria applied in the design of the RCS supports are that the specific function of the supported equipment be achieved during all normal earthquake, and pipe break conditions. Specifically, the supports are designed in accordance with the design loading combinations which are applied in the design of ASME Code Class I supports. These design loading combinations are categorized with respect to their plant operating conditions, which are identified as normal, upset, emergency and faulted (as defined in the
ASME Code,Section III, for Class I components). The following design loading combinations are applied: (DRN 03-2059, R14) a) Loading Combination 1 (Upset) The concurrent loadings associated with either the normal plant condition or the upset plant condition and the vibratory motion of the operational basis earthquake (OBE). b) Loading Combination 2 (Emergency) The concurrent loadings associated with the plant emergency condition. c) Loading Combination 3 (Faulted) (DRN 03-2059, R14) The combined loadings associated with the normal plant condition, the vibratory motion of the SSE, and the dynamic system loadings associated with postulated pipe ruptures as discussed in
Subsections 3.6.2, 3.6.3, and 6.2.1.2. This loading combination is assumed to occur at 100 percent
power steady state operation. (DRN 03-2059, R14) For supports that are an integral part of RCS components, the design stress limits of the parent component are used as follows: a) The design stress limits applied in conjunction with loading combinations 1 and 2 arein accordance with the rules of ASME Code Section III, Paragraph NB-3223 for upset conditions and Paragraph NB-3224 for emergency conditions, respectively.
WSES-FSAR-UNIT-3 5.4-27 Revision 307 (07/13)
Nonintegral supports are designed and constructed to the stress limits of ASME Section III, Subsection NF. The design stress limits used are as follows:
a) The design stress limits applied in conjunc tion with loading combinations 1 and 2 for plate type supports and linear type supports are in accordance with the rules of Subsubarticles NF-3220 and NF-3230, respectively.
For these supports the design stress limits applied in conjunction with loading combination 3 are
dependent upon the type of system or subsystem analyses used to establish design loadings. Elastic system or subsystem analysis us ed to determine the design loadings produced by loading combination 3 and the associated design stress limits applied in the design of supports are in accordance with the rules
of the ASME Code,Section III, Appendix F, Subsubarticl e F-1323, or plastic instability load criteria are applied in accordance with paragraphs F-1370 (d) and F-1370 (e) of Appendix F, including the effects of the resulting plastic deformation.
5.4.14.2 Description
Figure 5.4-12 illustrates the RCS support points.
A description of each supported component follows.
5.4.14.2.1 Reactor Vessel Supports
The reactor vessel integral supports consist of four pads welded to the underside of the vessel inlet nozzles. Vertically, the pads rest on lubricated bearing plates, and contain studs that act as holddown
devices for the vessel. Horizontal keyways inte rface with the pads. The arrangement of the vessel supports allows radial growth of the reactor vessel due to thermal expansion while maintaining it centered. The supports are designed to accept normal loads and seismic and pipe rupture accident loads.
Reactor vessel supports are shown in Figure 5.4-13 and Figure 3.8-34.
5.4.14.2.2 Steam Generator Supports (DRN 00-1059, R11-A; 03-2059, R14) The steam generator weight is supported at the bottom by a sliding base bolted to an integrally attached conical skirt. The sliding base rests on low friction bearings which allow unrestrained thermal expansion
of the RCS. Two embedded keys mate with keyways wi thin the sliding base to guide the movement of the steam generator during expansion and contraction of the RCS and, together with anchor bolts, prevent excessive movement of the bottom of t he steam generator during seismic events and following a pipe break. (DRN 00-1059, R11-A; 03-2059, R14)
A system of keys and snubbers located on the steam drum guide the top of the steam generator during expansion and contraction of the RCS and provide support during seismic events and following a primary side or secondary side pipe break.
(EC-8438, R307)
The mechanical/structural loads associated with the dynamic effects of a LOCA in the RCS hot leg and cold leg piping have been eliminated with the applicat ion of Leak-Before-Break (LBB) (Reference Section 3.6.3). Based on removal of the RCS dynamic pipe br eak loads, the shim plate pack (stop) on the reactor side of the keyway (underneath the RCS hot leg) was permanently removed from the SG-1 and SG-2 sliding base. Additionally, the SG-1 and SG-2 slidi ng base supports were modified to remove the shim plate from the perimeter of the SG support skirt fl ange. These modifications were performed as part of the changeout of the steam generators. (EC-8438, R307)
Steam generator supports are shown in Figure 5.4-14.
WSES-FSAR-UNIT-3 5.4-28 Revision 14 (12/05
)5.4.14.2.3 Reactor Coolant Pump Supports (DRN 03-2059, R14) Each reactor coolant pump is provided with four vertical support columns, four horizontal support columns, and one horizontal snubber. The rigid structural columns provide support for the pumps during normal operation, earthquake conditions, and any postulated pipe break in either the pump suction or discharge
line 2. An illustration of the pump supports is shown in Figure 5.4-15. (DRN 03-2059, R14) For the case of a pipe break in the pump discharge line, three structural stops are, provided to limit the pump motion. Pipe stop structures that limit pipe motion are also provided to prevent overloading of the pump support columns due to a pipe rupture at either the steam generator or reactor vessel nozzles. 5.4.14.2.4 Pressurizer Supports (DRN 03-2059, R14) The pressurizer is supported by a cylindrical skirt welded to the bottom of the pressurizer and bolted to the support structure. The skirt is designed to withstand dead weight and normal operating loads as well as the loads due to earthquakes and postulated pipe break. An illustration of the pressurizer supports is
shown in Figure 5.4-16.
5.4.14.3 EvaluationThe structural integrity of the reactor coolant component supports is confirmed by analyses, using the design basis presented in Subsection 5.4.14.1. The analyses employ dead weight, thermal, seismic and pipe break loadings, combined in the various load combinations. Dead weight and thermal loads are determined by static analysis of the RCS. The method of determining seismic loadings is described in Subsection 3.7.2. RCS response to postulated branch line pipe breaks (BLPBs) for extended power uprate to 3716 MWt is determined using non-linear time history analysis of a full three-dimensional ANSYS model of the RCS. The effects of BLPBs on the RCS component supports include the effects of pipe tension release, jet impingement, RV blowdown, SG subcompartment and component-to-support gaps.
Subsequent evaluations demonstrate the adequacy of the RCS supports under extended power uprate conditions to 3716 MWt, per the acceptance criteria provided in Section 3.9.3. Structures are provided to mate with the component supports to restrain and support RCS components.The loads at the support/structure interface locations are examined under normal, OBE, SSE, and pipe break conditions (see Section 3.8). Seismic and accident loads are determined by the same methods referred to in the first paragraph of this section, taking into account the structural characteristics at the support/structure interfaces. The design of each support is compatible with the design radiation levels
given in Section 3.11. (DRN 03-2059, R14) 5.4.14.4 Testing and Inspection Tests were conducted on materials similar to that being used for the reactor vessel and steam generator sliding supports to demonstrate that the maximum static coefficient of friction does not exceed 0.15 at a design loading of 5000 psi. Tests on sliding supports and the steam generator basis are in accordance with ASME Section III. In addition, all sliding supports are 100 percent liquid penetrant inspected at the vendor's shop. The steam generator base is ultrasonically and magnetic particle inspected. The specifications for the steam generator snubbers require that they be tested in the vendor's shop at the rated load capacity in both tension and compression. The piston creep during these tests must comply with specification limits. Tests are also specified for initiation of snubber action in both tension and
compression. (DRN 03-2059, R14) 2These pipe breaks, which have been eliminated via LBB (Section 3.6.3) and which have been replaced by postulated BLPBs for extended power uprate to 3716 MWt, continue to provide the enveloping design
loads on the RCP supports. (DRN 03-2059, R14)
WSES-FSAR-UNIT-3 5.4-29 Revision 14 (12/05
)Supports integral to the RCS components receive quality assurance inspections in accordance with the ASME Code,Section III, during fabrication. During preoperational testing of the RCS, the support displacements will be monitored for agreement with calculated displacements and/or clearances. Subsequent examinations of supports of RCS components
will be in acccordance with the ASME Code,Section XI. 5.4.15 REACTOR COOLANT GAS VENTING SYSTEM 5.4.15.1 Design Bases The Reactor Coolant Gas Venting System (RCGVS) is designed to allow for remote venting of non-condensible gases, which may collect in the RCS, via the reactor vessel head vent or pressurizer steam space vent during post-accident situations. The system may be used for normal RCS venting when required during plant outages. This system has been designed to meet the requirements of NUREG-0737.
The design criteria for the RCGVS are as follows: a) The system permits remote (control room) venting of the reactor vessel head or the pressurizer. b) The system is designed for a single active failure with active components poweredfrom their respective redundant emergency power sources. The system has parallel vent paths with valves powered from alternate power sources. A single active failure in the power and control portion of the vent system will not prevent isolation of the entire vent system when required. c) The system is designed primarily to vent noncondensible gas. d) The vent flow capability is based upon the following considerations: 1) The vent rate is sufficient to vent 4800 standard cubic feet of hydrogen per minute.2) Coolant liquid loss through the vent will not exceed the makeup capacity ofone charging pump in the event of a Safety Class 2 pipe break or inadvertent valve
operation, thus limiting leakage to less than the LOCA definition of 10CFR50. 3) The vent mass rate will not result in heat loss from the RCS in excess of the normal pressurizer heater capacity. e) Vent paths are provided to the quench tank which allows for cooling of gases andcondensing water vapor by releasing the vented gases-below the water level in the tank. The vent paths are safety grade and, being a part of the RCS, meet the same qualifications as the existing
RCS.f) A method of leakage detection is provided to identify and ensure that leakage in the WSES-FSAR-UNIT-3 5.4-30 Revision 301 (09/07) system is identifiable. This allows continued power operations at leak rates greater than 1 gpm but less than 10 gpm (refer to plant Technical Specifications).
g) The solenoid operated valves are powered fr om safety grade 120V ac power supplies.
Power is removed from the fail closed valves by utilizing key-locked control switches to minimize the possibility of inadvertent operation during normal operation.
h) Valve position indication (open-closed) is provided in the control room for all remotely operated valves.
i) The system is designed so that each v ent valve may be tested for operability during plant operation. Testing can be performed in acco rdance with Subsection IWV of Section XI of the ASME code for Category B valves.
See Figure 5.4-17 for flow diagram arrangement.
5.4.15.2 Description/Principal Modes of Operation
5.4.15.2.1 Startup
Venting of the RCS prior to plant startup can be accomplished using the RCGVS. The reactor vessel can be vented by opening either 2RC-E2559A or 2RC-E 2560B and the pressurizer can be vented by opening either 2RC-E2557A or 2RC-E2558B. Venting is routed to the quench tank. Air flow can then be routed to
the Waste Management System vi a the containment vent header.
5.4.15.2.2 Normal Operation
This system is not intended for use during normal power operation and administrative controls are
provided to minimize the possibility of inadvertent operation. In addition, power is removed from all valves during normal plant operation.
During normal operation, leakage detection is maintained by use of pressure instrumentation. A rise in pressure will indicate leakage past valves 2 RC-E2557A, 2RC-E2558B, 2RC-E2559A or 2RC-E2560B.
The ability to identify leakage above 1 gpm from the RCS may be accomplished in using either of two methods:
a) If the pressure increase is slow enough, the leakage rate can be determined by observing the rate of pressure increase per unit time.
b) The leakage would be diverted to the quenc h tank and the increase in quench tank level can be translated into a leakage rate.
5.4.15.2.3 Refueling Shutdown (EC-5000080002, R301) a) During refueling shutdown, valves 2RC-E2557A and/or 2RC-E2558B,2RC-E2559A and/or 2RC-E2560B are opened to align the reactor ve ssel to the pressurizer which is vented to atmosphere. Opening 1RC-V2590 A/B (RC-10111) and 1RC-V2506 (RC-101) provides an alternate vessel vent path. Either path prevent s vacuum formation during reduction of reactor vessel level. (EC-5000080002, R301) b) During shutdown for refueling, valves 2RC-E2557A, 2RC-E2558B,2RC-E2559A, 2RC-E2560B, 2RC-2561A and 2RC-E2562B are to be tested per In-service Testing requirements.
WSES-FSAR-UNIT-35.4-315.4.15.2.4Accident ConditionsPrior to system operation, the quantity of non-condensible gases in the RCS can be estimated. This canbe accomplished through evaluation of the inadequate core cooling instrumentation that will be provided in accordance with the requirements of NUREG-0737 II.F.2 (Section 1.9.30) for the case of gas in the reactor vessel and by the response of system pressure control methods or departure from saturation conditions in the pressurizer for the case of gas in the pressurizer. Additional indirect methods involving observing pressurizer level response to a given pressure change can be used as a backup method of gas volume determination.The vent system is aligned to vent from the reactor vessel or pressurizer to the quench tank. Smallquantities of gas may be vented to the quench tank without rupturing the quench tank rupture disc. This permits gas removal from the RCS without contaminating the containment atmosphere. Gas may be discharged from the quench tank to the Waste Management System.5.4.15.3EvaluationThe RCGVS may be required to operate during post-accident situations to remove noncondensible gasesfrom the RCS. To assure operability under those conditions, the components of the system required to perform venting operations have been environmentally qualified to operate under post-accident containment conditions. They are provided with emergency power and parallel valves powered from redundant power sources. Parallel valves assure a vent flow path to containment in the event of single active failure.The RCGVS is supplied with flow limiting orifices which limit mass loss from the RCS to an amount lessthan the makeup capacity of a charging pump.The components, piping and supports in the RCGVS are specified as seismic Category I. All valves havebeen qualified for operability during and following a seismic event.5.4.15.4Required Componentsa)Piping and ValvesAll piping and valves are constructed of austenitic stainless steel and are safety grade asrequired. The piping size is 1 inch Schedule 160. The six Safety Class 2 solenoid operatedvalves are designed to fail closed to minimize inadvertent operation. The Safety Class I orifices, one of which is included in the present RCS design, are sized to meet the specified flow requirements.b)Instrumentation and ControlsThe system is designed to be controlled remotely from the main control room. All safety-relatedinstrumentation is powered from emergency power sources. Position indication (open/closed) is provided for all remotely operated valves and displayed in the control room. Pressure instrumentation is alsoprovided to monitor system performance and any valve leakage. Pressure indication is located in the control room. Quench tank level is also located in the control room.
WSES-FSAR-UNIT-35.4-32 Revision 11-A (02/02) 5.4.16 REFUELING WATER LEVEL INDICATING SYSTEM (RWLIS) 5.4.16.1 Design Bases(DRN 00-1059)
The REFUELING WATER LEVEL INDICATING SYSTEM (RWLIS) is designed to perform the following functions.(DRN 00-1059)a)Monitor the water level in the Reactor Coolant System and the refueling pool during refueling operations.b)Monitor the water level in the RCS hot leg during maintenance evolutions requiring the RCS water levels at elevations within the range of the hot leg.c)Provide indication of the level locally in the containment building and remotely in the control room.d)Provide an alarm in the control room when the water level in the refueling pool exceeds a predetermined level or when the water level drops below a predetermined level in the
range of the hot leg.
Refer to Figure 5.4-18 for a diagram of the system showing its connection to the RCS and the major components.The design pressure and temperature of the RWLIS are 100 psig and 250
°F. With the design conditions, the system's normal operational conditions are modes 5 and 6 with the RCS vented and pressurizer level greaterthan 5% and less than 75% as indicated by the pressurizer cold calibrated level instrumentation. Thesystem is isolated from the RCS during plant operations. The RWLIS can be operated with the vessel head
in place.5.4.16.2System Description The RWLIS is comprised of a narrow and a wide range differential pressure transmitter attached through stainless steel piping to a primary system high point near the top of the pressurizer and to the hot leg drain on the RCS (see Figure 5.4-18). The signals from these transmitters are provided to a local indicator inside
containment, and to a remote indicator in the control room.The RWLIS is designed to operate with the reactor vessel head installed or removed. With the exception ofthe local RWLIS indicator in containment, the system is designed for permanent installation. The RWLIS
will compensate for slight positive and negative pressure variations in the RCS by utilizing a differential level
instrument arrangement to compare actual RCS water level to a reference leg vented to an optimum system
high point.
The low level sensing line is permanently attached to the drain line just below hot leg #1 in the vicinity of the shutdown cooling suction line.
The reference leg is permanently attached to the existing upper level tap of the pressurizer wide range level instrument.
WSES-FSAR-UNIT-35.4-33The transmitters and the local indicator are mounted on an instrument stand outside the biological shield.The location was chosen to minimize tubing runs and to allow easy draining of the reference leg to existing RCS level drain pipe to maintain the reference leg dry during normal RWLIS operation.The level signals are conducted to the control room panel and annunciator window. The RWLIS signalsare also provided to the Plant Monitoring Computer.5.4.16.3EvaluationThe valves which maintain the systems' isolation during plant operation are designed Class I and Class IIfor the isolation at the hot leg and the pressurizer, respectively, and are part of the Reactor Coolant Pressure Boundary. These valves are discussed in Subsection 5.4.12. The remainder of the RWLIS is installed as Class III. The tube routing and instrument stands are installed as Seismic Class 1. The RWLIS is not installed as safety-related instrumentation.5.4.16.4Tests and InspectionsInservice inspection of the boundary valves is performed during the plant life in accordance with ASMESection XI as discussed in Subsection 5.2.4.Operating procedures for the RWLIS require that the system be calibrated prior to draindown to ensureaccurate level indication.5.4.17REACTOR COOLANT SHUTDOWN LEVEL MEASUREMENT SYSTEM (RCSLMS) 5.4.17.1Design BasesThe Reactor Coolant Shutdown Level Measurement System (RCSLMS) is designed to perform thefollowing functions:a)Monitor the water level in the Reactor Coolant System (RCS) during non-poweroperation.b)Monitor the water level in the RCS hot leg during maintenance evolutions requiringthe RCS water levels at elevations within the range of the hot leg.c)Provide indication of the level locally in the containment building and remotely inthe control room.d)Provide an alarm in the control room when water level drops below a predeterminedlevel in the range of the hot leg.Refer to Figure 5.4-19 for a diagram of the system showing its connection to the RCS and the majorcomponents.The design pressure and temperature of the RCSLMS are 100 psig and 300
°F. With the designconditions, the system's normal operational conditions are modes 5 and 6 with the RCS vented and pressurizer level greater than 5% and less than 75% as indicated by the pressurizer cold WSES-FSAR-UNIT-35.4-34 Revision 12 (10/02)calibrated level instrumentation. The system is isolated from the RCS during plant operations by removing the temporary connections.5.4.17.2System Description(DRN 01-3975)The RCSLMS is comprised of the Mansell Level Monitoring System (level instrument) and a skid mountedsight glass. This RCSLMS is connected by stainless steel tubing to a primary point at the top of the pressurizer and to the hot legs on the RCS (see Figure 5.4-19). The sight glass provides local indication.
The Level instrument is installed on a temporary basis during refueling operations. The level instrument
provides indication to the existing control room annunciators and indicators. Indication also exists on the level instrument computer system that is temporarily mounted in the Control Room during refueling
operations to provide for system status of the instrument and level of the RCS.(DRN 00-1059)
The level instrument is comprised of two channels, each containing a reference (low-pressure) transducer assembly and a high-pressure transducer assembly that provides absolute pressure measurements to the computer system. The computer system computes the level of the RCS and displays the level in the control room. The computer system has a channel selectable output of 0-10VDC to the RCSLMS process
analog control cards for conditioning and transmission of the signal to the control room indicators.(DRN 00-1059; 01-3975)
The RCSLMS is designed to function during non-power operations, providing a redundant measurement ofRCS level. The system is designed to have temporary connections which are made only during non-power operation. The system is isolated and disconnected during normal operation.5.4.17.3E valuation During power operation, the RCSLMS is disconnected, providing isolation from the RCS pressure boundaryat the hot leg and pressurizer. The temporary connections are made after the reactor pressure has dropped
below 300 psig in a section of piping which is Cat. 7 NNS and has valving to provide isolation for making theconnections. These valves are part of Reactor Coolant Pressure Boundary and are discussed in Subsection
5.4.12. The tubing, tube supports and instrument skid are installed as Seismic Class 1. The RCSLMS is installed as non-safety-related instrumentation.
5.4.17.4 Tests and InspectionsVisual inspection of welds and hydrostatic testing of the added piping and valves shall be performed in accordance with ANSI B31.1-1977.
Operating procedures for the RCSLMS require that the system be calibrated prior to downdrain to ensure accurate level indication.
WSES-FSAR-UNIT-3 TABLE 5.4-1 (Sheet 1 of 2) Revision 14 (12/05)
REACTOR COOLANT PUMP PARAMETERS Parameter ValueNumber of Pumps 4 Type Vertical, controlled leakage, centrifugal Shaft seals, type, quantity Mechanical, 4 Byron Jackson SU seal:
Materials, stationary face Carbon A GR CCP-72 or US Graphite Graphitar G114 Rotating face body ASTM-A-351 Gr CF8 Rotating face ring Titanium carbide, Kennametal K-162 B Byron Jackson N-9000 seal Materials, stationary face Morganite CNFJ or US Graphite Graphitar G114 Rotating Face Tungsten Carbide, Kennametal KZ-801 Atomic Energy of Canada Limited (AECL) CAN4 seal: Materials, stationary face Zr, -2.5 Wt. % Nb Carbon Graphite Rotating face body XM-19, annealed (Nitronic 50)
Rotating face ring Tungsten Carbide Design pressure, psig 2485 Design temperature, F 650 Normal operating pressure, psig 2235 (DRN 03-2059, R14)
Normal operating temperature, F 543 Rated flow, gpm (@ 553F) 99,000 (DRN 03-2059, R14)
WSES-FSAR-UNIT-3 TABLE 5.4-1 (Sheet 2 of 2)REACTOR COOLANT PUMP PARAMETERSParameterValue MotorVoltage, volts6600Frequency, Hz/60/3Horsepower/speed, hot hp/rpm7800/1183Horsepower/speed, cold, hp/rpm9700/1183Service factor1.0 WSES-FSAR-UNIT-3 TABLE 5.4-2 (Sheet 1 of 2) Revision 307 (07/13) (EC-8458, R307)
STEAM GENERATOR PARAMETERS (a)(b) (EC-8458, R307)
Parameter Value Number of units 2 (DRN 03-2059, R14; EC-8458, R307)
Heat transfer rate, each, Btu/hr.
6.3793 x 10 9 (EC-8458, R307)
Primary side Design pressure/temperature (psig/ F) 2485/650 (DRN 06-1060, R15; EC-8458, R307)
Coolant inlet temperature, F 602 (DRN 06-1060, R15; EC-8458, R307)
Coolant outlet temperature, F 543 (DRN 06-1060, R15; EC-8458, R307)
Coolant flow rate, each, lb/hr 81.8 x 10 6 (DRN 03-2059, R14; 06-1060, R15)
Coolant volume at 68 F each, ft 3 2051 (EC-8458, R307) Tube size, OD, in.
3/4 (EC-8458, R307) Tube thickness, nominal, in.
0.044 for Rows 1 and 2 0.043 for rows 3 to 138 (EC-8458, R307)
Secondary side Design pressure/temperature (psia/ F) 1100/560 (DRN 03-2059, R14; 06-1060, R15; EC-8458, R307) Steam pressure, psia 837.2 (DRN 06-1060, R15)
Steam flowrate (at 0.10% moisture) each, lb/hr 8.3 x 10 6 (EC-8458, R307) (DRN 06-1060, R15; EC-8458, R307)
Feedwater temperature at full power, F 449.8 (DRN 06-1060, R15)
Moisture carryover, weight maximum, % 0.10 (DRN 03-2059, R14; EC-8458, R307)
Primary inlet nozzle, No./ID, in. 1/42 (DRN 03-2059, R14; EC-8458, R307)
_______________________________ a. Based on full power post-EPU (Extended Power Uprate) conditions with Replacement Steam Generators. b. The parameters presented in this table are based on a feedwater temperature of 449.8°F. The best estimate feedwater temperature is 447.7°F. An evaluation was performed to assess the impact of the difference in feedwater temperatures which concluded the difference between the two values is
insignificant. (DRN 03-2059, R14)
(DRN 06-1060, R15) (DRN 06-1060, R15; EC-8458, R307)
WSES-FSAR-UNIT-3 TABLE 5.4-2 (Sheet 2 of 2) Revision 307 (07/13)
(EC-8458, R307)
STEAM GENERATOR PARAMETERS (a)(b) (EC-8458, R307)
Parameter Value Primary outlet nozzle, No./ID, in. 2/30 (EC-8458, R307) Steam nozzle, No./ID, in.
1/34 (EC-8458, R307) Feedwater nozzles, No./size/schedule 1/18/80 (DRN 03-2059, R14; EC-8458, R307) Overall heat transfer coefficient (nominal) 1436 (EC-8458, R307)
Btu/hr-ft 2 - F Blowdown flow (gpm nominal, per SG) 165 (DRN 03-2059, R14)
WSES-FSAR-UNIT-3 TABLE 5.4-3 Revision 307 (07/13)
STEAM GENERATOR FABRICATION TESTING Component Test (a)
Tube sheet
Primary head (EC-8458, R307)
Forging UT, MT (EC-8458, R307)
Cladding UT, PT Secondary shell and head (EC-8458, R307)
Forging UT, MT (EC-8458, R307)
Tubes UT, ET Nozzles (forging)
UT, MT Studs UT, MT Welds (EC-8458, R307) Shell, circumferential RT, MT, UT ISI welds only (EC-8458, R307)
Cladding UT, PT (EC-8458, R307) Feedwater Nozzles to shell RT, MT, UT (EC-8458, R307) Tube-to-tube sheet PT Instrument connections MT (EC-8458, R307) All accessible welds - after hydrostatic test MT Primary side pressure nozzles PT Pedestal to Head RT, MT (EC-8458, R307)
- a. UT = Ultrasonic testing MT = Magnetic-particle testing RT = Radiographic testing (EC-8458, R307) ET = Eddy-current testing PT = Dye-penetrant testing (EC-8458, R307)
WSES-FSAR-UNIT-3 TABLE 5.4-4 Revision 307 (07/13)
REACTOR COOLANT PIPING PARAMETERS Parameter Value Number of loops (steam generators) 2 (DRN 03-2059, R14; EC-8458, R307)
Design Flow per loop, lb/hr 81.8 x 10 6 (DRN 03-2059, R14; EC-8458, R307)
Pipe size
Reactor outlet ID/wall, in.
Pipe 42/3-3/4 w/o clad Elbow 42/4-1/8 w/o clad
Reactor inlet, ID/wall, in.
Pipe 30/3 w/o clad Elbow 30/3 w/o clad
Pump suction Elbow 30/3 w/o clad Pipe 30/2-1/2 w/o clad
Surge line, (nominal pipe size in./
schedule) 12/160
Sprayline (nominal pipe size in./ 4/120 schedule) 3/160
Design pressure, psia 2500
Design temperature, F (hot & cold legs) 650 (surge line) 700 WSES-FSAR-UNIT-3TABLE 5.4-5REACTOR COOLANT PIPING TESTSComponentTest (a)Fittings (castings)RT, PT Piping (castings)RT, PT, or MT Pipe and elbowsUT, MT Carbon steel plateUT, PT WeldsCircumferentialRT, PT, or MT Nozzles to pipe runRT, MT, UTInstrument connections to pipePT or MTCladdingUT, PTSafe ends to nozzlesRT, PTa.KeyUT = Ultrasonic testing MT = Magnetic-particle testing PT = Dye-penetrant testingRT = Radiographic testing WSES-FSAR-UNIT-3 TABLE 5.4-6 (Sheet 1 of 2) Revision 309 (06/16)
PRESSURIZER PARAMETERS
Property Parameter Design pressure, psia 2500 Design temperature, F 700 Normal operating pressure, psia 2250 Normal operating temperature, F 653 (DRN 06-904, R15)
Internal free volume, ft 3 1519 (DRN 06-904, R15)
Normal (full power) operating water volume, ft 3 800 Normal steam volume full power, ft 3 700 (DRN 00-1673, R10; 01-1361, R12;05-892, R14, LBDCR 16-017, R309) Installed heater capacity, kW (Nominal) 1350 (DRN 00-1673, R10; 01-1361, R12;05-892, R14, LBDCR 16-017, R309)
(DRN 03-1268, R13)
Spray flow, maximum, gpm 490 (DRN 03-1268, R13)
Spray flow, continuous, gpm 1.5
Nozzles
Surge line (1 ea) nominal, in. 12, schedule 160
Safety valves (3), ID, in. 6, schedule 160
Spray (1) nominal, in. 4, schedule 120 (DRN 00-1631, R10;05-892, R14, LBDCR 15-004, R309, LBDCR 16-017, R 309) Heaters (30), OD, in. 1.25 (F4 Plugged / F3 & D4 CAPPED) (DRN 00-1631, R10;05-892, R14, LBDCR 15-004, R309, LBDCR 16-017, R309)
Instruments,
Level (4) nominal, in. 3/4, schedule 160
Temperature (1) nominal, in. 1, schedule 160
Pressure (2) nominal, in. 3/4, schedule 160
Dimensions
Overall length, including skirt and spray nozzle, in. 441
Outside diameter, in. 106-1/2
Inside diameter, in. 96
Cladding thickness, in. (minimum) 1/8 WSES-FSAR-UNIT-3 TABLE 5.4-6 (Sheet 2 of 2) Revision 15 (03/07)
PRESSURIZER PARAMETERS
Property Parameter (DRN 06-904, R15)
Dry weight, including heaters, lb 203,288 Flooded weight, including heaters, lb 297,712 (DRN 06-904, R15)
WSES-FSAR-UNIT-3TABLE 5.4-7PRESSURIZER TESTSComponentTest(a)HeadsPlatesUT, MT CladdingUT, PT ShellPlatesUT, MT CladdingUT, PTHeatersTubingUT, PT Centering of elementsRT End plugUT, PTNozzle (Forgings)UT, MTStudsUT, MTWeldsShell longitudinalRT, MT, UT Shell circumferentialRT, MT, UT CladdingUT, PT NozzlesRT, MT Nozzle safe endsRT, PT Instrument connectionsPT Support skirtMT Temporary attachment after removalMTAll welds after hydrostatic testMT or PTHeater assembly, end plug weldRT, PTa.Key:UT = ultrasonic testing, MT = magnetic particle testing,PT = dye-penetrant testing, RT = radiographic testing WSES-FSAR-UNIT-3TABLE 5.4-8QUENCH TANK PARAMETERSPropertyParameterDesign pressure, psig (internal/external)130/15Design temperature, °F ,350Normal operating pressure, psig3Normal operating temperature, °F120Minimum internal volume, gal2400Blanket gasNitrogenNozzlesPressurizer discharge (1) nominal, in.12 Sch. 40 SDemineralized water (1), in./rating2/3000 lb SW Coupling Rupture disc (1) in.20 flanged Drain (1), in./rating2/3000 lb SW Coupling Temperature instrument (1), in./rating1/3000 lb SW Coupling Level instrument (2), in./rating1/3000 lb SW Coupling Vent (1), in./rating2/3000 lb SW CouplingVessel materialASTM-SA-240 Type 304 DimensionsOverall length, in.151.75 Outside diameter, in.72Dry weight, lb6700 Flooded weight, lb26,850Code and dateASME Section VIIIDiv. 1 through Summer 1970 Addenda WSES-FSAR-UNIT-3 TABLE 5.4-9 Revision 15 (03/07)
PRIMARY SAFETY VALVE PARAMETERS
Property Parameter
Design pressure, psia 2500
(DRN 06-872, R15)
Design temperature, F 700 (DRN 06-872, R15)
Fluid Saturated steam, 2000 ppm H 3 BO 3 , pH=5.0 Set pressure, psia 2500 +/-3
Capacity, lb/hr. at set pressure, each 460,000
Type Spring loaded safety-balanced bellows. Enclosed bonnet.
Accumulation, percent 3
Backpressure Max buildup/max. superimposed, psig 500/130
Blowdown, percent 5
Materials
Disc ASTM A637, GR. 688