ML19298B515
| ML19298B515 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 10/04/2019 |
| From: | Tennessee Valley Authority |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML19298B540 | List: |
| References | |
| Download: ML19298B515 (72) | |
Text
{{#Wiki_filter:BFN-26 TABLE OF CONTENTS 6.0 EMERGENCY CORE COOLING SYSTEMS 6.1 Safety Objective................................................................................................................................... 6.1-1 6.2 Safety Design Basis.............................................................................................................................6.2-1 6.3 Summary Description-Emergency Core Cooling Systems ................................................................... 6.3-1 6.4 Description ........................................................................................................................................... 6.4-1 6.4.1 High Pressure Coolant Injection System ................................................................................. 6.4-1 6.4.2 Automatic Depressurization System ....................................................................................... 6.4-7 6.4.3 Core Spray System ................................................................................................................. 6.4-8 6.4.4 Low Pressure Coolant Injection System .................................................................................. 6.4-11 6.5 Safety Evaluation ................................................................................................................................. 6.5-1 6.5.1 Summary ................................................................................................................................. 6.5-1 6.5.2 Performance Analysis ............................................................................................................. 6.5-2 6.5.3 Integrated Operation of the Emergency Core Cooling Systems .............................................. 6.5-10 6.5.4 Emergency Core Cooling Systems Redundancy .................................................................... 6.5-15 6.5.5 Potential Plugging of Emergency Core Cooling System Suction Strainers ............................. 6.5-15 6.6 Inspection And Testing ........................................................................................................................ 6.6-1 6.0-i
BFN-28 EMERGENCY CORE COOLING SYSTEMS LIST OF TABLES Table Title 6.3-1 Emergency Core Cooling Systems Equipment Design Data Summary 6.4-1 (Deleted) 6.5-1 Plant Parameters Used in LOCA Analysis 6.5-2 ECCS Equipment Capacity Assumed in LOCA Analysis 6.5-3 Single Failure Evaluation Used for LOCA Analysis 6.5-4 Sequence of Events for Limiting Recirculation Discharge Line Break Battery Failure (SF-BATTlBA) (EXEM BWR-2000 LOCA Methodology (at 3952 MWt) 6.5-5 (Deleted) 6.5-6 (Deleted) 6.5-7 (Deleted) 6.0-ii
BFN-26 EMERGENCY CORE COOLING SYSTEMS LIST OF ILLUSTRATIONS Figure Title 6.3-1 (Deleted) 6.4-1 HPCI System - Mechanical Control Diagram 6.4-2 Core Spray System, Flow Diagram 6.4-3 HPCI System - Mechanical Control Diagram 6.4-4 Core Spray System - Flow Diagram 6.4-5 HPCI System - Mechanical Control Diagram 6.4-6 Core Spray System - Flow Diagram 6.5-1 (Deleted) 6.5-2 (Deleted) 6.5-3 (Deleted) 6.5-4 (Deleted) 6.5-5 (Deleted) 6.5-6 (Deleted) 6.5-7 (Deleted) 6.5-8 (Deleted) 6.5-9 (Deleted) 6.5-10 (Deleted) 6.5-11 (Deleted) 6.5-12 (Deleted) 6.5-13 (Deleted) 6.5-14 (Deleted) 6.5-15 (Deleted) 6.5-16 (Deleted) 6.5-17 (Deleted) 6.0-iii
BFN-26 EMERGENCY CORE COOLING SYSTEMS LIST OF ILLUSTRATIONS (Cont'd) Figure Title 6.5-18 (Deleted) 6.5-19 (Deleted) 6.5-20 (Deleted) 6.5-21 (Deleted) 6.5-22 (Deleted) 6-5-23 (Deleted) 6.5-24 (Deleted) 6.5-25 (Deleted) 6.5-26 (Deleted) 6.5-27 (Deleted) 6.5-28 (Deleted) 6.5-29 (Deleted) 6.5-30 (Deleted) 6.5-31 (Deleted) 6.5-32 (Deleted) 6.5-33 (Deleted) 6.5-34 (Deleted) 6.5-35 (Deleted) 6.5-36 (Deleted) 6.5-37 (Deleted) 6.5-38 (Deleted) 6.5-39 (Deleted) 6.5-40 (Deleted) 6.5-41 (Deleted) 6.0-iv
BFN-28 EMERGENCY CORE COOLING SYSTEMS LIST OF ILLUSTRATIONS (Cont'd) Figure Title 6.5-42 (Deleted) 6.5-43 (Deleted) 6.5-44 (Deleted) 6.0-v
BFN-21 6.0 EMERGENCY CORE COOLING SYSTEMS 6.1 SAFETY OBJECTIVE The objective of the Emergency Core Cooling Systems (ECCS), in conjunction with the primary and secondary containments, is to limit the release of radioactive materials to the environs following a loss-of-coolant accident, so that resulting radiation exposures are kept to a practical minimum and are within the guideline values given in 10 CFR 50.67. 6.1-1
BFN-22 6.2 SAFETY DESIGN BASIS
- 1. To provide adequate cooling of the reactor core under abnormal and accident conditions, various cooling systems shall be provided of such number, diversity, reliability, and redundancy that only a highly improbable combination of events could result in inadequate cooling of the core.
- 2. In the event of a loss-of-coolant accident, the Emergency Core Cooling Systems shall remove the residual stored heat and heat from radioactive decay from the reactor core at such a rate that fuel clad melting is prevented and any core mechanical deformation does not limit effective cooling of the reactor core.
- 3. The Emergency Core Cooling Systems shall provide for continuity of core cooling over the complete range of postulated break sizes in the nuclear system process barrier.
- 4. Emergency Core Cooling Systems shall be initiated automatically by conditions which sense the potential inadequacy of core cooling to limit the degree to which safety is dependent upon operator judgment in a time of stress.
- 5. Operation of the Emergency Core Cooling Systems shall be initiated regardless of the availability of offsite power supplies and the normal generating system of the plant. The system shall be able to accommodate any combination of real or spurious accident signals from one of the other two units (real coincident with a spurious signal, real followed by a spurious signal, and spurious followed by a real accident signal).
- 6. Action taken to affect containment integrity shall not negate the ability to achieve core cooling.
- 7. To provide assurance that the Emergency Core Cooling Systems shall operate effectively, each component required to operate in a loss-of-coolant accident shall be testable during normal operation of the nuclear system. The inboard isolation check valves can only be tested during cold shutdown (MODE 4 or 5).
- 8. The components of the Emergency Core Cooling Systems within the reactor vessel shall be designed to withstand the transient mechanical loadings during a loss-of-coolant accident, so that the required core cooling flow is not restricted.
6.2-1
BFN-22 9.The equipment of the Emergency Core Cooling Systems shall withstand the physical effects of a loss-of-coolant accident so that the core can be effectively cooled. These effects are missiles, fluid jets, high temperature, pressure, humidity, and radiation.
- 10. The Emergency Core Cooling Systems shall be capable of withstanding earthquake ground motions without impairment of their functions.
- 11. To provide a reliable supply of water for the Emergency Core Cooling Systems, the qualified source of liquid for cooling the reactor core after a loss-of-coolant accident shall be a stored source located within the primary containment. The source shall be located in the primary containment in such a manner that a closed cooling water path is established during operation of the Emergency Core Cooling Systems.
6.2-2
BFN-17 6.3
SUMMARY
DESCRIPTION - EMERGENCY CORE COOLING SYSTEMS During normal operations, when normal electrical power for the plant auxiliaries is available, heat is removed from the reactor core through the boiling-water, steam-turbine, condenser-feedwater cycle during power operation, or during shutdown through use of the Residual Heat Removal System (RHRS). For postulated accident conditions, coolant is lost from a breach in the nuclear process system. The reactor is shut down by a low reactor water level or high drywell pressure scram. High drywell pressure plus low reactor vessel pressure, or various reactor vessel low-water-level signals, will automatically start one or more Emergency Core Cooling Systems to maintain core cooling. As the water level in the reactor vessel continues to drop, the main steam line isolation valves are automatically closed on a low-low-low reactor water level signal. The Emergency Core Cooling Systems are filled at all times to prevent the possibility of water-hammer. The LPCI and Core Spray Systems are filled from the head tank which has ties to the pressure suppression chamber water-transfer pumps for pumping power, or from the condensate transfer system. The HPCI System is filled from the condensate storage tank. The head tank to ECCS to torus to transfer pumps back to head tank system assures that the proper torus water level is maintained. Two Class D check valves exist between the ECCS and the head tank and between the ECCS and the Condensate Storage and Supply System to prevent loss of ECCS flow. The transfer pumps are redundant and have separate power supplies. The instrumentation associated with the system is redundant. Also, the discharge piping of the RHRS and Core Spray are periodically vented from the high point of the system and water flow determined in accordance with technical specifications surveillance frequency requirements. Verification that the systems are vented and filled is also required by plant procedures following periods of inoperability or maintenance on these systems as required to demonstrate system operability. The Emergency Core Cooling Systems (ECCS) consist of the following: High Pressure Coolant Injection System (HPCI), Automatic Depressurization System, Core Spray System, and Low Pressure Coolant Injection System (LPCI), an operating mode of the RHR. The Emergency Core Cooling Systems are designed to limit clad temperature over the complete spectrum of possible break sizes in the nuclear system process barrier, including the design basis break. The design basis break is defined as the complete and sudden circumferential rupture of the largest pipe connected to the reactor vessel (i.e., one of the recirculation loop pipelines) with displacement of the ends so that blowdown occurs from both ends. 6.3-1
BFN-17 The individual Emergency Core Cooling Systems (HPCI, Automatic Depressurization System, Core Spray System, and LPCI) are described in the following paragraphs. A summary of the principal design parameters of the ECCS--core cooling capacity, flow, pressure, and backup systems--is given in Table 6.3-1. See Section 6.5 for system flow requirements utilized in the ECCS analyses. Section 6.0 gives the safety analysis of the Emergency Core Cooling Systems from the integrated system operation viewpoint. Other sections of this report which give further specific details are the following: Reactor Vessel Internals (Core Spray), Subsection 3.3, Nuclear System Pressure Relief System (depressurization valves), Subsection 4.4, Residual Heat Removal System (RHRS), Subsection 4.8, and Emergency Core Cooling Control and Instrumentation, Subsection 7.4. The system flow diagrams, along with the functional control diagrams, are included in Subsection 7.4, "Emergency Core Cooling Control and Instrumentation," which also evaluates the controls and instrumentation for all of the ECCS. The equipment and operation of each of the ECCS are discussed, followed by an evaluation of the capability of the integrated ECCS operation to meet the safety design bases for the ECCS. The section concludes with a discussion of the testing and inspection which are performed to provide assurance that the ECCS will operate as required. 6.3-2
BFN-28 Table 6.3-1 EMERGENCY CORE COOLING SYSTEMS EQUIPMENT DESIGN DATA
SUMMARY
Number AC Power Installed- Design Flow Pressure Required Individual (each) Range for Source Backup Function Capacity flow psid* psig Initiation of Water Systems HPCI 1-100% 5000 gpm @ 1120**** to 150 1174**** to 150 None Condensate Auto Depress. 3600 gpm @ 1174 Storage Tank with either Core and Pressure Spray or LPCI Suppression Pool Automatic 6- 25% 865,000 lb/hr @ 1105 1169***** to 50 None N/A Remote-Manual Depressurization Main Steam System Relief Valves Core Spray 2-100%*** 6250 gpm @ 105 289 to 0 Normal Aux. Pressure Suppression LPCI or Standby Pool Diesel Gen. LPCI 4- 50% 20,000 gpm @ 0 319.5 to 0 Normal Aux. Pressure Suppression Core Spray (2 pumps per loop) or Standby Pool System 10,800 gpm @ 0 Diesel Gen. (1 pump per loop)
- psid-pounds per square inch differential between reactor vessel and primary containment
- (Deleted).
- 2-100% capacity systems each having 2-50% capacity pumps
- Full rated flow of 5000 gpm is only required between 150 and 1120 psid. See Table 6.5-2 for flow rates credited for LOCA for pressures above 1120 psid.
- 2-100% capacity systems each having 2-50% capacity pumps
- (Deleted).
Otherwise, pressures above 1120 psid are representative of conditions where HPCI is required to serve as backup to RCIC which produces significantly less flow (600 gpm).
- 1169 is calculated from the lowest MSRV setpoint + 3%.
BFN-16 Figure 6-3-1 Deleted by Amendment 9.
BFN-27
6.4 DESCRIPTION
6.4.1 High Pressure Coolant Injection System The High Pressure Coolant Injection System (HPCI) consists of a steam turbine assembly driving a constant-flow pump assembly and system piping, valves, controls, and instrumentation. The HPCI mechanical control diagram is shown in Figures 6.4-1, 6.4-3, and 6.4-5. The principal HPCI equipment is installed in the Reactor Building. The turbine-pump assembly is located in a shielded area to assure that personnel access to adjacent areas is not restricted during operation of the HPCI. Suction piping comes from the condensate supply header and the pressure suppression pool. Injection water is piped to the reactor feedwater pipe at a T-connection outside the primary containment. Steam supply for the turbine is piped from a main steam header in the primary containment. This piping is provided with an isolation valve on each side of the drywell barrier. Remote controls for valve and turbine operation are provided in the plant control room. The controls and instrumentation of the HPCI are described, illustrated, and evaluated in detail in Subsection 7.4, "Emergency Core Cooling Systems Control and Instrumentation." The HPCI is provided to assure that the reactor is adequately cooled to limit fuel cladding temperature in the event of a small break in the nuclear system and loss of coolant which does not result in rapid depressurization of the reactor vessel. The HPCI permits the nuclear plant to be shut down, while maintaining sufficient reactor vessel water inventory until the reactor vessel is depressurized. The HPCI continues to operate until the reactor vessel pressure is below the pressure at which LPCI operation or Core Spray System operation maintains core cooling. If a loss-of-coolant accident occurs, the reactor scrams upon receipt of a low-water-level signal or a high-drywell-pressure signal. The HPCI starts when the water level reaches a pre-selected height above the core, or if high pressure exists in the primary containment (drywell). The HPCI automatically stops when a high water level in the reactor vessel is signaled. The HPCI automatically resets and will restart if vessel water level again reaches a pre-selected height above the core. The HPCI is designed to pump water into the reactor vessel for a wide range of pressures in the reactor vessel. Two sources of water are available. Initially, condensate water from the condensate storage header is used instead of injecting the less desirable water from the pressure suppression pool into the reactor. This provides reactor-grade water to the reactor vessel for the case where the need for the HPCI is rapidly satisfied. The condensate water from the normally assigned condensate storage tank for each unit is sufficient for maintaining an adequate 6.4-1
BFN-28 reactor coolant inventory for 5.5 hours at hot shutdown conditions (MODE 3) (non-accident). The condensate tanks for all three units may also be aligned to provide condensate supply to any individual unit. In addition, two spare 500,000-gallon condensate storage tanks may be aligned to supply HPCI on any unit. When the level in the condensate header falls below a predetermined setpoint or the level in the pressure suppression pool increases to a predetermined setpoint, the HPCI pump suction is automatically transferred to the pressure suppression pool. With HPCI taking suction from the condensate storage tank and injecting to the reactor vessel, there is sufficient inventory in the tank such that the high suppression pool level suction transfer will occur before a low condensate header level would be created. This transfer may also be made from the control room using remote controls. This establishes a closed loop for recirculation of water escaping from a break and satisfies safety design basis 11. Water from either source is pumped into the reactor vessel via the feedwater line. Flow is distributed within the reactor vessel through the feedwater spargers to obtain mixing with the hot water or steam in the reactor pressure vessel. The pump assembly is located below the level of the condensate header and below the water level in the pressure suppression pool to assure positive suction head to the pumps. Pump NPSH requirements are met by providing adequate suction head and adequate suction line size. The water supply lines for the HPCI are also shown on Figures 6.4-1, 6.4-3, 6.4-5, and 7.4-1b sheets 1, 2, and 3. Trapped air is initially removed with vent lines connected to the high points of the HPCI System. Whenever HPCI is lined up to take suction from the condensate storage tank, the discharge piping of the HPCI is periodically vented from the high point of the system and water flow observed in accordance with Technical Specifications surveillance frequency requirements for system operability. The formation of air pockets in the water supply line can create water-hammer problems upon initiation of the systems. Maintaining the pump discharge lines of the HPCI System full of water ensures that the ECCS will perform properly upon demand. This will also prevent water hammer following an ECCS initiation signal. One acceptable method of ensuring that the lines are full is by venting the system. Constant hydrostatic overpressure on the water supply line in the HPCI System prevents air pockets from forming. 6.4-2
BFN-28 The HPCI turbine-pump assembly and piping are located so as to be protected from the physical effects of design basis accidents such as pipe whip and high temperatures. The equipment is located outside the primary containment. This arrangement satisfies safety design basis 9. The HPCI turbine is driven by steam from the reactor which is generated by decay heat and residual heat. The steam is extracted from a main steam header upstream of the main steam line isolation valves. The two HPCI isolation valves in the steam line to the HPCI turbine are normally open in order to keep the piping to the turbine pressurized to permit rapid startup of the HPCI. To permit a controlled warmup of the steam line following a HPCI maintenance outage, a motor-operated bypass valve around the outboard containment isolation valve is provided. Signals from the HPCI control system open or close the turbine stop valve. A condensate drain pot is provided upstream of the turbine stop and HPCI steam supply valves to prevent the HPCI steam supply line from filling with water. The drain pot normally routes the condensate to the main condenser, but upon HPCI initiation isolation valves on the condensate drain pot line automatically shut. Condensate in the supply lines during this time will discharge via drain lines from the turbine stop valve seat, steam chest areas, and turbine casing exhaust drains. The HPCI pump set flow rate is controlled by a speed governor which controls the turbine speed in response to a demand signal from the flow controller. This arrangement allows for a constant HPCI flow rate over the range of discharge pressures which HPCI must operate. As reactor steam pressure decreases, the HPCI turbine throttle valve is modulated by the governor control system to control turbine steam flow in order to provide the required pump flow called for by the flow controller. The capacity of the system is selected to provide sufficient core cooling to limit cladding temperatures while the pressure in the reactor vessel is above the pressure at which core spray and LPCI become effective. Exhaust steam from the HPCI turbine is discharged to the pressure suppression pool. A drain pot at the low point in the exhaust line collects moisture present in the steam. Collected moisture is discharged to the gland seal condenser. Therefore, condensate will not collect in the turbine exhaust line and the turbine drain line to create steam compression problems upon initiation of these systems. After operation of the HPCI system, the HPCI turbine exhaust drain line to the gland seal condenser may be full of water. A globe stop check valve stops the backflow of suppression pool water into the turbine through the turbine exhaust line. 6.4-3
BFN-27 The HPCI turbine gland seals are vented to the gland seal condenser and part of the water from the HPCI pump is routed through the condenser for cooling purposes. To prevent over-pressurization of the gland seal condenser, an orifice and relief valve are provided upstream of the condenser. Noncondensable gases from the gland seal condenser are pumped to the Standby Gas Treatment System. The system piping material is carbon steel and is designed to USAS B31.1.0, 1967 edition. The pumps are designed to the ASME Boiler and Pressure Vessel Code, Section III, Class C, 1965 edition. The pumps are also designed and tested in accordance with the standards of the Hydraulic Institute. A condensing sparger is attached to the HPCI turbine exhaust line to reduce the noise made by the exhausting steam and improve turbine exhaust pressure stability. The sparger is capped at the bottom end and contains holes drilled through the walls for discharging the steam. The uppermost row of holes is under approximately two feet of water. Additional holes are spaced at regular intervals down to the capped ends. The lower row of holes is under approximately eight feet of water. For the HPCI there are 1296 holes, 1 inch in diameter. The ratio of total hole area to the cross sectional area of the pipe is about 2.5 to 1 (the discharge path area available in the sparger design is about 2.5 times that in the open pipe design). A vacuum breaker is installed on the HPCI turbine exhaust lines to prevent unacceptable vibration and movement of the line. The breaker prevents intermittent negative pressure in this piping run from pulling water out of the torus and causing a water-hammer problem. The Unit 1 and Unit 3 vacuum breaker consists of 2-inch check valves with connecting piping and the Unit 2 vacuum breaker consists of 3-inch check valves with connecting piping which allows torus air to enter the exhaust line whenever negative pressure exists in that line. To ensure operability and at the same time prevent steam bypassing the pressure suppression pool due to a check valve sticking open, four check valves are installed on the system in a one-out-of-two-twice configuration. Thus, the vacuum breaker piping will be arranged with a common line penetrating the torus wall that includes a normally open manual valve and a normally open motor-operated valve (MOV) in series. Downstream of the MOV on Unit 1, the line branches with separate lines going to the HPCI and to the RCIC vacuum breaker connected to the turbine exhaust lines outside the torus, but downstream of the respective stop check valves. Units 2 and 3 have separate HPCI and RCIC turbine exhaust vacuum breaker lines. This arrangement is shown on Figures 6.4-1, 6.4-3, and 6.4-5. Electrical power to operate the MOV is normally disconnected and indication is provided in the Main Control Room to warn of any inadvertent connection. Controls 6.4-4
BFN-27 and electrical power for the valve are located where they are accessible following a LOCA. The HPCI equipment, piping, and support structures are designed Seismic Class I equipment (see Appendix C). This satisfies design basis 10. Assuming the system is tested every two months, the system is designed for a service life of 40 years, accounting for corrosion, erosion, and material fatigue. The following Aging Management Programs (see Appendix O for summary descriptions) are used to manage the effects of the HPCI for the 60 year operating life: ASME Section XI Subsections IWB, IWC, and IWD Inservice Inspection Program (Section O.1.4) Chemistry Control Program (Section O.1.5) BWR Stress Corrosion Cracking Program (Section O.1.10) Flow-Accelerated Corrosion Program (Section O.1.14) Bolting Integrity Program (Section O.1.15) One-Time Inspection Program (Section O.1.26) Selective Leaching of Materials Program (Section O.1.27) Systems Monitoring Program (Section O.2.1) Material fatigue for the 60 year operating life has been evaluated as a Time Limited Aging Analysis (TLAA). The summary of this evaluation is provided in Appendix O, Sections O.3.2.3 and O.3.2.4. Startup of the HPCI is completely independent of AC power. Only DC power from the plant batteries and steam extracted from the nuclear system are necessary. This satisfies safety design basis 5. The various operations of the HPCI components are summarized as follows. The HPCI controls automatically start the system and bring it to design flow rate within 30 seconds (see Section 6.5 for value assumed in Emergency Core Cooling System analyses) from receipt of a reactor vessel low-low-water-level signal or a primary containment (drywell) high-pressure signal. The HPCI turbine is shut down automatically by any of the following signals:
- a. Turbine overspeed--This prevents damage to the turbine and turbine casing.
- b. Reactor vessel high water level--This indicates that core cooling requirements are satisfied.
6.4-5
BFN-27
- c. HPCI pump low suction pressure--This prevents damage to the pump due to loss of flow.
- d. HPCI turbine exhaust high pressure--This indicates a turbine or turbine control malfunction.
- e. Automatic isolation signal--This indicates a HPCI steam line rupture or rupture disc failure.
- f. Low steam supply pressure.
If an initiation signal is received after the turbine is shut down, the system is capable of automatic restart if no shutdown signals exist. Because the steam supply line to the HPCI turbine is part of the nuclear system process barrier, certain signals automatically isolate this line, causing shutdown of the HPCI turbine. The system is not capable of automatic restart following isolation of the steam supply line. Automatic shutoff of the steam supply is described in Subsection 7.3, "Primary Containment and Reactor Vessel Isolation Control System." However, automatic depressurization and the low pressure systems of the ECCS act as backup, and automatic shutoff of the steam supply does not negate the ability of the ECCS to satisfy the safety objective. In addition to the automatic operational features of the system, provisions are included for remote-manual startup, operation, and shutdown (provided initiation or shutdown signals do not exist). During a HPCI initiation, the following valves automatically actuate: HPCI pump CST flow test throttle and block valves HPCI pump CST suction valve (if closed) HPCI pump injection valve HPCI pump discharge isolation valve (if closed) HPCI turbine stop and control valves HPCI turbine steam supply valve HPCI gland seal condenser drain line isolation valves (close if open) HPCI turbine inlet drain pot isolation valves HPCI minimum flow valve Startup of the auxiliary oil pump and proper functioning of the lube oil system are required to open the turbine stop and control valves. Operation of the gland seal condenser is required to ensure that turbine seal outleakage is minimized. Gland seal condenser startup is automatic and its failure does not prevent the HPCI System from performing its core cooling objective. When the turbine stop valve 6.4-6
BFN-27 begins to open, it signals the speed governor to begin ramping turbine speed up. Once the speed governor ramp signal exceeds the flow controller demand signal, turbine control is switched to the flow controller. The flow controller maintains the HPCI flow rate at a constant rate as reactor pressure fluctuates. A minimum flow bypass is provided for pump protection. Following an initiation signal, the bypass valve automatically opens due to a low-flow signal, and automatically closes on a high-flow signal or a HPCI turbine trip. When the bypass is open, flow is directed to the pressure suppression pool. A full-flow test line is provided to verify system operation. The line directs flow to the condensate storage tank when the test return line throttle and block valves are open. These valves are sequenced to close by the signal which actuates system operation and are interlocked closed when either suction valve from the pressure suppression pool is open. As previously stated, the HPCI System is not dependent upon AC power for operation. In the event of a loss-of-all AC power, the time limits for the operation of HPCI would be dependent upon the power available from the 250-V batteries which supply the DC power to the HPCI system. Best-estimate analyses have been performed for an assumed scenario of loss-of-all AC power and it is indicated that the batteries have sufficient capacity for 4 hours. 6.4.2 Automatic Depressurization System In case the capability of the HPCI is not sufficient to maintain the reactor water level, the Automatic Depressurization System functions to reduce the reactor pressure so that flow from the LPCI and the Core Spray System enters the reactor vessel in time to cool the core and limit fuel cladding temperature. The Automatic Depressurization System uses six (6) of the nuclear system main steam relief valves to relieve the high pressure steam to the pressure suppression pool. Per Subsection 4.4, between four and six ADS valves are required to meet the requirements for ADS dependent upon the recognized single failure. As shown in Table 6.5-3, due to a limiting failure of a battery which results in the loss of the normal power supply to ADS logic, Units 1, 2, and 3 are designed with an automatic transfer scheme to provide four automatic ADS valves. The design, description, and evaluation of the main steam relief valves are discussed in detail in Subsection 4.4, "Nuclear System Pressure Relief System," and it is shown that safety design bases 5, 9, and 10 are satisfied. Discharge pressure indication of one LPCI pump or two core spray pumps combined with one of the following initiation paths will cause the ADS main steam relief valves to open: (1) reactor vessel low water level and primary containment (drywell) high pressure in conjunction with a 120 seconds timer timed out; or (2) sustained reactor low water level for 360 seconds. The time delay provides time for the operator to 6.4-7
BFN-27 cancel the automatic depressurization signal if control room information indicates the signal is false or is not needed. The controls and instrumentation of the Automatic Depressurization System are given in Subsection 7.4, "Emergency Core Cooling System Controls and Instrumentation." 6.4.3 Core Spray System Two independent loops are provided as a part of the Core Spray System. Each loop consists of two 50 percent-capacity centrifugal pumps driven by electric motors, a spray sparger in the reactor vessel above the core, piping and valves to convey water from the pressure suppression pool to the sparger, and the associated controls and instrumentation. Figures 6.4-2, 6.4-4, and 6.4-6 show a flow diagram of the Core Spray System. In the case of low-low-low water level in the reactor vessel or high pressure in the drywell plus low reactor vessel pressure, the Core Spray System, when reactor vessel pressure is low enough, automatically sprays water onto the top of the fuel assemblies in time and at a sufficient flow rate to cool the core and limit fuel cladding temperature. (The Low Pressure Coolant Injection System starts from the same signals and operates independently to achieve the same objective by flooding the reactor vessel.) The Core Spray System provides protection to the core for the large break in the nuclear system when the control rod drive water pumps, RCIC, and the HPCI are unable to maintain reactor vessel water level. The protection provided by the Core Spray System also extends to a small break in which the control rod drive water pumps, RCIC, and HPCI are all unable to maintain the reactor vessel water level and the Automatic Depressurization System has operated to lower the reactor vessel pressure so LPCI and the Core Spray System can provide core cooling. The core spray pumps for each unit receive power from the plant 4160-V shutdown boards. Each core spray pump motor and the associated automatic motor-operated valves for one unit receive AC power from different buses. Similarly, control power for each pump of the Core Spray System for one unit comes from different DC buses. This arrangement satisfies design basis 5 (see Subsection 8.5, "Standby AC Power Supply and Distribution," and 8.6, "250-V DC Power Supply and Distribution"). The core spray pumps and all automatic valves can be operated individually by manual switches in the control room. Operating information is provided in the control room with pressure indicators, flow meters and indicator lights. 6.4-8
BFN-27 The major equipment for one loop is described in the following paragraphs. When the system is actuated, water is taken from the pressure suppression pool. Flow then passes through a normally open motor-operated valve in the suction line to each 50 percent pump. Each valve can be closed by a remote-manual switch from the control room to isolate the system from the pressure suppression pool in the case of a leak from the Core Spray System. This valve, which is normally open, is located in the core spray pump suction line as close to the pressure suppression pool as practical. A local pressure gauge by each pump indicates the presence of a suction head for the pump. The core spray pumps are located in the Reactor Building below the water level in the pressure suppression pool to assure positive pump suction. The pumps, piping, controls, and instrumentation of each loop are separated and protected so that any single physical event, or missiles generated by rupture of any pipe in any system within the containment drywell, cannot make both core spray loops inoperable. The switchgear for each loop is in a separate cabinet for the same reason. This arrangement satisfies safety design basis 9. A shaft seal drain line is provided from the seal housing to the Radwaste System. A normally closed test valve is connected to the line in order to provide a means for measuring seal leakage during primary containment leak rate testing. A low flow bypass-line is provided from the pump discharge to below the surface of the pressure suppression pool. The bypass flow is required to prevent the pump from overheating when pumping against a closed discharge valve. Two orifices in series limit the bypass flow. A manual valve normally open is used to close the bypass line for maintenance. The piping diagram for the Core Spray System is shown on Figures 6.4-2, 6.4-4, and 6.4-6. Flush and drain connections on the high and low points (flush and drain systems not shown) permit solid filling of these systems. A hydrostatic overpressure is maintained on the system through two check valves and a normally open hand valve. Thus, water-hammer problems are avoided in the Core Spray System by periodically venting the discharge lines while they are under a hydrostatic pressure in order to eliminate any air pockets which might form. Discharge line check valve seat leakage from hydrostatic pressure will drain into the torus. A relief valve, set for 500 psig, protects the low pressure Core Spray System upstream of the outboard shutoff valve from reactor pressure. The relief valve discharges to the equipment drain sump and thence to the Radwaste System. 6.4-9
BFN-27 A full-flow test line permits circulating water to the pressure suppression pool for testing the system during normal plant operations. A normally closed, motor-operated valve in the line is controlled by a remote-manual switch in the control room. Orifices and partial opening of the valve in the test line provide rated core spray flow at a pressure drop equivalent to discharging into the reactor vessel. A restricting orifice is provided to minimize vibration in the pump test lines. A flow indicator in the control room signals that water is or is not flowing to the core spray sparger or test line. Two motor-operated valves are provided to isolate the Core Spray System from the nuclear system when the core spray pumps are not running. These valves admit core spray water to the reactor when signaled to open. Both valves are installed outside the drywell to facilitate operation and maintenance, but as close as practical to the drywell to limit the length of line exposed to reactor pressure. The valve nearer the containment is normally closed to back up the inside check valve for containment purposes. The outboard valve is normally open to limit the equipment needed to operate in an accident condition. By closing the outboard valve, the inboard valve can be operated for test with the reactor vessel pressurized. A drain line is provided between the two shutoff valves to measure leakage through the inside check valve or the inboard shutoff valve. The drain line is normally closed with two valves. A check valve is provided in each core spray pipeline just inside the primary containment to prevent loss of reactor coolant outside containment in case the core spray line breaks. A normally open manual valve is provided downstream of the inside check valve to shut off the Core Spray System from the reactor during shutdown conditions for maintenance of the upstream valves. The two Core Spray System pipes enter the reactor vessel through nozzles 120 degrees apart. Each internal pipe then divides into a semicircular header with a downcomer at each end, which turns through the shroud near the top. A semicircular sparger is attached to each of the four outlets to make two practically complete circles, one above the other. Short elbow nozzles are spaced around the spargers to spray the water radially onto the tops of the fuel assemblies. Core spray piping upstream of the outboard shutoff valve is designed for the lower pressure and temperature of the core spray pump discharge and is fabricated from carbon steel. The outboard valve and piping downstream are designed for reactor vessel pressure and temperature. The high pressure piping portion of the system is designed to USASI B31.1.0, 1967 edition. Material for the portion of the system piping inside the drywell to the second isolation valve is high toughness (A333 Grade 6) carbon steel and type 304 stainless steel. 6.4-10
BFN-27 The core spray equipment, piping, and support structures are designed in accordance with Class I seismic criteria (see Appendix C) to resist the motion at the installed location within the supporting building from the Design Basis Earthquake. The Core Spray System is assumed filled with water for seismic analysis. It is concluded that safety design basis 10 is satisfied. Upon signals of reactor low-low-low water level or drywell high pressure plus low reactor vessel pressure, the automatic controls turn on the core spray pumps and restore valves to the spray mode. When reactor pressure decreases, the core spray shutoff valves are signaled to open. Flow to the sparger begins when the core spray discharge pressure is sufficient to open the injection check valve located inside the drywell. Subsection 7.4, "Emergency Core Cooling System Controls and Instrumentation," contains further details and evaluation. For Units 1 and 2, a Core Spray signal will initiate the ECCS preferred pump logic to trip any running RHR and Core Spray pumps in the opposite unit. In the event of a LOCA in conjunction with a spurious accident signal from the opposite unit, the ECCS preferred pump logic will dedicate the Division I pumps (1A and 1C) to Unit 1 and the Division II pumps (2B and 2D) to Unit 2. This will ensure that the shared Unit 1/2 4KV shutdown boards are not overloaded in the event of a real and spurious accident signal. 6.4.4 Low Pressure Coolant Injection System In case of low-low-low water level in the reactor or high pressure in the containment drywell plus low reactor vessel pressure, the Low Pressure Coolant Injection System (LPCI) mode of operation of the Residual Heat Removal System pumps water into the reactor vessel in time to flood the core and limit fuel clad temperature. (The Core Spray System starts from the same signals and operates independently to achieve the same objective.) The LPCI has been modified for Units 1, 2, and 3. This modification has been described in reports entitled "Browns Ferry Nuclear Plant Units 1 and 2 Emergency Core Cooling Systems Low Pressure Coolant Injection Modifications for Performance Improvement" (October, 1977) for Units 1 and 2 and for Unit 3. The reports were submitted to the Nuclear Regulatory Commission. The LPCI system modifications to BFNP Units 1, 2, and 3 provide a significant reduction in peak fuel clad temperature following a postulated recirculation line break. This, in turn, results in increased MAPLHGR limits. The modifications eliminate the LPCI recirculation loop selection logic and permit simultaneous injection into both recirculation loops. Additional resistance in the form of orifices in the pump discharge line was added to prevent the RHR pumps from going to a full runout flow in the event the RHR pumps are discharging to a broken recirculation loop. The orifices (and resultant increased NPSH) assure the availability of the RHR 6.4-11
BFN-27 pumps for long-term cooling. For the modified design, the LPCI injection valves and RHR cross-tie valve are normally closed. (The Unit 1 LPCI loop cross-tie valve is removed; and the corresponding cross connection is removed by a combination of a blind flange and cut, capped connections.) On receipt of an initiation signal following a recirculation line break, both LPCI injection valves are signaled to open (when the pressure interlock is met), both recirculation pump discharge valves are signaled to close when the reactor pressure decreases to 230 psig, and the LPCI flow from two RHR pumps (one LPCI loop) is directed to the unbroken recirculation loop. Flow from the other two RHR pumps (one LPCI loop) is directed to the broken recirculation loop. The LPCI logic was again modified prior to Unit 1 restart to support combined Units 1 and 2 operation. The Division I RHR (LPCI) initiation logic will only start the Division I pumps (A and C) and open the Division I inboard injection valve. The Division II RHR (LPCI) initiation logic will only start the Division II pumps (B and D) and open the Division II inboard injection valve. For Units 1 and 2, a RHR (LPCI) signal will initiate the ECCS preferred pump logic to trip any running RHR and Core Spray pumps in the opposite unit. In the event of a LOCA in conjunction with a spurious accident signal from the opposite unit, the ECCS preferred pump logic will dedicate the Division I pumps (1A and 1C) to Unit 1 and the Division II pumps (2B and 2D) to Unit 2. This will ensure that the shared Unit 1/2 4KV shutdown boards are not overloaded in the event of a real and spurious accident signal. The pumps, piping, controls and instrumentation of the LPCI loops are separated and protected so that no single failure or any single physical event for which LPCI operation is required can make both loops inoperable. LPCI pumps and piping equipment are described in detail in Subsection 4.8, "Residual Heat Removal System," which also describes the other functions served by the same pumps if not needed for the LPCI function. The portions of the RHRS required for accident protection are designed in accordance with Class I seismic criteria (see Appendix C). It is concluded that safety design basis 10 is satisfied. 6.4-12
BFN-16 Table 6.4-1 (Deleted by Amendment 13)
9so~ l-£L-O l93Lt-Z l"I L9 NOTES:
- 1. THE HIGH PRESSURE CCX>LANT INJECTION SYSTEM IS DESIGNED TO COOL THE CORE WHEN THE REACTOR VESSEL IS ISOLATED AND REACTOR FEEDWATER IS UNAVAILABLE. WHEN THE WATER LEVEL IN 1 --, 2-13oE928-2. 3 I THE REACTOR VESSEL REACHES SETPOINT OR THE DRYWELL PRESSURE REACHES SETPOINT. THE HPCI SYSTEM IS AUTOMATICALLY STARTED STEAM LINE SPACE TEMPERATURE I BY ACTIVATION OF REACTOR LEVEL SWITCHES LIS-3-58A, LIS-3-58C, (4 PLACES) - -
r---,---r---,-- 1 I
- - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - ---,----,---r---r--- - - - -
- --.,...- ---,- --,- - --,- - -- - - -- --~---------------
LIS-3-588 AND LIS-3-58D, OR ORYWELL PRESSURE SWITCHES PIS-64-58A-D (SEE 2-47E610-3-1 AND 2-47E610-64-1J ARRANGED IN ONE-OUT-OF-TWO-TWICE LOGIC TO OPEN VALVE FCV 6. THE SAME SIGNAL OPENS PUMP DISCHARGE VALVE FCV-73-44 ANO CLOSES DRAIN LINE VALVES FCV-73-6A, FCV-73-68, FCV-73-17A AND FCV-73-178. H
>- {Tu FRCJ-t TR-69-29 9-3 RI RI RI RI :I RIRIRIRI RI RI RI RI RI RI RI RI
- 2. A MECHANICAL SPEED GOVERNOR PREVENTS TURBINE OVERSPEED 1 2-45E620-1 73-55 73-2J 73-2K 73-2L 73-2M 73-2N 73-2P 73-2R 73-2S BY TRIPPING THROTTLE TRIP VALVE FCV-73-18 WHEN A TURBINE
..._, 73-2A
~
73-28
~
73-2C
~
73-2D
~ I 73-2E
~
73-2F
~
73-2G
~
73-2H
~ OVERSPEEO OCCURS. A CONTROL GOVERNOR, FIC-73-33, EQUIPPED WITH AUTOMATIC AND MANUAL MEANS FOR SPEED SET POINT CONTROL REGULATES TURBINE SPEED VIA "TURBINE SPEED CHANGER* TO r-1-----------------
I
-----------------------------------------------1-----------------1----1-----------------1----t----,
CONTROL HPCI MAIN PUMP DISCHARGE FLOW.
- 3. DELETED TO TR-69-29 t--l - - - - - - - - - - - - - - - - - L . . . . . . l - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - , I I I (2-HE610-69-1.G2~-+-- _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ L..l _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - - , I I 4. THE TURBINE CASING IS PROTECTED FROM OVERPRESSURE BY TWO RUPTURE DISCS. PS-73-20A THRU D DETECTS DISC RUPTURES AND RE~=WATER I ~ I A A A ANNUNCIATES ON PA-73-20 IN THE MAIN CONTROL ROOM.
CLEANUP L _ _ _J TE \ I I LOG 11 LOG I ( TE\ / TE\ ( TE\ 5. DELETED 73-55A I 173-3 Ol 173-3 cu 73-558 73-55C 73-550 6. NORMALLY MAKEUP WATER FOR THE HPCI PUMP SUCTION IS
..._, HPCI ROOM I I ..._, TORUS R11-U ..._, TORUS R10-U ..._, TORUS R9-T SUPPLIED BY THE CONDENSATE STORAGE TANK THROUGH FLOW CONTROL ELEV 519' : 1 CISS II CISS 1 9-3 ELEV 550' ELEV 550' ELEV 550' VALVE FCV-73-40. AN ALTERNATE SOURCE OF WATER IS AUTOMATICALLY OBTAINED FROM THE SUPPRESSION PCX>L THROUGH
~--------------~ I I B II B I @ -- --------~ ~--------------~ ISOLATION VALVES FCV-73-26 AND FCV-73-27 WHEN THE CONDENSATE 1
t-----r-----1 1 I ©,-"--,---©
! 9-3 r1 - - 73.:§6 FROM cNDs &
DMNRLZ WATER HEADER LEVEL DECREASES TO SETPOINT. PUMP DISCHARGE FLOW ROUTES ARE AS FOLLOWS: A. NORMALLY VIA THE FEEOWATER LINE TO THE REACTOR c::iliD_ I CISS I I l.- 9-3 WLZl-73-3A I MIMIC I ~-----.r---------~2~4~7~E12~61~0~74~1=.~,2~ THROUGH ISOLATION VALVES FCV-73-34, FCV-73-44, AND CHECK VALVE FCV-73-45.
~ . A --, ,---r-, I I 7~A r;.cl+--r.-<M2i 3 : r--------,....:~RAGE SYSTEM 9-19 8. DURING TESTING VIA TEST LINE TO CONDENSATE STORAGE THROUGH FLOW CONTROL VALVES FCV-73-35 AND FCV-73-36.
G I c1ss I ~ ~ ~ FROM AUX BOILER C. MINIMUM FLOW BYPASS TO THE SUPPRESSION PCXJL THROUGH c::iliD_
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~-3 FLOW CONTROL VALVE FCV-73-30.
- 7. IF HPCI SYSTEM OPERATION IS REQUIRED WHEN IT IS IN THE TEST MODE. THE CONTROL SYSTEM IS DESIGNED TO AUTOMATICALLY RETURN THE SYSTEM TO THE OPERATING MODE.
©,-- ~ ~
9- 3 I ZI-73-2 I I ~ -I<'.: - - I @ , I I 'H5' r-...../ HS, Zl-73-38 I 73-4DA
----© 73-18A 73-188 7~~*
73 - -<a~ci-'c! 73-4A 8. DELETED I --Ii 1 1 - - - 9. ALL INSTRUMENTS ARE PREFIXED UNIT 2 UNLESS OTHERWISE NOTED. I 9-3 ----. 7~8 r.-(_ 4 ! 2-45E714-2 f--t-- LS ti - - - - - - - - - i i (A)IL-73-188
! 2-45E779-13 f- JI "Zl-73-2 liL-73-2 I >-&- - I I 73~68 1 I LOG T 10. COMPLETE LOGIC FOR HIGH PRESSURE CCX>LANT INJECTION IS SHOWN ON MECHANICAL LOGIC DIAGRAMS LISTED IN REFERENCE DRAWINGS.
I
~ ~
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+*cc I
I TO FCV-73-26, -27 (THIS DWG, 86 f-J I L-------1. I I Zl-73-408 C5058 I t--J 11. UNIDS ON DRAWINGS ARE FOR REFERENCE ONLY AND ARE ABBREVIATED AS SHOWN IN THE EXAMPLE TO MEET SPACE CONSTRAINTS. REFER TO FCV I FCV I L I )t--{)'Fc _i ___ --4-_.....{HS\ r - - - - - - - - . . J . _ _ _ _ ..J MEL FOR COMPLETE UNIDS. ALL UNIDS ARE IN UNIT 2 UNLESS I IM1M1cl 73-2 73-3 2-45E714-2 r1 ' I 73-40 7~B I r-r---, OTHERWISE NOTED. CABLES ARE IN THE SAME SYSTEM AS THEIR FROM MAIN L ___ ,1.__...J_ ___ ,-J ~cc,_ 1--,------.J .)....IL-73-4D ! ,FROM FCV-73-26, -27 I ~S8¢T~8 ~~ \ r"i 2-45E714-2 I I I END DEVICE LEADING ZEROES SHOWN IN MEL AS PART OF THE UNID ARE NOT DEPICTED. FOR ADDITIONAL GUIDANCE REFER TO NEOP-4. r'S~TM=Lcc!NccE~B~--:::,:,.--,.; FE 2-47E610-1-1,F7 / 73-1T
',-----------t*3--e::::J~IL.,;-T - - -
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©© ZI-73-15 EXAMPLE: MEL UNID BFN-0-SHV-018-0502 DRAWING UNIO SHV-18-502 X-11 I _ I CISS 11 LOG I 9-J\r--------'---.,-.....~...- - - - - - ~ BFN-1-HS-031-0160A HS-31-160A 9-3 ~PDA 73 _ I **1L-~73-81 1 -
ZI-73-818 ZI-73-81A
~~R G ~~ R G - ._~*~
- * *73-81 o
- 2 - 4 7 E 6 1(0 - 7 3 - 2 , E 6 TO PCV 73-188 L! - - 1 < TO XM-73-19 THIS DWG
- E4 f- - - , ~ ~ 11,-1,-16~ MCC ~ 9 - 3 ~ s9-3
- 12. DELETED BFN-2-PT-003-0204C PT-3-204C
~-~"',ONDENSING I I R ------- -- -1 I CISS 11 LOG I I Zl-73-16 ~ L~ r-r-!..-i- *- 1---r-, J-5 REFERENCE DRAWINGS,
/ POT
---+----J Mcc Fcv ,-lo-*5E779-43 I (es\ ~ 9 - 3 e1J-81 c, 25-50 9-19 9-3 ~ ~ ~ ,.r:; Fcv ~ x!-< 1--- 4 F I/
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_ Z ~ A /' I< I 2-47E605-SERIES ........ MECHANICAL LAYOUT OF CONTROL BOARDS 2-47E600-SERIES ........ MECHANICAL INSTRUMENT AND CONTROLS Y f---l Z MEL .... ,,,,,,, .. ,,,,, .. MECHANICAL INSTRUMENT TABULATION HPCI
\..._.., ~ L-1---------~---.J 7 t",9 8 7~ r I 8 ,k.~s I T
CPOT-73-18 73 1A
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,l. DMNRLZ 25-50 I I 1)
LOG 11 A5161 9-3 I I
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ZI-73-18 '>cC9:_3
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ZS 73-18A 7 ~ ' ( ... 7 / T )'-- HS 73-16c
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HS 73-168
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-...- _J 73-16 -
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=:;.... I I 2-47E812-SERIES ........ FLOW DIAGRAM-HPCI 2-47E610-1-1 ........... CONTROL DIAGRAM-MAIN STEAM SYSTEM 0-47E610-2-2 ........... CONTROL DIAGRAM-CNDS & DMNRLZ WATER STORAGE SYSTEM 1~~~nn~~~ -
--.,--;--, -;A
-.--1-'-! 1- - ,
1 PS I I ZI 73 36A
- - ri 2-45E714-4 I PT 13 28
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2-47E610-3-1 ........... CONTROL DIAGRAM-REACTOR FEEDWATER SYSTEM 2-47E610-64-1 .......... CONTROL DIAGRAM-PRIMARY CONTAINMENT SYSTEM 2-47E610-69-1 .......... CONTROL DIAGRAM-REACTOR WATER CLEANUP
~~~J SYSTEM I I I I I I ,11o ::::: 25-7~*..7L-'~L~-A_~lf--rt_7~-l~.-78 7i.A5-78%.:;:;;-J i IL-73-36 y : - 3 _._, 2-730E92B-7 I ~-B2~A-3 i~ _J 25~ SM 7~~;[&µ-~ ...~-1---------::::---, 5 ' _ , 6,:5 2-47E610-71-1 & -2 ..... CONTROL DIAGRAM-RCIC SYSTM 0-47E800-2 ............. MECHANICAL-SYMBOLS & FLOW DIAGRAM DRAWING
* --,I - ~ I I '1 ~
7YB (7YDPS' t-~MCC : r- 4@'-1--+---;,-,2-730E928-3, -41 i 7')f: 1--- 7'T9 7b9 ~ ~ ~ 1&J- ~ ~ INDEX 0-47E610-12-1 ,, .. ,, ,, , .CONTROL DIAGRAM-AUXILIARY BOILER SYSTEM
'<;s_J _ _ _ _ _ _ _ _ -' 2-47E852-2 ............. FLOW DIAGRAM-CLEAN RADWASTE &
DECONTAMINATION DRAINAGE (CRW) 2-47E2610-74-1 & -2 .... CONTROL DIAGRAM-RESIDUAL HEAT REMOVAL l 2 - 45 E614 - 10 f----,~_J I
'-@) * ~c _J ~-63 9-3 Jr~L~OG~J I (73xx_""'19 I 12-47E610-73-2,E71 '-© ~I ~I 7~ 8 1----1~....--.., 12-730E928-8LJ 0-47E800-1 ............. SYSTEM
~.
r.::,._ I ,C.-,. ~ 73-36 FROt.t RCIC SYSTEM ~ s E TABL 1 * ! FLOW DIAGRAM-GENERAL PLANT SYSTEMS ( HS \...~--1rsv, A5 2 * ~ ;;;.,.--- HPCI OIL SYS. Zl-73-19 X 2-47E2847-SERIES ....... CONTROL AIR SUPPLY 73-62 73-62 FSV 73-63 HS 73-63
/2 47E610 71 1,F6 73-29-1 73-298 I I 25-49 L ______________ ,l... ___ j_ ___ l 2-45E614-10 ............ WIRING DIAGRAM 120V AC/250V DC VALVES &
..._, ~ ..._, y L __ I - - - - - - - - - - - - - - ______ 1 I MISCELLANEOUS SCHEMATIC DIAGRAM
~ 1 :;!.; ;;i..,"'
- 9-4?...L..
0-45E714-3 ............. WIRING DIAGRAM 250V DC REACTOR MOTOR
*
- L-j2-45E614-10 I I ~ JQ I ~----L--------f""----, OPERATED VALVE BOARD 1A SCHEMATIC DIAGRAM
. .. - r1 o-45E714-3 I l STEAM TRAP 2-45E714-2 ............. WIRING DIAGRAM 250V DC REACTOR MOTOR
* * * * - - - - .. .. - FG' ' I zi~s-3,,,5~3 i....!!:....l ._ -j !
I 25-63 7~4 I FROM
~
I I I TRP-73-5 OPERATED VALVE BOARO 2A SCHEMATIC DIAGRAM
- ct 73-513 I FROM FIC-73-33 I ZS-73-1 BB I 9 3 PISTON 2-45E779-12,-13. ,, ,, , .. WIRING DIAGRAM 480V SHUTDOWN AUXILIARY
' 2 730 928 7
- f_ ~HIS owG.c1)-L.q'H1s owG.F+I 6~:~A !~:~ A 3 ~ I ~ POWER SCHEMATIC DIAGRAM E
* *
- 5 REMOVED
- 1 73-JSA - - E - 7 29 ;;;i"'
- I I ....._, CPLG-73-548 (TE\ /TE\ 73-19A 1 ~ 0-45E779-43 ............ WIRING DIAGRAM 480V SHUTDOWN AUXILIARY POWER SCHEMATIC DIAGRAM
* *
- I I 9-19 CPLG-73-54C - CPLG-73-54A 73-548 73-54C --9-3 I ~ 2-45E714-4 ............. WIRING DIAGRAM 250V DC REACTOR MOTOR OPERATED VALVE BOARD 28 SCHEMATIC DIAGRAM
' .ltFCV\5 _JI @ I T : ~ I ', 2-47E611-73-SERIES ..... MECHANICAL LOGIC DIAGRAM HIGH PRESSURE
~J's 73-31 J_ 7cJ,9B , ~ CCX>LANT INJECTION SYSTEM 2-47E611-64-SERIES ..... MECHANICAL LOGIC DIAGRAM PRIMARY CONTAINMENT
', ,ex,25 - 5o y I I "" > - - - - - ~ I d- 3 Pl 73-37 ISOLATION SYSTEM f
LOG HPCI - - SPEED f-e-...L HPCI 2-45E670-SERIES ........ WIRING DIAGRAMS-ECCS
~ IL 35
.l...~ r----,-1 f SJ l I I A5162 BOOSTER IN- MAIN HPCI 5
- - ( .. E.. ) ...
I _,___~~I--- -L-..--t_W_8R--~_!T~5-E4--~ 7 C I 1 7 1
~ ~ZI-73-45A
~
I LOG I IA730321 713-318 25-50 I ) d-3
~
- PUMP PUP-73-29 _f- _CREASE_r---1_ PUMP PMP-73-54 ~
~ f>A A51659-3 L01G
- Zl-73-6A © © GE DRAWINGS:
2-730E928-SERIES .......... HPCI ELEMENTARY DIAGRAMS L DISC POSN
, 9-3 J.. J:]:Q[J I~
73-45
~-:~
73-54-14
..._, 73-32 PT 73-31
-~ 73-31A GR-73 I I
I L-.-----,.-~ t!?i,t (PI\ 73~8
~
f PI 7b9 L ~ ~/\~ L!....,~ 7-tJ.6 I I 1 9-3~5ZI-73-68 3-GRAHAM MFG CO. : C16720-1 .................. HPCI GLAND SEAL CONDENSER TURBINE DRAWING: 2-E-4707-CONTRACT #66C60-90744,91750 .... HPCI TURBINE I .. PUMP DISCHARGE ~ - TW FE 1 I I
-- 73-32_.__,........,*73_33......- - - - - - . . - - - - - - - - - - - - - - - - - - - - - - - - - - " " " I I ,e:-,. I SYMBOLS:
}-----J 1~----~I
- - - ABANDONED IN PLACE (TYP) 5 25 - 63
,....iz:> I /FSVL.....L ~ QUICK DISCONNECT WITH SINGLE END SHUTOFF r-----;c;-~IL-73-44 r I ----- -~IL-73-34 t---l~--,S25-63 *" ( PS' C
L-, 2-13oE928-8 I I Gsi-3 ::r 73-44 ~ 7 4
~
- 73-~6
~
1 ~A - : ~ QUICK DISCONNECT FEEDWATER LINE A I I 73-44A MCC I I I
~ MCC 9-3 L-.../Hs'\
7~4A I ~5~63 73 y 1 l ,--. .,
~ .:.
z 0,: D... PRESS TEST VENT X v-.=FCV
-V7~A H I I
~ CONTROL OIL D
t-- I
!2-47E610-J-1,H5 ) -
I Zl-73-44A I Zl-73-34A U __FT_).J y 1
~
D... en i_. . - I __ J_~-~ I [R)' (c'5 7 3 1@:tj ~ I g 0,: 0,: 9-3 a; - ... I {--, 730E928-6 I t---------T-X r--10-4SE714-3 I r------i I 7~3 1--1-....J
-z 0 ."'
I I I I 8 L IL-73-3D >- 1 L--j 2-45E71 4-2 -I 2- 73 oE 928 - 7 I 1 I 1 ucc '.ii')' tJ
- , "z "z _J ....J J!5-63 25-63 r--------1I ___ ....J I I~ J, 9- 3 en -' -' r---- ~
~
l2-730E928-51 FCV I- LP~) A~~3 ', *
- I 7 ~ I-- -' I I 7~3-20A 7~3-2D8 I ~ - J . - -,
8 tJ tJ 9-19 I 73-68
'-,Y'jl : ~ D... K
- c :c K
_[i5-63 +. l 2 5 - ~ - - - 25-63 25-63 @ n FCV ~ °l'
-1 z
-"' "' "' (rs) diS ( rs) 73-21 y H
©-@1fW © I 0 en 1i-9e 1z32 7~-yA PS 7?D 7?C PS 9_3 I 25-50 FRCM FCV-71-68 X 7:t=§JI
-t--
m
~
- 'l ZI-73-23
)'i,\'
>i" 9-3 (ISv\
73 _23
< TO XM-73-19 THIS DWG,E4 f--- 1 LOG 1:
1
~-3 73 _33 Zl-73-308 1:Zl-73-30A (HS\...-lj @9-3 en D...
D... en en C z z C
@~--<fJ 2-47E610-71-1 ,DJ)>--
,---j 2-45E620-1 1 -
n N
*y A5167 73-308 ..._,
73-30A en
-' -' - LOG I I
..._, I
" " A5160 1---~ 9-3
~
L--1 _ , TURBINE EXHAUST
- - . 3 TO MAIN STEAM CONDENSER "A" C
1~- I 3-8A A
-' . -* r-.--.JL --j 2-13oE92B-4 11 L......12-47E610-1-1 ,A4 f-- ~ - - -
+
~
lR)Il-73-64
- <:2-47E2610-74-1,C9 TURBINE EXHAUST t - ~ ~ -.....
_J 1~ LS -,---....l
- @--T-@ili IL-73-10A I IL-73-108 TO STANDBY GAS 2-47E610-64-1 ,F3 TREATMENT SYS DRAIN 3-88
~
~ '~c MCC
{xi:) - 93 POT ~ ,-730E928-8 1 (iis'\9-3 _-+-__/HS\
©T-
~
ew9-3 _, 7~A I 7c!9B X "~64 7~4 ©-~. . . --p-45E714-3 I TO FCV-73-26 AND FCV-73-27 (THIS DWG.8-6~--'1 73-57 I "" I G I R
~
ZI-73-64 (HS)480V REACTOR 12-730E928-7 FROM LS-73-57A,B f-----, I FROM LS-73-57A,B
!2-730E928-7 f---,
I PRESS. PRESS. E-IL-73-10 C E) ) " I ' I 73 _64 Mov BOARD 20
..._, X-214--.._ I~ X-222TH1S I DWG.8-8
>- 1 I ITHIS DWG,8-8
>- -, I TEST TEST
,--l---,
/c,/"'
- GLAND SEAL I '"-~-..JV FROM LS-73-56A.-56Bi I FROM LS-73-56A.-5681 : , CONDENSER BLOWER I I I ">--+ ~ GLAND SEAL BLW-73-10 LA_
L-,2-45E779-12 ITHIS DWG. ITHIS DWG, F6>---+ 1 t-1 LS TO LT-64-54 X-220 - - - ' 11*-aTQRUS~,'~~ ~: ci-;s : ~ r [©r -©-©Tj rg;j , - 1©-©Tl
~ rS@Tii@ @-3 @-3 CONDENSER
'--'---;CND-73-703~--'--'
I 7YA 7~~SJB
- 2-47E610-64-1,A6 )
I I LOG 11 CISS lzl-7J-26A I R.11 G
~
I R Zl-73-2681 G I I R G I Zl-73-27A I I I I R Zl-73-2781 G I 7~3 -l r (es\ h© IL-73-158 I HS 9-3 y 5A 73-158 73-15A y ,- LS B I .1,-2, c,.. A . 9-2. 1 - ~ 1 (i?\ 9: : 1 1 - ~ r CISS I l~L~OG~'I '-r-L-U-BE--,-, I -r 1 rr 1 J 8 I 1/3-27 CLI OIL 73-158 I 73-15A I I I 73 -l5A 1 L,2-730E928-2 x-2059--/
~73-~-3 1,-26A IL-m26 I
~ r,;'I° 73-268 1
..._, I I
1 73-27A L - IL-~27-~--f-Ll,---~:::i-73-21* 1 I- - CISS I I LOG I B I 173-27 a TW COOLER CLR 702 Y1Y t-I
- l ~ - - - ... ---1-....fLS\
I I 73-158 HOT - AMENDMENT 28 l--j 2-73DE928-1 I IL-73-58A ~ Gs'\ 9-3
~ HPCI AUTO ISLN LOGIC A """f2-730 E928 _2 I .,,_ 13 25 ucb[_,3::l6_y_ _ _J t.eeT __ ~-[__ _J A 73-j.2 I IL-73-15C
~-©: I Y.__"'i__:W~E:L:L_J IL-73-59
¥
~
¥ L __
73-58A I RESET
'::"°_J r.:::,.,9-3 LXS'\
73-59 r HPCI AUTO INIT ___ --j~-JJOE 928 _3 RESET
- I L--I1------~*T+*---~~~---~~----~
TO FCV-73-40 (THIS DWG, F5 f----_j ____________ j 1 I I GLAND SEAL CONDENSER CONDENSATE PUMP I I MCC G-,----------------~ L--j 0-45E714-J I POWERHOUSE UNIT 2 PMP-73-15 L--~-J
~ G')9 - 3 HPCI AUTO ISLN LOGIC B -t2-JJOE 928 _3 X X L, 2-45E714-2 I - ' BROWNS FERRY NUCLEAR PLANT IL-73-588 Xf"- 73-588 r I L _ _':::'_J RESET *
.----------~=======::-::-===========;----------------------------------<73-11 /y)_ FINAL SAFETY ANALYSIS REPORT
~-3 LHS'\
73-61 HPCI (AUTO-INIT)
..._, -MANUAL isDLAlION- - -
-j 2- 7JOE 92 B-J I
CRW DWG ZONE (TYP) ECCS-ANALOG TRIP POWER TABLE 1 ..._, MATCH POINT (TYP) -(!)@J-- ~IV ------<2-47E610-71-2.E5 I TEMP. ELEMENT LOG TE-73-054B A73054B HPCI SYSTEM A
~i riv ------<2-47E610-71-2,E4 I TE-73-054C A73054C COMPANION DRAWINGS:
I 0-45E610-73-1 MECHANICAL CONTROL DIAGRAM I I
~J 2-45E610-73-2 DIVISIONAL POWER SOURCE (HP)
D-- REF CONTROL DIAG (TYP) FIGURE 6.4-1 8 7 6 DIVISIONAL POIER SCHEME 5
-- 3
9,o~ l-tl83Lt-Z L9 NOTES:
- 1. CRW DRAINS ARE CONNECTED TO REACTOR BUILDING EQUIPMENT PRIMARY CONTAINMENT SHV-647 VTV-646 DRAIN SUMP.
1/2* TELL-TALE"""\ 2' rl"'4--~--, - 2' OVERFLOW 2. ORW DRAINS ARE CONNECTED TO REACTOR BUILDING FLCXJR DRAIN SUMP. 1/2 *--rt*-~ SHV-72 YI SHV-649 CKV-609 cKV-s10 SHV-611
'N 3. DELETED NOTE 15 "- 1-1/2" VENT J/4" 3/4' SHV-5848 VTV-5858 'y' DRW TV-642 SHV-643 SHV-645 TV-644 4.
5. ALL VALVES ARE THE SAME SIZE AS PIPING UNLESS OTHERWISE NOTED.
~lAE~~~~~u~5T~~~ TEST CONNECTIONS ARE 112* UNLESS H
DRW 'N FRO~ CNDS STORAGE VTV-78 2 ,, S~8
. m ANO SUPPLY SYSTEM D>--<.~-ll.Ji/1----TO ATM
- 6. ALL VALVE BONNET VENTS, BODY DRAINS AND PACKING LEAKOFFS ARE 1/2" UNLESS OTHERWISE NOTED .
;; m 4"
~
w '
, 2-47[818-1,84
~
I I I ~
~
'?(NOTE 14) cKv-s8oe cKV-s81e SHV-5828
. 7. VALVE POSITIONS SHOWN ARE FOR NORMAL SYSTEM OPERATION (FULL POWER).
l CKV -54 TESTABLE w, ECKV~ i lrcv\1 is:5J
-'i"-""f
,..,- SYSTEM FILL
- a. BACK FLUSH (TYP 2) 1:J, ;_.1-.....
2 --,1" (TYP) J \~{~;) CKV-628 8. ((]ETC. DENOTES DESIGN PRESSURE ANO TEMPERATURE AS GIVEN IN TABLE THIS DRAWING.
- ../ ~ -
1 O' 12' CHECK VALVE - 75-561'""a.,f.
- jl ~IXA~¥~x-
- 9. UNIDS ON DRAWINGS ARE FOR REFERENCE ONLY AND ARE ABBREVIATED AS SHOWN IN THE EXAMPLE TO MEET SPACE CONSTRAINTS. REFER TO MEL FOR
~
~
MPEN-100-168 X-27C- {i~
~
<~
"' PS 1" r-1/2" TELL-I RFV-583 RELIEF VAL VE TV-788 2'
F
\ 75-78A ALL PU¥PS.TRIP EL 647 -0 "SEE
~PLETE UNIOS. ALL UNIOS ARE IN UNIT 2, SYS 75 UNLESS OTHERWISE NOTED. LEADING ZEROES SHOWN IN MEL AS PART OF THE UNIO ARE NOT DEPICTED. FOR ADDITIONAL GUIDANCE, REFER TO NEOP-4.
REACTOR ~ ffi ~7 2 . RTV-52 TALE SET o 150 PSIG
++~~- ~ TANK EL 644'-3"
-~~+-*
RTV-78AB DRV-78A .. NOTE 12 EXAMPLE: MEL UNID DRAWING UNID
~ ~ ~ 1:~ - -,_ " ~ DRV-53 3 4*
t;; ,~ ~ I (SEE NOTE 13) RTV-78BA VTV-788 BFN-O-SHV-018-0502 8FN-1-HS-031-0160A SHV-18-502 HS-31-160A y
'------- NOTE 17
~ ALL
.. 1 " 2\ 8FN-2-PT-003-0204C PT-3-204C
.----------'1-"_ _ _ ~ .., ~ ~
SHV-71 N NOTE 17 1:1'""'!, -1.R~~~28: IO ffi A *ffi ~/4" TV-78A i.. 75-788 PUMPS STOP EL 646'-6"
- 10. THE DESIGN PRESSURE AND TEMPERATURE OF ALL DRAIN AND VENT LINES THROUGH THE LAST ISOLATION VALVE SHALL BE THE SAME I .J ~* ',t ' .,"' .,7 i~ ~ $ -~>I [j--,....-gj DRW
' ~SEE AS THE PROCESS LINE. G
*, X-27D l.--- 1.1.1 c ,,, 1-1/r VENT 1" TO RAOWASTE SYSTEM PSC HEAD TANK DRV-788 NOTE 12 11. TOLERANCES OF +2-1/4"/-1* TO BE APPLIED TO INDICATED
.- < ;:: * ~ ~ DIS :!it;; ~
- 1- ~ 2" TNK-78 RTV-7888 ELEVATIONS FOR LS-75-78A.8,C &. 0.
~ ~ ~ i::: ~ 75~8 blii r!_ ~ >>I- SHV 584A VTV 585A Y-I 1 1/2" 0 47E830 5,GS CAPACITY: 3090 GAL.
' RTV-78CA VTV-78C 12. FITTING INSTALLED IS PART OF A 1/4" UNION, TYPICAL .3 PLACES.
*~'
CD LJ'I, >u <.- .- - - .;
" i".~~-
;, ~(NOTE w 14) ~~ ~
~t, ,. 1" ::
- .; DRW f2l-+-<al i=..1 I '-=-' ~~ Is "'
1 "'I -~ 1ST PUMP STARTS EL 645'-6"
- 13. THE CONDENSATE HEAD TANK MAY BE USED TO SERVE RHR AND CORE SPRAY DISCHARGE PIPING INSTEAD OF UNIT 1, 2 OR 3 PRESSURE SUPPRESION CHAMBER (PSC) HEAD TANK. WHEN UTILIZED FOR THIS
~ ~ t e:~ ~
72-7!\C PURPOSE. THE CONDENSATE HEAD TANK SHALL NOT BE REMOVED FROM CD Z fFCv'\ ffi 1,:-3.s fFCv' \ i r-D~--"' cK~-58oA CK v-s81 A sHv-ss2A 1 m-+rn *N - ...._ nSEE
.uNOTE 12 SERVICE TO CALIBRATE LEVEL SWITCHES 0-LS-2-172A. 8, C AND 0.
SHV- 6'\ll,~~---------------------...1-_;;;;_____ (FCV \ 7*y5-27 ;;CKV-26 75-23 2' (REFER TO 1-47E818-1, D4) LOV-23 tr "'" V>. * " 1~ RTV-78C8 DRV-78C
'" TESTABLE ' N 1 4. VALVES 2-FCV-75-25 &. 53 WILL SEAL LEAK TIGHT ONLY IN THE
~t:=:=::=:=:3' tt-:~-'.!!:~::::io-l.,k:J---_;:_d.~1!2:*
I ~1 . N5AJ 1--{IJ
-;-y:,:~~CH~E~C~K=V~A~L~V:!;E-;:'.-j-1""-~
m c.. < MPEN-100-16A .J
"~f.-------~~~~~,+::i""'.'".".".~-~~--!-l-------~~7:"'~""'.-----,
~
- :_ ~
~
1
- 3/4" TV-648 SHV-637; ~
ID 10"
. . .- - - - - - - ~ *
[I}-+IB f rev,-------..
~
'----------r--~----"T---1"-
LOV-50 7f:.50 l
, 7 .25" PRo-'Crot--"3'-"4"-1"CX,:1-----i>I----,,--,
N -~ 1" RTV-78DA VTV-780 LS 2ND PUMP STARTS DIRECTION OF REACTOR TO CORE SPRAY SYSTEM DUE TO PRESSURE LOCKING MODIFICATION WHICH DRILLED A 1/4" HOLE IN THE REACTOR SIDE FACE OF THE VALVE.
- 15. CAP IS OPTIONAL.
~ ffi ~ 1::24 - 1 * ~ ~ ~ ~ ~
75-78D EL 645'-0"
*N -.......- SEENOTE11(TYP) 1 6. VENT, DRAIN AND TEST CONNECTIONS 1-1/2" AND BELOW CAN BE
~>' 314* u ~,5.;-2 ' i- FE RTV-49L ffi ;. ~ ~ ,,, __... RTv- 2~ ~ ~ ~ t 1" PROVIDED WITH PIPE CAPS OR HOSE CONNECTION FITTINGS WHERE TV-638 SHV-639 ,-... 10" E-t PUMP TEST 7 *~t_ 75 -j.9 3/4" _ '1~d '1~J RTV-78DB REQUIRED BY PLANT PERSONNEL. THIS CONFIGURATION IS SUPPORTED BY ENGINEERING CALCULATION C0-00999-923399.
m ,,,
~
~
~
> ~ ~"
tn ~ ,
~
ls ~ =-~ g J J t;; i
,,,~ PR0-669 r"-C<~-Df-r-----,
,1r- 3/4" FE RTV-21 L..,.~ 1s'
~
, , v, 0
~
RTV-49H ORV-78D 17 . FOR ORIFICE COUPLING, SEE 0-47W600-20, DETAIL P20.
> > ~ v-,,, ...,..., L<- ~ -,_~T_) (_~S_) > REFERENCE DRAWINGS:
112 * ~ 1-, ,, 7 1- .., '7Lllr1n--!-~-!2.2L;2u i::: - 14
- r1<,-.1,g
,I ...
~
~
~
1s-3~14
- l-~!£)~-Df-1--_J !'l
.~ 1 ~1 1- N m MEL ................... VALVE MARKER TAG TABULATION F LJr, l/2.!.....
- z .,, > 2 ,,RTV-21H t 0-47E800-1 ............ FLOW DIAGRAM-GENERAL PLANT SYSTEM 0-47E800-2 ............ MECHANICAL-SYMBOLS&. FLOW DIAGRAM DRAWING i .!.fl
...1
~
- i *1 .. 1-y~ INDEX 1 .~ VTV-640 SHV-641 2
- 0 3/4" 2-47ES17-1 ............ FLOW DIAGRAM-REACTOR WTR RECIRCN SYS
~ 3 TO RHR SYSTEM 47W2458-SERIES ........ PIPING
'N 1O' PUMP TEST
~
o.... 622 3/4" TEST 623 1/4' - TW 7..5-J9 .- DRV-93
--t__:2~4:!_7~E8~1~1~1~,H~3~_,) MEL ................... INSTRUMENT TABULATION 2-47E610-75-1 ......... CONTROL DIAGRAM-CORE SPRAY SYSTEM 1O" PUMP TEST
~ FCV .
N RELIEF VALVE 2-47ES11-1 ............ FLOW DIAGRAM-RHR SYSTEM 2-47E813-1 ............ FLOW DIAGRAM-RCIC SYSTEM
~ ~ SET o 50 PSIG/ 2-47E815-4 ............ FLOW DIAGRAM-AUXILARY BOILER PUMP MINIMUM FLOW 3" 3/4" TEST ,1 /2" I2~ ~,,... * ~.7~7 FCV r 11 SHV-88 TO RHR SYSTEM 2 47E811 1.H7 R'F~V-629 / r 2-47E817-1 ............ RX WTR RECIRCN, DRAINS, VENTS&.
SLOWDOWN SYSTEM SHV-595 TV-594 1" RELIEF VAL VE SHV-87 ~- 9 l TO CRW o . N 2-47E818-1 ............ FLOW DIAGRAM-CNDS STOR AND SUPPLY SYSTEM 0-47E830-5 ............ FLOW OIAGRAM-RAOWASTE SYSTEM SET o 500 PSIG 3/4* TEST 1 /2' EL 621.25 2-47E837-1 ............ FLOW DIAGRAM-REACTOR WATER CLEAN UP DMNRLZR r ------------------------1.2..*_ _,"'i----i:11 PR0-543A 1 "-7'.l 543A
,-L RFV-SHV-597 TV-596
- BYV-89 2-47E837-1, 83 L..:.._=-:c.;__:..:....:.::...._..,)>--------' 0-47ES20-1 ............ FLOW DIAGRAM-CONTROL ROD DRIVE HYDRAULIC SYS 2-47E854-1 ............ FLOW DIAGRAM-STANDBY LIQUID CONTROL SYSTEM 2-47E852-1,2 .. , ....... FLOW DIAGRAM-DRAINAGE PT '----!C:XX:J---1 3" 3" PUMP MINIMUM FLOW 2-47E610-85-SERIES .... CONTROL DIAGRAM-CONTROL ROD DRIVE SYSTEM 75-20 0
"'m. 2-47E610-68-1 ......... CONTROL DIAGRAM-REACTOR WATER RECIRC SYSTEM
~ RTV-20 N RELIEF VALVE 14" 12" PRIMARY CONTAINMENT N ' SET o 500 PSIGt - 1" 11----it-------T---C+-----------------, SYSTEM PRESS.-TEMP DATA
"'mw RFV-5438 ~ I 1* BYPASS 1" BYPASS ~ PR0-5438 I MPEN-100-2238
'fr.,-~~-'f'\. -~ 75-18 *-
BYV-10 Oilt--~'i'\. 7~0 _(SHV' !;;: N ' <2-47ES52-2.B4 w v L---!:XX:J->. PT 75-48 LINE DESIGN DESIGN MK NO. PRESSURE (PSIG) TEMP(°F) E
- RTV-48 y ~ 1250
~
~
BYV-18 "'-~~:,f-....J.1_"--IIC>il 1" 1 575
- N SHV-18A VTV-18 ~ 2 500 350
~
w SHV-10A VTV-10 c:-,. *' ~ 1" BYPASS 3 150 210 LOV-574C LOV-574A 4 123 130
~
' {:_H4~*-. -*
{;;/~/Y----1~ 5 90 130 0 1/4' 1/+" , .. __... ,Al'---~-a 110 130 SUPPRESSION POOL 4" I 3' 3' I 4" ~...-'--~:if---' BYV-46 ~.......J.1_"-tlCIIII-" eYv-38 1 /4' 6 7 50 150 1" VTV-46 SHV-46A VTV-38 SHV-JSA
~EN-10D-223A = sHv-s14c . w 8 120 103 HIGH WATER LEVEL 536' -7" .
'N
~
N LOV-5740 LOV-5748 ~ 9 10 75 100 130 103 II~ ~
.~
~
-=~
~ 1 /4" /4' u '
~
~
4" - - 3' 3'
- ** ~-
' ::; . SHV-5748 I 7~ ,
LOW WTR LVL fs36'-1-3/4" WITH DW/WW*P 1536'-0-3/4* W/0 OW/WW.AP
.. 75-17
~
-~"' a
.:~ 75-8 ->'
~
N
@1----1---ill 3* CKV-570C ~ u 4" , PR0-6.710 3 ..
---1'1-"-l/'1--l<:l-:-i .500" PR0-668 T\ PRO-
{f667 1" PR0-671C
,1,
- 4'
/PS\ 3 * .9.375 CKV-570A 4, C CKV-5708 . 9375
- I 2'
.9375" 75T-16 ...---1::1---1./'-'1--Kl...;.i ' ::;"' CORE SPRAY PUMPS RATED CAP . .3125 GPM o ~ RTV-35 uu 1/4" LEAK 5.41 :-) 1/4' LEAK
. 1/4" LEAK PR0-671A
" 582 FT TOH TEST m
n:sT m PI n:sT--, ~ 75-35 *
'~ ~
2" DRV-619 75-16 ~ * ~ t PS u ~ PS '" D G---
~
RTV-16 ~ 3" .9375" .... 75-35
~ ~
~
77:Y_ 7 r;,,~P""I"\I PUMP {PS\ STRAINERS
~
0 w 2' ., DET 2A l~'ir SHV-148 3/+" 5-7 RTV-J DISCH ORJ
"'"----1::1---*i..--,.,--1<1--1 75-44 1 I+" LEAK 3/4' i".' CKV-5.37 1 6'
>' PR0-671D CKV-570D J"'""t=flc>f---..~.:.;.., /\ oEST r,.;_;-i....1--..f--f"",-- SHAFT SEAL OR 1/2" (A,8,C,D) 12 ,,
STN-663 STN-664 STN-665 STN-666 cc
~
~
ORV-618 i SHAFT SEAL DR
}/4" TEST_ _ '-. ~- ORV-HB
~
m N 1/2" SHAFT SEAL DR - --1-+:+--'tc~:, 3/4" TESL - ' SHV-58 DRV-58 3 1c;4c,,""1 G?') DRV-338 SHV-338
- CORE SPRAY
-~ /7.:-':--'::-""-.&..:."""".,
./ 3/4" TEST H~-r---ac4--- u q~-,,f-itifTYP-4 PU-MPS- 1 ' CORE SPRAY '
~
0
- :2 ~
~ SHV-14C '-. _.... ~ N ' TV-SA PUMP RTV-44 7..::!-4 :S ..*_P_u_M.P.,.z_e_P-IMICP-..3..3-<)I.... .;/\ SHV-33C TV-3jA ~ ~
0
. , 'f
~
z u TV-HA ~ m SHV-5C '--.....1,..,,,,~P~U~lll"-;..-..:2:A~P-M;Pl_-:5_,..j DISCH DR--,,
- t
~
~
0 g ! jg i IT~ j §> PUMP ~ly2C SHV-14A DRV-14A
- w
~
f5
' , SHV-5A DRV-SA 3/4" ,
~
SHAFT SEAL OR 1 /2" .
~
DRV-33A SHV-.33A ~ i:li. VENT, >2" 0 CYCLONE CYCLONE SEPARATOR w o
>I-
- PMP-H
- CASING DRAIN &: ;::, 1 1" CASING DRAIN :_ . - DRV-428 SHV-428 3/4' TEST 1" CASING DRAIN ~
SEPERATOR SEP-5 (2A PUMP)
~
1 in
- I CORE SPRAY N
.500" SEP-14 12c PUMP) u I< 2-47E852-2, es r ;: ;,
r SUPPRESSION RING HEADER POOL SHV-76 llf:: 1 /2"
.... ID m c-i N
L_ C "' m N PI 75-4 RTV-4 m
- < 2-47E852-2, 85 I--
. i--:-:~~~1-PUMP 2D PMP-42 11. ~ SHV-42C N
i". TV-42A
,T" N PI 75-32 RTV-32 "'mw
~
v SEP-33 28 PUMP) SEP-42 (20 PUMP) SHV-75 1--[:xx:]f-->.75-13 PI ~ ~ 0RV-42A SHV-42A :0
~
~
N v DET 2A 1
> m.
<2 47E852 2.84
'----~----/ RTV-13 --
LI.I IIIO 1 1" CASING DRAIN *o
. N TYPICAL FOR CORE SPRAY PUMPS A,8,C,D
~ s
' 1.1.1 1 /4'
~
1" RELIEF VALVE 114
- LOV-512C N
1 !;: t.!. LOV-512A 1* RELIEF VALVE v m
- w
-l------lll( '
1 RFV-5078 N
"'mw
' DS iii
- SET ,*. -.1-50_P_s_I_G-t. ... t,t-*-*-1~*~--1,J--v_3::_'~,_._ __3_'-:::::\--:~'::--/-;:!~*~:::--+1 (J/4f';'-'1'~S.:E_T_o~1~5-..0,PS IG VENT 1 " RELIEF VALVE SET PI m N N' 1" RELIEF
~
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- TO CORE SPRAY 2' 2" SYSTEM II 16" FLOW DIAGRAM COMPANION DRAWINGS:
NONE FIGURE 6.4-2 8 7 6 5 4 3
8£0~ l-£L-Ol93Lt-£ V'J L9 NOTES;
- 1. THE HIGH PRESSURE COOLANT INJECTION SYSTEM IS DESIGNED TO COOL THE CORE WHEN THE REACTOR VESSEL IS ISOLATED AND REACTOR FEEDWATER IS UNAVAILABLE. WHEN THE WATER LEVEL IN r--,3-13oE928-2. -31 THE REACTOR VESSEL REACHES SETPOINT OR THE DRYWELL PRESSURE REACHES SETPOINT. THE HPCI SYSTEM IS AUTOMATICALLY STARTED STEAM LINE SPACE TEMPERATURE I BY ACTIVATION OF REACTOR LEVEL SWITCHES LIS-3-58A, LIS-3-58C, 4 PLACES I LIS-3-588 AND LIS-3-580, OR DRYWELL PRESSURE SWITCHES PIS-64-58A-D (SEE 3-47E610-3-1 AND 3-47E610-64-1) ARRANGED IN I ONE-OUT-OF-TWO-TWICE LOGIC TO OPEN VALVE FCV-7J-16. THE SAME r---,---r--,-- - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - r - - , - - - r - - ~ - - - - - - - - - - , - - - r - - ~ - - - r - - + - - - - - - - r - - , - - - r - - , SIGNAL OPENS PUMP DISCHARGE VALVE FCV-73-44 ANO CLOSES DRAIN LINE VALVES FCV-73-6A, FCV-73-68, FCV-73-17A AND FCV-73-178. H 9-3 Lili 73-55
~73-2A
~73-28
~73-2C
~73-2D I
1 ri3-45N365+-1 I
~~~~
73-2[ 73-2F 73-2G 1J-2H
~~~~
73-2J 73-2K 73-2L 1J-2M
~~~~
73-2N 73-2P 73-2R 1J-2s
- 2. A MECHANICAL SPEED GOVERNOR PREVENTS TURBINE OVERSPEED BY TRIPPING THROTTLE TRIP VALVE FCV-73-18 WHEN A TURBINE OVERSPEED OCCURS. A CONTROL GOVERNOR, FIC-73-33, EQUIPPED y WITH AUTOMATIC AND MANUAL MEANS FOR SPEED SET POINT CONTROL I -------- -------- --------
"I3---4-5-E6_2_0--1~t--. - - - ------"1---------- -t-l------------------------------------------ -------T---------------------,---------*----------1 REGULATES TURBINE SPEED VIA TUR8INE SPEED CHANGER" TO 0
CONTROL HPCI MAIN PUMP DISCHARGE FLOW. r ------T--------- -t---------------------------------------------*-------r--------------------~----------~---*---, II 3. DELETED TO XM-69-29 3-47E61D-69-1,G2 r ------;-----------+------------------------------------------- -------+-------- --------, I I 4. THE TURBINE CASING IS PROTECTED FROM OVERPRESSURE BY TWO RUPTURE DISCS. PS-73-20A THRU D DETECTS DISC RUPTURES AND REACTOR WATER l ------t---------- -~-----------------------------------------------, I I I I I ANNUNCIATES ON PA-73-20 IN THE MAIN CONTROL ROOM.
- 5. DELETED CLEANUP 1 L- _---0E')
7t;JA I I A 73-558 I A 7t;JC
- I A
7t;JD
- I
- 6. NORMALLY MAKEUP WATER FOR THE HPCI PUMP SUCTION IS SUPPLIED BY THE CONDENSATE STORAGE TANK THROUGH FLOW CONTROL VALVE FCV-73-40. AN ALTERNATE SOURCE OF WATER IS
' HPCI ROOM I II I TORUS R18-U I TORUS R17-U I TORUS R16-T AUTOMATICALLY OBTAINED FROM THE SUPPRESSION POOL THROUGH ISOLATION VALVES FCV-73-26 AND FCV-73-27 WHEN THE CONDENSATE I I ELEV 519' I
LOG 73-3 o LOG 73-3 cL (G:\ 9-3 I ELEV 550' I
, 1 I ELEV 550' I
~--+-'--~--------~ 1 1 ELEV 550
- HEADER LEVEL DECREASES TO SETPOINT. PUMP DISCHARGE FLOW ROUTES ARE AS FOLLOWS:
f- ------,.----1 1 CISS CISS I 3-45E620-1 1-,-- ~ L ~-21 FROM CNDS. I A9-21 i --?9-21 A. NORMALLY VIA THE FEEDWATER LINE TO THE REACTOR THROUGH ISOLATION VALVES FCV-73-34, FCV-73-44, AND B , 1 B I I 7~-5§8 DMNRLZ WATER L_...£.m L-...l.m CHECK VALVE FCV-73-45. 73-55A 1 '-- Gs'\ 9-3 i i : - o-47E610-2-2,G3 1e;;c 1e;;D
- 8. DURING TESTING VIA TEST LINE TO CONDENSATE STORAGE THROUGH FLOW CONTROL VALVES FCV-73-35 ANO FCV-73-36.
r ---r-,
,~ G
, 1~, 9-3 STORAGE SYSTEM C. MINIMUM FLOW BYPASS TO THE SUPPRESSION POOL THROUGH 1 1 1 1 FLOW CONTROL VALVE FCV-73-30.
I MIMIC r-------,--, 7. IF HPCI SYSTEM OPERATION IS REQUIRED WHEN IT IS IN THE I 3-45E779-1l f-~ f-,-- I I
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J __ ULl 9-3 TEST t.40DE, THE CONTROL SYSTEM IS DESIGNED TO AUTOMATICALLY RETURN THE SYSTEM TO THE OPERATING t.ODE.
- 8. COOLING WATER FOR THE PUMP. TURBINE, AND LUBE OIL COOLER IS SUPPLIED FROM THE PUMP DISCHARGE LINE THROUGH PRESSURE CONTROL VALVE PCV-73-43.
- 73L~: 0 I : 73~r CL: 1, ~~ _ _ J, l ../Hs'\ I MZI-73-3 , +--i J-JJOE 975 -s 73-+0A 1 73-18A 73-188 9-3 7~;!8 73 4
- 73-4A
- 9. ALL INSTRUMENTS ARE PREFIXED UNIT "3-" UNLESS OTHERWISE NOTED.
1 o . CCMPLETE LOGIC FOR HIGH PRESSURE COOLANT INJECTION IS SHOWN
~ I- 73-38 I ! 3-45E714-2 f-1-- LS !---------J:'. 1- - , - - , 'y' 'y' (AJIL-73-188 ON MECHANICAL LOGIC DIAGRAMS LISTED IN REFERENCE DRAWINGS.
- 11. UNID'S ON DRAWINGS ARE FOR REFERENCE ONLY AND ARE ABBREVIATED
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** -'-, ,c T 73-40 1 FROM 73-408 I r-r---,
FROM MAIN L ___ _tMIM~--. _J X-11 ~cc __ T ___ J ' CISS ' ' CISS ' J_,, - ~73-26.-27 ..___, I r"'13-45E714-2 I : I rs:TM.:t,L~I~NE~8::;,~7',,~~-/. FE ),~~~~~-"'_1___ IJ-47E610-1-1A,F7/ \.7J-1 13___2~~-,t>l<J-~-8::::::~~IL.;;,7~3~3~8~:"<1--~~'~~~~'T"~~M*A-I_N~s-T_E_A~M~S~U~P~P~L~Y~~"""."-~*...;~~-*-~*-*~-~*.;..;1:~~,c~c~:""'"-~~~*~
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47E605-SERIES .......... MECHANICAL LAYOUT OF CONTROL BOAROS 47W600-SERIES .......... MECHANICAL INSTRUMENT AND CONTROLS MEL .................... MASTER EQUIPMENT LIST-HPCI 3-47E812-1 ....... , ..... FLOW DIAGRAM-HPCI 3-47E610-1-1A .......... CONTROL DIAGRAM-MAIN STEAM SYSTEM
* * * \J3-}5 73-31 ~ PI 0-47E610-2-2 ........... CONTROL DIAGRAM-CNDS & DMNRLZ WATER STORAGE
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73-51
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73-37
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~ ** *..#
rl 3-730E928-5 J_yI r,i>. ___, ~..___, 6 ', ', SYSTEM Zl-73-45A (ii', (CJ G [J]i[l _ - --, PT 25-50 11 9-3 - M PMP-73-29 r - PMP-73-54
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N FEEDWATER ' LOG (CKV' ~ fu471 73-31 -r- 73-31A F/ V\ A6165 ~ A LOm..,. I
- Zl-73-6A@ © 9-3~5ZI-73-68 0-47E800-2 ............. MECHANICAL-SYMBOLS A FLOW DIAGRAM DRAWING INDEX c6o53 L-t 9- 3 7~-;-5 1 ~ ~==-..___,~~ 1 'LI/'.'/ ~ ~ 'Ll0 Ji~ 0-47E610-12-1 .......... CONTROL DIAGRAM-AUXILIARY BOILER SYSTEM LINE B X-9A J *
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~ - z >- 1 - - I OPERATED VALVE BOARD JA SCHEMATIC DIAGRAM 3-45E714-2 ............. WIRING DIAGRAM 250V DC REACTOR MOTOR OPERATED VALVE BOARD 3A SCHEMATIC DIAGRAM D
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C16720-1 .................... HPCI GLAND SEAL CONDENSER Zl 23l.\ (lsv\ @- ~ 1 " ' - - - - 7 ~=; 3 .J 1
~
JV .18. 11
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C z UL\ r~1~1-t,_-_-_""__ I __[PT\ TO MAIN STEAM r w TURBINE DRAWING: C888X ....................... HPCI TURBINE
,Y. 73-23 ~.J ..___, /HS\__i HS 9-3 :::,
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- ~: ~"13-730E~928-:_1 3
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-* . ' ~-I 3-47E810-64-1.F3f-A
- f----.1 I3-730E928-5 f--...l L~ L--t-_J HS
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~-47E610-74-1C.E3 . ~ - I 3-8 TREATMENT SYS
- L ,--,---r--r-,
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I I I I ( e) I I 13-730'928-7 f - - - - - , 13-730'928-7 f----1 1---1 a PRESS. PRESS.
! 3-45E620-1 1----4 _j_
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..___, x-214 r~
L_, x-222 I I I FROM LS-73-57A,B V,------, I I FROM LS-73-57A.B
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- I @::)------f L-~-=--=--=--f 1-1--~@l I @)------4
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r111, Y--< Y--< I Y--< Y--< 9-3 9 -3 CONDENSER II I I rl3-47E610-64-1,A6 =& \. ,Qv_ ,Cg)_ JV .IQ ,Qv_ .IQ
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- AMENDMENT 28 7~A - - 7~A COOLER 9-3 r-x-3
~~ ~~
_ _ r.;;'{ LX$l HPCI AUTO ISLN LOGIC A _j 3_730E928 _2 I 73-158 HOT IL 73 SBA 73-58A L _ _'::::'_J r RESET !
- r;A-73-25 (!Sv'\-3
>-f' 73-25 MCC 7~-3,6 _J MCC 73~7 j
- I t--13-730E928-3 I y'----t__:W:_:;E~L::_LJ
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IL-73-26 L-1 L._ IL-73-27 TW-73-42 GLAND SEAL l!L-73-15C t-13-45E714-3 I POWERHOUSE I*- I TI UNIT 3
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L __ 73-59 I RESET
- -"_J TO FCV-73-40 r- r, CONDENSER CONDENSATE{ \ L-4 3-45E620-1 I
_ _ 9-3f4i IL 73 588 *L--="-J ~
~9-3 73-588 I HPCI AUTO ISLN LOOIC 8 _j 3_730E928 _3 I RESET ! -
NC
~----l------------J NC L, 3-45E714-2 I PUMP PMP 73 15 G )
BROWNS FERRY NUCLEAR PLANT
@-3 73-61 HPCI (AUTO-INIT)
-MANUAL ISOLATION----i 3- 73 oE 92 a- 3
~---~
I =r-r-, r-i--r: FINAL SAFETY ANALYSIS REPORT
..___, * © © ' '4------L-----------------------------------------------1_73-11 .(jw)_
CRW ..___, FCV ZI-73-178 ZI-73-17A FCV 73-178 73-17A TABLE 1 HPCI SYSTEM A TEMP ELEt.tENT LOG 3-TE-73-0548 A730548 MECHANICAL CONTROL DIAGRAM 3-TE-73-054C A73054C 3-47E3847-8 3-47E3847-8 COMPANION DRAWINGS, 1-.2-47E610-73-1 2-47E610-73-2 3-47E610-73-2 FIGURE 6.4-3 8 7 6 5 3
l-tl83Lt-£ L9 ~ VE "\. . " SYSTEM PRESS.-TEMP DATA PRIMARY CONTAINMENT 1iJ2 NOTE 13 VTV-646 ;1;~
~ S : 4' 1 /2' TELL-TALE""'\
.--I.,._-~--,
"j"
. - . - "' 2" OVERFLOW 2" LINE DESIGN DESIGN MK NO. PRESSURE (PSIG) TEMP(°F) s SHV-72/1-1/2" VENT+
SHV-649 CKV-609 m 5:'
- CKV-610 TV-644 SHV-611 *N 2 3
1250 500 150 575 350 210 J ( 130 w SHV-5848 VTV-5858 DRW (/) 3/4" - :Ire 4 5 6 123 90 110 130 130 H w<
~>
~ DRwY .N FROM CONDENSATE SYSTEM VTV-78 2
- ST~78 7 50 150
\ :;--... "'<lo<:: . - ..... 4" ~ 4" VALVE 3-2-724 [J,--=-rtu./f-.....~To AT" 8 120 103
..-...J.~~~~~~~~~~~.l..~-!,,"1-~~-i..--'1-....
~ ~ ~
CCI ,._;~1i<>f--,",~--<3-47E818-1.B5 9 75 130 I I .., .... t-CJ "':;i "' I I N m U,W wee
~~ SYSTEM FILL cKv-s8oe cKv-s81e SHV-5828
.... . . NOTES,
~
I .,,..-
- 1. CRW DRAINS ARE CONNECTED TO REACTOR BUILDING EQUIPMENT
\ (/) 75-53 i,-- & BACK FLUSH (TYP 2 PLACES) 1 - - ~ ~ ~ RTV-78AA 628 DRAIN SUMP *
~ 2. DRW DRAINS ARE CONNECTED TO REACTOR BUILDING FLOOR 1O" 12" ' \ ,--CKV-54 X-168"" '>--J' - - DRAIN SUMP.
- 2* ~ T V 78A TV-788 ( LS ALL PUMPS TRIP '- ALL VALVES ARE THE SAME SIZE AS PIPING UNLESS OTHERWISE TV-
,?48_
SHV-
'4A
;;; . ~ .... 2'
\
75-78A EL 647'-0" 4. NOTED. ALL PRESSURE AND TEST CONNECTIONS ARE 1/2" UNLESS REACTOR SHV- \ 548
- TV-54A PS 75-52 RTV-52 1 *
- "'s. N
~
1/2' TELL-TALE""'\ 7~1 _"f_
- I
-+--+--
It TANK EL 644'-3" RTV-78AB DRV~78A
- 5.
OTHERWISE NOTED. ALL VALVE BONNET VENTS. BODY DRAINS ANO PACKING LEAKOFFS ARE 1/2" UNLESS OTHERWISE NOTED.
"'-s.
RELIEF VALVE SET
----"' .,._NOTE 16 3/4" TEST- r
"'r "'
w cc ,-,SHV~-*71--l.,.l--,_ I o 150 PSIG 1
- 2' \
(SEE NOTE 12) RTV-788A L ~ 6. [I]ETC. DENOTES DESIGN PRESSURE AND TEMPERATURE AS GIVEN IN TABLE THIS DRAWING. NOTE 1sT X-27C 1 " RTV-5 DIS ~ RFV-583 ')tl,,.f---'-~ ~TV~788 7. HYDROSTATIC TESTING SHALL BE IN ACCORDANCE WITH APPLICABLE DRwY CODES. rl",,
- N Q LL~. ALL PUMPS STOP
- l ECKV- 1~ :; '?
.. ~ 3 /!I} g_ ::l -
1" TV-78A 75-788 EL 646' -6" y ** THE DESIGN PRESSURE AND TEMPERATURE OF ALL DRAIN AND VENT LINES THROUGH THE LAST ISOLATION VALVE SHALL BE THE SAME
,< X-27D 1~ 75-56 .'\.-1/2" VENT y PSC HEAD TANK AS THE PROCESS LINE. G r
"' 'f w
r
~ ~
~
ao
>'1
~
fcKV- 7..5-2J8 -fTO PDT 3-68-52 SHV-584A VTV-585A J V DRW TNK-78 CAPACITY: 3090 GAL. RTV-7888 RTV-78CA DRV-788 9. TOLERANCES OF +2-1/4"/-1" TO BE APPLIED TO INDICATED ELEVATION FOR LS-75-78A, 8, C, AD. TO CLEANUP 10. PUMP CAPACITIES:
> 1..1.1< 75-28 ...__ 471610-68-1. A6 DRWI BACKWASH ' ;;; - -~TV-78C ffi > ~ lzl-+-fal fsl--1--fBl PSC HEAD TANK PUMPS - RATED CAPACITY 40 GPM o 152 FT TOH.
~
n I-1
<lo<::
~~
CCI in
-{TO PI 3-85-19 47W610-S5-1A,B3 m
~ '
L.=J I ~ 4 r-- I - 2" TRANSFER PUMPS 1 0-47E830-5, (G2) & (F2) FL EL 639.0 j i' /_L~_\ 1ST PUMP STARTS CORE SPRAY PUMPS - RATED CAPACITY 3125 GPM o 582 FT TOH. DRAIN PUMPS - RATED CAPACITY 120 GPM o 150 FT TOH.
.~.,.,' .~\B I 75-78C EL 645'-6"
~ /SHV1 Z frCV\
ffi '1's'::'.i5
~
75-23
*S:., '
- cKv-ssoA cl<v s*s1A SHV-582A ~
~
*N -....;., 11 . UNIDS ON DRAWINGS ARE FOR REFERENCE ONLY AND ARE ABBREVIATED AS SHOWN IN THE EXAMPLE TO MEET SPACE CONSTRAINTS. REFER TO MEL FOR
'r-f . .,__.. . !t- ~ 2' COMPLETE UNIDS. ALL UNIDS ARE IN UNIT J UNLESS OTHERWISE NOTED.
' 75-27A
-ElE=~*=---,.:tJ,~)-~~-1:.::OO~'""'-~:....-i:,k:--r:~----J.....:.;12;.:*_4,-1';...C:..K:..v_-2:..6;..._ _ ...,~~=$r--~,~..
-------r--~ -~1Z,2j::"' ,--L-;--------H~::::::~;,;~~;;;:::::::::::~-------------
N
~ ID-f-ill SHV-700 7rrr5-50 _ __
tT_ RTV-78C8 ORV-78C RTV-78DA
- ~Tv* 78D LEADING ZEROES SHOWN IN MEL AS PART OF THE UNID ARE NOT DEPICTED. FOR ADDITIONAL GUIDANCE. REFER TO NEDP-4.
EXAMPLE: MEL UNID DRAWING UNID 1 N5A
/~-
1-{IJ CD ~ X-16A_/ ;1i r.1 PS RTV-24 1 .,
~
- !_ 4 u,
10"
. . .- - - - - ~------+
( FCV\ ... 1 7
- 25 "
PR0-670 IL... (LS\ 2ND PUMP STARTS BFN-0-SHV-018-0502 BFN-1-HS-031-01 60A BFN-2-PT-003-0204C SHV-502 HS-1 SOA PT-3-20+C
~' '1 ffi ;. 7z:3 4 ', /2" ;. ffi ~ SttV-~37 TV-648 7 5-{22 :, L1 ~ ~ RTV-49L 75-780 Y
EL 645'-0" SEE NOTE 9 12. THE CONDENSATE HEAD TANK MAY BE USED TO SERVE RHR AND CORE SPRAY
".,, ,o
~
n-... n' ..,.
'1 n
'1 .
co o J/4" ~
- .i TC 1o*
1 PUMP TEST 7 .25* 1
>I iJi W
1- 3/4" -
~
112*
'--'-::::!?:!::~~f,t_( TYP)
DISCHARGE PIPING INSTEAD OF UNIT 1, 2 OR 3 PRESSURE SUPPRESSION CHAMBER (PSCJ HEAD TANK, WHEN UTILIZED FOR THIS PURPOSE, THE i I >
~
~
~
~ ~ SHV-639 TV-638 1 ~
1L-I ~ 0 m' :. FE Gi'i @) RTV-78DB DRV-78D CONDENSATE HEAD TANK SHALL NOT BE REMOVED FROM SERVICE TO CALIBRATE LEVEL SWITCHES 0-LS-2-172A, 8, CANO D. (REFER TO PR0- 669 iJ *N 75-49 75-49 75-49 TO RHR HEAD SPRAY I-" 4 1-47E818-1. 04) *
~ 3/4* ~ -*~ 3/4" _ ='.) ~ I 314' RTV-21L
--.,- 3/4'
_._,_., 1 " * ._,_., VALVE 3-74-801
- 13. VENT FOR TEST ONLY.
.~ t~ ~
- 14. VALVES J-FCV-75-25 AND 3-FCV-75-53 WILL SEAL LEAK TIGHT ONLY IN
~ ~:::;---f'3':-~47__7~E81!_1~11_:-:_11_c.~H~4,'
3/4" TEST rI} --{[! TC !- I;;
~
i ~ r;;--." RTV-49H
. THE DIRECTION FROM THE REACTOR TO THE CORE SPRAY SYSTEM DUE TO F
'1;-12} '1;:2~ ~
FE \ jlRV-1J TO RHR CONTAINMENT A PRESSURE LOCKING MODIFICATION WHICH DRILLED A 1/4" HOLE IN
.lr-~ 75-21 N THE REACTOR SIDE DISC FACE OF THE VALVE *
.:t. CJ ; : SPRAY SYS FILL
.., VTV-640 SHV-641 6 co ~ ; .:t_ 'M""
-1... - -- ._,_.,1/2" "'~ I L--,:f?.?:;J::---,'i; ~--j:X,:l-,~-,:J-47E811-1.H7
- 15. VENT, DRAIN AND TEST CONNECTIONS 1-1/2" AND BELOW CAN BE PROVIDED WITH PIPE CAPS OR HOSE CONNECTION FITTINGS WHERE 1=**---l...f-;-.3/74:S.;-'"'.i g f.~ "~
D.. f"l N (_rw_J RTV-21 H :5HV-87 T ~ 1 SHV-88 REQUIRED BY PLANT PERSONNEL. THIS CONFIGURATION IS SUPPORTED
~
V I ~ ~ ~ BY ENGINEERING CALCULATION CD-00999-923399. NOTE13 ~ a.. 1 75-J9
- 16
- FOR ORIFICE COUPLING SEE 0-47W600-20. DETAIL P20.
. BYV-89 RELIEF VALVE
~ -:~
. PUMP TEST 2' "'
N
,__ 1o*_ _ _ _ _ _...:....;;;;;;_...;..:.;;.;_ PUMP TEST _ _ _ _ _ _ _ ___, 3/4" TEST
. 2* 1 .. RFV-5438 RELIEF VALVE SET+/- 500 PSIG SET o 50 P_SIG/
RFV-¥9,/ REFERENCE DRAWINGS: 0-47E8D0-1 ......... FLOW DIAGRAM - GENERAL PLANT SYSTEMS 0-47E800-2 ......... MECHANICAL SYMBOLS A FLOW DIAGRAM DRAWING INDEX
,--------'P"U::;M::,P_M:::l:.:N:.:I:.:M:;UM:;..aF.:L.:O::.W_ _ _ _ _ _ _ _ 3 _," Js-H..V.,_.51-'95=-'=T'-v.._"s*.-4-1:>1 RELIEF VALVE~ -- *N *N MEL ................ INSTRUMENT TABULATION 12"
,-~~~~~~~~~~~~~~~~~~~~~~R-T-v---2o""lr=--t.>t-'--l!l("+~~~A 2"~
SET o 500 PSIG 1" J
- 3/4" TEST - L.....I 3-47E852-2, 84 "),
<._ 3-47E852-2, G7 2* 3-47E610-85-1,-2 ... MECHANICAL CONTROL DIAGRAM-CRD HYDRAULIC SYSTEM
~~-I 3-47E610-68-1 ... ,, .MECHANICAL CONTROL DIAGRAM-REACTOR RECIRCULATION SHV-597 TV-596 SYSTEM EL 621.25 3-47E610-75-1 ...... MECHANICAL CONTROL DIAGRAM - CORE SPRAY SYSTEM PT .f---!:0:1-i 3' 3" PUMP MINIMUM FLOW MEL ................ VALVE MARKER TAG TABULATION 75-20 * ~ J-47E811-1 ......... FLOW DIAGRAM - RHR SYSTEM
~
DRW 3-47E815-5 ......... FLOW DIAGRAM - AUX BOILER SYSTEM 14* 3-47E818-1 ......... FLOW DIAGRAM - CNDS STOR AND SUPPLY SYSTEM PRIMARY CONTAINMENT J-47E817-1 ......... FLOW DIAGRAM - NUCLEAR BOILER
,-x-2230 1" BY-PASS
..I.,
,JIii-~~~ 75-18A SHVli.:\'.ATE OR GLOBE 1" BY-PASS
,j, SHV I - PT 75-48 3-47E852-1 *......** FLOW DIAGRAM -
(DRW) J-47E852-2 ********* FLOW DIAGRAM - FLOOR & DIRTY RAOWASTE DRAINAGE CLEAN RADWASTE A DECONTN E
,Jllf---~\s'\ 7~0A RTV-48 ~ DRAINAGECCRW)
BYV-18 '\. - - 1" BYV-10 1" L 0-47E830-5 ********* FLOW DIAGRAM - RADWASTE J-47E813-1 *......** FLOW DIAGRAM - RCIC SYSTEM SHV-188 VTV-18 SHV-108 vi-v-10
~ 1" BY-PASS 1" BY-PASS CYCLONE SUPPRESSION POOL LOV-574C SEPARATOR ',
LOV-574A 1.5~+,,;-;,9._4*~ 4" . 3' 3' _, ... - 1" ....._,, / eYv-4s 1" ~~--~* BYV-38 SEP-5 (JA PUMP) SEP-14 (JC PUMP) SEP-33 (38 PUMP) CKV-537 SEP-42 (JD PUMP) ;' {A,8,C,D)
,..-X-223A . SHV- SHV-
. VTV-46 SHV-468 VTV-38 SHV-388 N 574C
-w 574A 1a.:r-<.5oo*- _
N N f2l- LOV-574D LOV-5748
- 1 112*
HIGH WATER LEVEL 536'-7" 4" 1/4"_ - I 3" 3" - .)
~
- 4" SHV '
- i SHV- N 75-36
- t. ~
LOW WATER { 536'-1-3/4" WITH OW/WW hf> 1 '____/SHV\ ' *N 5748 _..- VENT LEVEL 536' 3/4" W/0 OW/WW /j,p ' >
~
,Ll--.,...J5!QOQ.O"'.....(
j ~ u 15 21 .500"-112' I 3* 4 **-to::1-L...I. PI
1----1.,,'
I - .... * -+::H RTV-16 - - 75-16
......_. Q p I '-RrlT)VX-_J-35-1 (i)
. 9375" CKV-570C 3* .9375" CKV-570A 4" ' , fsHV'- . :;i
~
PR0-6718 PRO-yPR0-668 667 PR0-671C "'-----l'/'l--l<l--1 llr - I j ~15'::'.4~ 7L5-~
~5 t-4'""-t,1>1--1' -.,J- *---1:l,I---'"
- r-1 /4" LEAK 5 .41 "
1 /4" LEAK TEST~ 1f4" LEAK PR0-671A Pl ~ j r- ~' PS CKV-5708 .9375" 1 TEST SHV-5C SHV-33C T ST~ ~ ...-R-[TXVX:_f-7....--{75-7
/pj" SHV-14C
~ *~
D m' 'm CJ 7~5 s.r PS 3" .9375" f_!l) ~ SHV-42C STRAINERS SHV-148 75-7 ,..__-l't---l...-'1--!<J-°'4-"-I
*I -
RfV-44 7~4 1/4" LEAK t ', 1 12 *' SHAFT SEAL DR 3/4'
~1._/2~'-'_,_~SH~A~F~T-"SE~A~L'--"D~R....._,.-J',J--.,4----c:4~3~~4~'*'~~
PR0-6710 CKV-570D n:sr m 3/4"
.....-....c:4--lPf--CJ.,,-.
DRV-J3e sHv-3Je PMP- / SHAFT SEAL DR 1/2" DET 3A 1 STN-663 STN-664 STN-665 STN-666 2* 3/4" SHV-HC ~ P~P- '4- DRV-148~--t~ :<'PMP-s sHv-5e oRv-5e r--1 cc N PUMP 3B 33 tf" 3 /4" TEST TYP FOR CORE SPRAY PUMPS PUMPS A, 8, C, AD
>-... I>
3/4" TEST * ' z 2
- DRV-619 CJ TV-14A TtST
~UMP ~i( ._,( ~
t TYP 4 PUMPS 1, *
.;t:
f;j
--tK:.4.li.:'--!p,.:4--r{v' PUMP 3 A [;l
~ ~ J/4 '
~
' l" foo-'-IIOlll---.:4-.../
DET 3A1 SHV-33C n TV-33A
~
Q 2"
- ii;:'
1/2" ofr 3Al 112* , 0 TV-SA in,
- 1,.
SHV-SC l/,::-IIC::'-~D-E~T.3::A.:.1~..,.*Jll.:",.~,.. 1/2" ~ : ~ ,...,..,:--11e11-1--,-....... -, SHAFT SEAL DR 1/2' ORV-33A SHV-3JA 1/2" VENT~ I')
~
CJ 7'ii;:*
~
- c.,,u 0
~
~ DRV-618 N Q SHV-14A DRV-14A
- l CASING DRAIN ~
!;* "r-VENT SHV-SA ORV-SA I
,.... a.. _ DRV-428 SHV-428 PMP-/
/-rllcol-'3"/i-4-'_T:.:E:::Sc.T..,. *~
- s. I>
0 DRW , ' C 421 Q m ~ m I 7
~ (TYP 4)
-i':;
I
~~ "' >
I j ~0 Q !; 4 i--VENT 1/2" ~ PUMP DET JA 3D1 '\.. 'SHV-42C ... N TV-42A _.. ~3-47E852-2, 84-µ r1.,... -,--11e111---.:.-.../' - { SUPPRESSION RING HEADER POOL u Z::E
<~
r~ cc_ m 1 rDRW RTV-13 Pl PI 5-DRW
~
~ DRV-42A SHV-42A 1/2" VENT~
~
.~
uN N 1/2" Pl 75-32
......_. RTV-32 1" RELIEF VALVE SET o 150 PSIG
'-~----~---~/ .,1,_._'
,- .2 "-t::ci:J--, <
QZ w< - ' ts 112* -
. - 75-13
~
RTV-4
~~
... . ~~~ N'
"'mw RFV~ (I
~ --it SHV-76 ::c2i DRW a,
. m Q SHV-75 1" RELIEF VALVE SETO 150 PSIG 3* 3*
1" RELIEF VALVE SET o 150 PSIG N I 112*
~
5078 ..l 1-1/~~ _ _1,..,_.-4_ .. 1-1/2L .. l - 4" sHV~* 4"
- )
N
~
m w 1" RELIEF VALVE SET o 150 PSIG =
+--to::]f-1_
n 75-41 PI 1/2"' RFV- SHV- CKV-S14C 2"
- CKV-514A <~
1/2' L ,)- 512C .--~-11~ 512A RFV-
~ I--'+~ RTV-41 ...__,,. N Pl :-~C<J()--'r~LL 507C
- 507A , ' C
~ ~
N 3H---i5 . l~A RTV-7SA
~ G 7 0 A RTV-75A. L '
DRW
~.-
Q ..... 3::. 1" -
, - DRW 15
- -31
{PI\ {PI\ DRW 15 61 V "' .... m RFV-5070 "'mw r N N ' PMP-76
\_ m i
PMP-75 [fil >J.
\ -~ DRV-604 DIS i";.
~
f~ c3
~
DRV-515A '\...cNDS FLUSH DR 75~3 75-64 6 VTV-67 41 N N'
"' 4 ..
LOV-5120
-- 3 ..- --I V
~
3" 3" ,
~-~128 I
. "'mw 7~8 Pl RTV-768 en
~
'---1/4",
~-
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-1/~- *
;.:;J:,-C,'.4
~
<}:."'(~
75-74
~
*~*~
cc
~
- 1. j::
I
,,.-t,Pa,M,!P-,-1--,
I " I m w
~
Ll,...:t.i)l-"--+'+---r,t--*-r--l,*---1,..-1--....
-1 SHV-
.:.!_ 20 I
CKV-514D CKV-5148 SHV-5128
-,<l-,c,-
4" ,- 7 LOV:76 LOV-75 _ :., LOV-520B"T" PUMP 3B S.iJi ...~ 2*__...;.__________ PUMP 3A ....:,"'.!..[4.L;--....;.-------- cc , .. "f-...L-~--,,...~67 I *
,3, -,5, 1/4"1io1HI 2"- - 2" 3"- ::!;
. .- I )
0:::
- {PI\
L..::..J L.=.I - -. STN-64 1O" STN-74 SHV-612 *
- SHV-614 ' . TO AUX BOILER sHV-s2.oe 75-66 DRAIN PUMP 3A
- 1" CNDS ..fs1 CASING DRAIN 2'
...f---~~~~~~~~~~~--,N ~
10*
~
FEED PUMP "A" 1-----a:3-47E815-5.B6 FLUSH DR_.. "'
~
~
- I
-Isl- VTV-68 I
Y I COMPANION DRAWINGS: 1-47E814-1 VENT 1* < r-- >
~
' ~
~
PMP-1-- 68 2-47E814-1 PSC HEAD . ~ 3-47E852-2, 84 TANK PUMPS N oRw VTV-511C SHV-510C 10" 10* STN-66 , ) $ m, N i";'
- 5**
TO RCIC PUMP SUCTION RATED CAPACITY 40GPM o 152 FT TDH
"'m 0
ii;:' DRV-11 SHV-11A FCV 75-11
~
DRV-2 SHV-2A SHV-28 VTV-2 DRAIN PUMP 3B ....._ FROM CONDENSATE SUPPLY
.~.~'
c, Ir=~~-~-
~
~
3-47E813-1,A7 Loy-5_2JA 112" 1
'>-=-- "'
SYSTEM I DRW 16" sHV-11 e Vrv--11 N I THIS DWG. A4
,/ PUMP SEAL B
FCV 1 FCVtJ-~'-tK:.4~-t<'4--C: SHV-52JA e9 -;
- SHV-J9A DRV-39 75-30 /
,:;,._;;,.T SHV-30A DRV-30
. 3/4" TEST ~SHV-SJ
..,._--IK:.4-../
3" 1;4*
~::i'(J--
53H3_s;., 7~
~
X VTV-39 SHV-398 VTV-JO SHV-308 AMENDMENT 28 8
~
LOV-5238
,~g ~<
"I r~-=~-i 1
- VENT 1"VENT 1-~=~.,;;; ~ 1/4" _
' POWERHOUSE
~ ..., ~~ ~ .
N
~ UNIT 3 j ,
~
~
N 1" VENT o 1- ;;-; 0 in Q I ;;:; ~ ~CJ ' :;!. ;: 2 1 /2" ' E:i ::!. I- LI.I ...J I- LtJ SHV- CKV-5228
~~ . . 1 , .., , @ ffi ;i........
0:::1..1.1 8::~
=i...1
~r.1 FROM CONDENSATE SUPPLY VALVE J-2-744 1
~>
' ~
.. ~ ...
i >' '"
~
5238 1[3~-4-(;4il*I *~ CCI BROWNS FERRY NUCLEAR PLANT
~ ~ ~... "'
~
- 0
_,1,-.-lli ~8 I ~ s. FROM 24" CONDENSATE SUPPLY HEADER rf1 I T 1~2 14
- 7 I 1.
11+* J ;: i 1-
- j:: ~:c
~....18 r-(J-47E818-1. ss
'1
~
SHV_)-L-I 75-40'
~
- **-t3
<. 6"?\ >
....- SHV 31
~
i:&. w r cc
"' FINAL SAFETY ANALYSIS REPORT I3-47E818-1,C7 Lav 3
- ' ' DRW a.a..
TO DRAIN PUMP JB {SEAL WATER) co
~
N I 1* '"2" I 1* *2* PUMP SEAL I - - v N ' FROM 2+* CONDENSATE " ' " " ' " ' - " ' - - - - - - - - - - - - - - - ' - * " ' - " - - - - - - = : : : . . . = = - - - - - - - - - - - - - . ; - - - - - - - - - - - - - - - - - - - , [ ) c X ] , - " - - ' - - - , : T H I S DWG. 82 "' m A SUPPLY HEADER 14 .. I
/>------------------------------------------------------------.,------------------------------------------~-----
I sHV-J1so 1
'V" 14 ..
w
~
. CORE SPRAY SYSTEM
.m 13-47E818-1,C7 I 2" 2"
()' N FLOW DIAGRAM SYSTEM II 16" I 4" 16" 2-1/2" FIGURE 6.4-4 8 I 7 I 6 I 5
- 4 I 3
1 2 3 5 7 8 9 10 11 12 9ZO~ l-fL-0 l93Lt- l N L9 NOTES:
,- --j1 -730E928-2, _, I 1. THE HIGH PRESSURE COOLANT INJECTION SYSTEM IS DESIGNED TO COOL THE CORE WHEN THE REACTOR VESSEL IS ISOLATED AND I REACTOR FEEDWATER IS UNAVAILABLE. WHEN THE WATER LEVEL IN THE REACTOR VESSEL REACHES SETPOINT OR THE DRYWELL PRESSURE I REACHES SETPOINT, THE HPCI SYSTEM IS AUTOMATICALLY STARTED A
9-3
-r----r-------------------------------------- BY ACTIVATION OF REACTOR LEVEL SWITCHES LIS-3-58A, LIS-3-58C, LITS-3-58B, AND LITS-3-58D, OR DRYWELL PRESSURE SWITCHES PS-64-58A-D (SEE 1-47E610-3-1 AND 1-47E610-64-1) ARRANGED A
FROM 1-TR-69-29
!1-47E610-69-1,G2r .........
1 I I
@' -c IN ONE-OUT-OF-TWO-TWICE LOGIC TO OPEN VALVE FCV-73-16. THE SAME SIGNAL OPENS PUMP DISCHARGE VALVE FCV-73-44 AND CLOSES DRAIN LINE VALVES FCV-73-6A, FCV-73-6B, FCV-73-17A AND FCV-73-17B.
I
- 2. A MECHANICAL SPEED GOVERNOR PREVENTS TURBINE OVERSPEED I
TO 1-TR-69-29 r- --------------------------------------------- 1
---, BY TRIPPING THROTTLE TRIP VALVE FCV-73-18 WHEN A TURBINE OVERSPEED OCCURS. A CONTROL GOVERNOR, FIC-73-33, EQUIPPED
/~- --------------------------------------------- ---, I WITH AUTOMATIC AND MANUAL MEANS FOR SPEED SET POINT CONTROL
~-47E610-69-1,G2 ~ - - - ---------------------------------------------
1 1 --~ I I REGULATES TURBINE SPEED VIA "TURBINE SPEED CHANGER" TO CONTROL HPCI MAIN PUMP DISCHARGE FLOW. REACTOR WATER ' CLEANUP '-- _../TE'. 7~A I I j-1:IT~ Jc:iill_~ 7t:;;B Ire\ 7t:;;C Ire\ 7t:;;D
- 3. DELETED
+. THE TURBINE CASING IS PROTECTED FROM OVERPRESSURE BY TWO HPCI ROOM I I Eill._~ TORUS R4-U TORUS R3-U TORUS R2-U RUPTURE DISCS. PS-73-ZOA THRU D SET AT 10 PSIG DETECTS DISC 8 EL 519' I L--------f-m-~ EL 550' EL 550' EL 550' RUPTURES AND ANNUNCIATES ON PA-73-20 IN THE MAIN CONTROL 8
+----r------, ROOM.
z,~,~-1 1,~~o I FROM CNDS .t 5. THE HPCI TURBINE WILL BE AUTOMATICALLY ISOLATED FROM THE DMNRLZ WATER I ~ 9- HsI ~ 1 ,o 47E610 2 2. GJ I STEAM SUPPLY BY CLOSING ISOLATION VALVES FCV-73-2, FCV-73-3 AND FCV-73-81 1F.
)a{.... _ _,.~
1
.l1-+5E779-1>~-1 . I 7~A I STORAGE SYSTEM A. A HIGH PRESSURE DROP OCCURS ACROSS THE HPCI FLOW f- - -
r I I I ~ 11 NOZZLE FE-73-1 BECAUSE THE STEAM LINE RUPTURES. fi
! 1-+5E714-2 ~-19 THE HIGH FLOW IS SENSED BY EITHER PDIS-73-1A OR 9-J
~Efil_ I------"tiI II I i ~ 9 - J MIMIC PDIS-73-1B AND ANNUNCIATED BY PDA-73-1.
Zl-7J-40A ~IL-73-188 a-, B. HIGH AREA TEMPERATURES INDICATING A STEAM LINE BREAK m-i ~_,
~12:J"'-----rl tL-3:sz"-7>-> ).._
I ©_, ~->1 (PI\5-50 p;,-,o I ARE DETECTED BY ANY OF FOUR SETS OF TEMPERATURE SWITCHES (TS-73-2A THROUGH TS-73-2S). 1 I 1 I I 1 II , U:07'-596-3 LS r .-[_ L~.) I ~ ' ©*-> ~ I 3-4 - - , - 4A EACH SET IS CONNECTED IN ONE OUT OF TWO TWICE LOGIC. FOUR INDEPENDENT TEMPERATURE SENSING ELEMENTS
.../HS\ 1.- 1~A 1~8 4-P. 73-40A
~
73-18A y 73-188 y I 7~-:!B (TE-73-55A THROUGH TE-7J-55D) ACTIVATE THE MAIN CONTROL CISS I I y I I I I Y 21-13-400
~
A 9-J I IL-1,-2 r- 1,-,8 I 1 , 1 1 ~ - -, 1 --~---1 ----1---1--wi I ,~L-oG~, I I I ......... ROOM ANNUNCIATOR TA-73-55. C. LOW REACTOR PRESSURE IS DETECTED BY PS-73-1A THROUGH t,~~ -l: ti C MIMIC I 'l;;'{ I .......,. I I I I I IDIG0301 I I I PS-73-10 EACH SET AT 100 PSIG AND ARRANGED IN C
~~ I Zl-73-2 MZI-73-2-~+MCC IL-73-3 Zl-73-38 I I I I I I I I I +--.J ONE-OUT-OF-TWO-TWICE LOGIC.
~fil--,I I~~~~
I 9-> 1 FCV I
>- 2 _J
~MCC L
~~
II Iro FCV-73-26 1 l1-45E714-2f-4----1-----_J A -21 I '- c.::,.,
-< l~:-*.1,,-- - - - - ! - - - ~
73-40 -'I FRo.1 FCV-73 73-40B
-26 a, -27 ..._,
r---.j.,----+----.J I r TO XM-73-19
-jrn,s owG. 09 ) r-l 1-+5E71+-2 !
- 6. NORMALLY MAKEUP WATER FOR THE HPCI PUMP SUCTION IS SUPPLIED BY THE CONDENSATE STORAGE TANK THROUGH FLOW CONTROL VALVE FCV-73-40. AN ALTERNATE SOURCE OF WATER IS FE L L_J_']'_j__.l..._ r:._J -- T - - - - - ' _J !THIS DWG. J>f--' MAIN STEAM SUPPLY {R'f IL-73-40 L-jrHIS DWG, J4-5! I I AUTOMATICALLY OBTAINED FRc>>.i THE SUPPRESSION POOL THROUGH I ISOLATION VALVES FCV-73-26 ANO FCV-73-27 WHEN THE CONDENSATE
,---{7
~v " - - I I ~~
MCC I I TO WOODWARD GOV f---..J I *,
~
,_, HEADER LEVEL DECREASES TO EL 552'-6". PUMP DISCHARGE FLOW coNDENSING
~ I x-11 r-Tf-_______ ~~
~ ~
z1-1J-s1e z1-1J-s1A
~~ rm-~ -
I I JO PSIG 1-730E928-2
~ I {ii:).
9-3 Zl-73-16B
~::::i::;;T (Hs'\
3- 5 ROUTES ARE AS FOLLOWS: A. NORMALLY VIA THE FEEDWATER LINE TO THE REACTOR LJ-m POT " r- C: IL-73-16
~ ~
THROUGH ISOLATION VALVES FCV-73-34, FCV-73-44, AND
":,..[::-- z 1 _,._,,._.r~....1 , 1 I ~ _ - --' e1 \ I 73-18 ~ ~l1J-16B 9_391 -< 1-¥MCc CHECK VALVE FCV-73-45.
- e!. I ~ L ___
- 9-19 r------73~ 7~::r' LOG ~ Rs\_ F~V ~ / ...._ I 73 B. DURING TESTING VIA TEST LINE TO CONDENSATE STORAGE THROUGH FLOW CONTROL VALVES FCV-73-35 AND FCV-73-36.
i:: I {Hs'\ ~
>-1A -:3-18 "
~ r-,.
'f I
I
,I Mee+ 73-811
~
IL-73-81 11I rlo-+o*n*-<> 11
- ,1-730E92S-1 I y 73-818
~-> L 7J-8JA c1ss B
LOG 73-81-CL ~ 7~8 ~
~
I 1 7'
~
PI 18A C...,.. DRAIN PCV 9 - 3}&:- {zs'\ 73-18A 21-73 7'J - 188
*v l
- Zl - 7'J ..J w -¥ t'J:;: 3 fH5' fi.........
1k: I 7. C. MINIMUM FLOW BYPASS TO THE SUPPRESSION POOL THROUGH FLOW CONTROL VALVE FCV-73-JO. IF HPCI SYSTEM OPERATION IS REQUIRED WHEN IT IS IN THE D - Lt--f-------~---' I rl I -1FCV -1.6~-. 73-16-A 7"~S16C R G _J D W.1_.LLJJ_.Ljf---.L--~~--=:-' I ....l.25-7 J....25-7 I ---- TO CNDS 0-47E610-2-2, f3 A DMNRLZ
> 25-50 I) d - J I 73-~8 f Pl \ 7~-!,!'C PCV t 7J-1~'-'J'---'-'==.,_ ~ J§l _____ ....._... ~ ......-'"-1*"
.i..;:,i.,.;:,uo,'--,
8. TEST MODE, THE CONTROL SYSTEM IS DESIGNED TO AUTOMATICALLY RETURN THE SYSTEM TO THE OPERATING MODE. COOLING WATER FOR THE PUMP. TURBINE, AND LUBE OIL COOLER nnnn111111nn- - ~ I (!s.~ (!sj I I rir-73-362"1i {&A ~~5~~ R§. ijijijijijijijijij 111111
- ~*7 PDT I 25-7 7yA
+-
I ~s-7 ~;:.1-7yC
--+- _J I II II R jQ_
WATER STORAGE SYS rl1-45E71+-+ I I 7j_T28 i-_- 0_;----,---_J'
--~- I i
x&t SPEED7l:1jlll..Z_J-18 DRAIN
>St-9;:_73-18 7c_!JA ZS T
@ 3-1 3-1 73 TURBINE INLET DRAIN POT MCP-73-5 9.
IS SUPPLIED FROM THE PUMP DISCHARGE LINE THROUGH PRESSURE CONTROL VALVE PCV-73-41 SET AT 75 PSIA. UNIOS ON DRAWING ARE FOR REFERENCE ONLY AND ARE ABBREVIATED TO MEET SPACE CONSTRAINTS. REFER TO MEL FOR COMPLETE UNIDS .
*: I,,' °"\*=** -:,- *~* 1 T I 73:J..B I ( PS' ( PS' I I
I I 7~A L,1-1JoE92a-1 tI
~
- DRAIN p ~
Cl
~
" - ~
I ""JC,(.. ~ ~ I ~
- 10. ALL INSTRUMENTS ARE PREFIXED UNIT "1-" UNLESS OTHERWISE NOTED.
~I
** r-=-------.J 1 1 1>-1* 1,-10 1 73-29A PI -
>; ~!1-47E610-11-2)-jo1v¢"1-_J I *y* *y* I ~--' ~ - 6,
~-> ,--
30 PSIG r-----------~ ~ RAIN 7~c o 9-3 L1~ SM'4 xs). (__PI_)- 1.L;:.-" 1~9 _ 1~*
-µ__PI_) 11. COMPLETE LOGIC FOR HIGH PRESSURE COOLANT INJECTION IS SHOWN ON MECHANICAL LOGIC DIAGRAMS LISTED IN REFERENCE DRAWINGS.
I Eccs-ANALOG I I LOG ffi'ss J ~I cc:_, 12. DELETED
~
TRIP POWER ~* I I ===,---=]-- - 1 (_scy_ v~~~t* IN
~ Bo:: ~
_lJ,-81 -c .,.1_9-82 bis) II HS7381A 8 ~--=L=~~R=*~-l~L:.~7:-3:-:-'-'------<F~R~DM~R~C~l~C~SY~S~T~E~M
, ,1 47E610711.E6 1>..;:~-1 7~8 1
I I I r-~--, I I I I I I 1,-1e-r1..:.:;2.e
/ xM 5_+!._ 1_ 1_
r-----, 1 ..11.&1 E IVV * * ** 0 <... 73-1A i 73-,s i E i:: 1~ y rj1-45E62o-q y r--T-----,-j1-45E61+-1D I '" I 25-6> -1.... I A I A I A I I
--1.... A I I
-1.... I l'y,'.1~
~ 9->
** * * .: i'l:!.
.i: ~e:
I I L...~---
~-,
3-1 I
--L-~--_J I . 1 21 _7 ,_-.CR 1
~ ~
+~a rl ----.-j1-1,0Es28-1 I
I PS (re\ ( TE' ( TE' ( TE' (TE\ ( TE' (re\ I L-r~ I I 1i-:.,H 1i-;,G
,J
{HS\_lfisv\ {HS\_lfisv\ I >>A L-jo-+5E71+-> 7~9 7l-;tJ 73-~E 7t"~B 1t-;,c 7t-~l 1 7b9A : 7 PI 1 ;;;; 102 1J-s2 1~3 1~ -~J 1 (ru 9-3 _j (TE\ I ___[Ig[J I~ 7~7 8 r ~ ::- ZI-73-45A ct
~
- - r ~
.1.3 uh
~-:r I 73-35A
~
{PI\5-50 @_,, iI -
.-~'-------,
HPCI BOOSTER PUMP PMP-73-29 f-- SPEED CHNGR GR-7J-54 _
-~_,._-I I/ -
HPCI MAIN PUMP PMP-73-54 - - CPLG-73-54A
~
,<,//<:T-1 HPCI TURBINE TRB-73-54 r-'-1 73-54F
~
)
~
_ J__SE_ 1.3-~~1
~
-+
I I
--.uu 9-3 1.3-~1
..\
f'j"
~
1 I I 1 1 C DISC~f POSN 73-513 L"*MCC IL-73-35 @ l,J-J1B 25-50 9-3 L- ,__1 _____. LOG A4165
...=.
t- - , - -, I 0 -47 ~ UL\ r===={:J,~ Il T J£>r.::!----'; ~ ~ ~ ~ r-, PT LcPLG-73-5+C r----, F --T- ~ --I TE \ _!,.. ~----~-~ A /:':"\. I'/, .,1. @9-3 I") co x-,-, F
- r~i
!oigg2,! 7~ 1 ~78.fA X-9A CPLG-73-548 I ' PI' f_!l.J-- //'** PA Vl IO ,.!, :c I u l 1~~15 ~ NJi ~I
.,f 73-54-1+ 73-32 r
11-730E928-Bf-i_ L,._:;;;;;;;;;;;;;;;;;;;:...:;.;;,.:i:;3:...-;:::;:;:::;------.l,~t----:--:--_;c:;R::;*:.__J__ L- 73-32 rw FE 7._3-~3 J--------.-----------------P. . P. . u...M.. o_,_sc__HA..R..G..E___I - - - - - I ABANDON IN PLACE _J r,,.., I'\.
- .."' 7...3-48 73-49
/;
// /
'--ABANDON I IN PLAC~25 _63 y -
O u o
~
"j" O Cl 1:l .....
LL Cl N Cl eia:: :c"'.... Vl . .
~
Cl I I Zl-73-6A I I
~-J
~;Q.
Zl-7J-68 II II
~
- .... - - - - __ ___)j;<MCC t-----.':::"-¥Mcc, -' I :c r-.,
' ID1\~~2sl I I ~ I fiL-73-34 r ~
'JC{' "25-63 - ,-- "° PS ~ .=. .._ .!:.
Ill: i...
- c
.!::. I I I I C 9-i cy FCV 73 44 IL-73-H CV 73-34
'-(Re
~-1_~1~}-
7~3 "i co E~
'7
- I 7J-j.6 e: I I I I (FsN\._
73-6A _..J_ T ..£fl'...
~
I I ZI-JJ-4-4-A 9-3 I 1 I IL-73-34A 9 _3 L..- - - , 25-63 g ~ z
~
....w
~ .>
~
~ r ___ _J I
PRESS. TEST
.-. ,--1 X 'j?...:!'.._'.l'.._l)-,,*+--j"'......-,** I I r----' I I r---.J 1 ~
L<lli 1 I I I
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~
r-rI @A Yr-* L-1-....1
~
w
~
w
....< ....< I ,-,::{-'
V
~
I I I I I I REFERENCE
- r----: ~~l ~1 DRAWINGS, f
- ......... I z
~ -.... ""' i....td-1 0
G lL-7J-JO u
~
z "z ! 1 0
*'* r ________~ I L_.J I I I 1-47E605-SERIES ...... MECHANICAL LAYOUT OF CONTROL BOARDS 1-47E600-SERIES ...... MECHANICAL INSTRUMENT AND CONTROLS G
~ "' 1-47E610-64-1 ........ CONTROL DIAGRAM-PRIMARY CONTAINMENT SYS
~ ~
]Bl J§l L--11 I
r __ .J. _ _JI 0 0
!1-730E928-7f,----..J MCC :~5--6~25 1 , - - - - - - - + - - . J 0-47E800-2 ........... MECHANICAL SYMBOLS AND FLOW DIAGRAM lo-45E714-3
, r: __ _J I t"' r---1--r---, j zi-1,-,0* /'HS\..u. 11 FCV .~
8 0 u X
~
u 0 X
~
ps ps 25-50 I I (FsN\._ __ J DRAWING INDEX 1-47E1847-SERIES ..... CONTROL AIR SUPPLY
!1-73DE928-6~1
~J l1-1,0ES28->I-.J 1 I Loe 1,_,,I
- I I
I
.-J,.z- 3 LeJCl 1,-,08 : I I I 1Li-,
L.......,.OG
,-,-"Tu--,
I - "' z 0
~
~
w
~
w I
~ - - ~ - - ~ L--1-+----+
I I7~A 7~8 I /.1....1.5-6>~>
@,-19 (PI) 73-;J 8 I I 11
- 73-~8 Bi 1-45E614-10 .......... WIRING DIAGRAM 120V AC/250V DC VALVES AND MISCELLANEOUS SCHEMATIC 0-45E714-J ........... WIRING DIAGRAM 250V DC RMOV BD 1A SCHEM
!1-+5E71+-2
~
~
I I I I 7J->> C4057 ..ls- L_ I w-J I
.w
~
~
0 z
~
"'0 z
~
~
7~,8
- 25-6>
1u2 PS 1i-9A 25-6> 1( PS' ( PS' 117~C 7~0 *-> I y 20-00 X 'j?...::..:..-0-,,4==:t-""l*j V ~ 1-45E714-2 ........... WIRING DIAGRAM 250V DC RMOV BD IA SCHEM 1-45E779-12.-13 ...... WIRING DIAGRAM 480V S/0 AUX PWR SCHEMATIC 0-45E779-4J .......... WIRING DIAGRAM 480V S/0 AUX PWR SCHEMATIC
"' " " (n'\. I PT 1-45E714-4 ........... WIRING DIAGRAM 250V DC RMOV 18 SCHEMATIC ZI-73-23 TO XM-73-19 ~- I 3 I LOG 1 73-JOA I I I ~
1-47E610-55-SERIES ... CONTROL DIAGRAM ANNUNCIATOR SYS
);;'( ~
't'~':(3 (THIS OWG. F1D f - - - ! - - - ~ . J ID!G0291 I
'l;;'{~
I I I I 7J-21A "t 73-21'1-~--1
......... rdir, ......... ." 1-45E670-SERIES ...... WIRING DIAGRAM ECCS SYSTEMS 1-47E812-1 ........... FLOW DIAGRAM - HPCI
. - TURBINE EXHAUST ZI 30AL--"'Ct .....J I lJBII TO MAIN "'~ 1-47E-610-1-1 ........ CONTROL DIAGRAM - MAIN STEAM SYSTEM 0-47E610-2-2 ......... ~~~\~iL DIAGRAM - CNDS & DMRLZ WATER STORAGE H TURBINE EXHAUST CONDENSER C, :2 fij
>1 1-47E610-3-1 ......... CONTROL DIGRAM - REACTOR FEEDWATER SYSTEM H
- . L..-1--1* mo,- 0-47E610-65-1 ........ CONTROL DIAGRAM - STANDBY GAS TREATMENT SYSTEM RCIC SYSTEM '
' ! ~J DRAIN POT MCP-73-8 ~~~--(~.,_,.___.
~-3 oz -w zw
< w ,..."
1-47E610-69-1 ........ CONTROL DIAGRAM - REACTOR WATER CLEANUP SYSTEM 1-47E610-71-1 ........ CONTROL DIAGRAM - RCIC SYSTEM 11 47E610711,Bi' -- . -
. 11-7JDE928-4 f- +---, I I
~
"' ..,w
~ 0 e::
< 0-47E800-1 .... , ...... FLOW DIAGRAM - GENERAL PLANT SYSTEMS 0-47E800-2 ........... MECHANICAL SYMBOLS AND FLOW DIAGRAM DRAWINGS INDEX I 1--.!.- LS~ r---,---~---fl];) MEL .................. ~CHANICAL INSTRUMENT TABULATION - HPCI A 11-45E620-1 f-..l. __ ...J 73-8A.: y8B I t--j1 -730E92B-8 ! 45E615 SERIES ........ CONTAINMENT ISOLATION STATUS SYSTEM (CISS)
I 1-730E928 SERIES .... HPCI SYSTEM ELEMENTARY DIAGRAM
~~--J ~
Gi:S-3 73-57
~:
~ I I
I I L____ _ I TO FCV-73-26 a. -27 (THIS DIG, J4-5~---t I If ;t
~
~
*r:: PRESS.
TEST ~ PRESS. TEST I I I TERRY TURBINE CO.: C888-X ...................... HPCI TURBINE GRAHAM MFG CO. : I r, f PCV\ I GLAND SEAL CONDENSER C16720-1 ......... ,., ........ HPCI GLAND SEAL CONDENSER
-!--i + ...l..----------,.,AJ~
~
t-H-S.c~~~~6- A + - ~ - l r<THIS DWG, 111 FROM LS-73-57A/578 FROM LS-73-57A/57B
!THIS OWG, 11>- 1 7~
I BLOWER BLW-73-1D I ,--__:.I_.._
" I GLAND SEAL
~:fi:
I I FROM LS-73-56A/56B FROM LS-73-56A/56B I fl]\,__+ LS L-f.. L~} TO LT-64-54 I\
--11~1----11--1 I r(THIS DWG, col !THIS OWG, co>-1 I CONDENSER CND-73-703
~
~ I 7~ ' 7~ 8 1 +7E610 6+ 1.A6 > '\ J fCISSfA i I I I r ____ J --!o..,..,...=,-_...,t- IL-73-158 IL-73-15A I Cus'i-3 I I ~ ~ I ~-,:~-, ~-,
I I 11-730E928-7u 7J-26A -l....f-- 73-26B
. I fl ...__.. I ...__..
~->
73-27A
-+1 I 73-278 I ..._,
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BFN-27 6.5 SAFETY EVALUATION 6.5.1 Summary In order to satisfy the safety design basis, four means for emergency core cooling are provided. They are: High Pressure Coolant Injection (HPCI) Automatic Depressurization System (ADS) Low Pressure Core Spray (LPCS) Low Pressure Coolant Injection (LPCI) These are in addition to the feedwater, control rod drive, and RCIC Systems which also supply core coolant. For reliability, each emergency core cooling system uses equipment with as few components required to actuate as feasible, and those are operable for test purposes during normal operation except the inboard isolation check valves can only be tested during cold shutdown. To provide diversity, two different cooling methods are provided--spraying and flooding. Evaluation of the reliability and redundancy of the controls and instrumentation for the Emergency Core Cooling Systems (ECCS) shows that no failure of a single initiating sensor either prevents or falsely starts the initiation of these cooling systems. No single control failure prevents the combined cooling systems from providing the core with adequate cooling. The controls and instrumentation can be calibrated and tested to assure proper response to conditions representative of accident situations. As stated in the safety objective and in safety design basis 2, the ECCS is to remove the residual and decay heat from the reactor core so that fuel clad melting is prevented and fuel clad damage is acceptably minimized. The design basis used is that nowhere in the core will the fuel cladding reach the calculated melting temperature of 3370F. The intent of the ECCS temperature criterion is to prevent gross core meltdown or clad damage. Since core cooling cannot be easily shown to be effective unless the geometry is defined, a no-cladding melt criterion should suffice. However, because under some conditions highly oxidized Zircaloy is known to fracture upon cooling, concern exists that the core geometry may not remain defined after cooldown. Based on the worst case experimental data, cladding fragmentation upon cooldown can be prevented for the time scale of interest here, if the maximum cladding temperature is limited to less than 2200F.1 Thus this is the design temperature 6.5-1
BFN-27 criterion against which ECCS performance must now be judged, i.e., the systems shall prevent cladding fragmentation upon cooldown which translates to a peak temperature of 2200F. It should be noted, however, that since the cladding fragmentation would not occur until after the core is cooled and flooded, it is most likely that a considerable portion of the core could be fragmented without leading to gross core melting. In addition to this margin inherent in the criterion the actual performance of the core cooling systems is such that peak temperatures lower than 2200F result across the complete break spectrum. All of the safety design bases for the ECCS are shown to be met by the previous descriptions, the referenced descriptions and evaluations of the individual and combined ECCS. Peak cladding temperatures are determined in accordance with approved analytical methods. 6.5.2 Performance Analysis The manner in which the ECCS operates to protect the core is a function of the rate at which coolant is lost from the break in the nuclear system process barrier. The HPCI is designed to operate while the nuclear system is at high pressure. The LPCS and LPCI are designed for operation at low pressures. ADS is provided to automatically reduce nuclear system pressure if a break has occurred and vessel water level is not maintained by the HPCI and the other water addition systems. Rapid depressurization of the nuclear system is desirable to permit flow from the LPCS and LPCI to enter the vessel, so that the temperature rise in the core is limited. If for a given size break, the HPCI has the capacity to make up for all the coolant loss from the nuclear system, flow from the low pressure portion of the ECCS is available for core protection before the nuclear system pressure has decreased below the pressure at which the HPCI turbine steam stop valve shuts due to low steam supply pressure (see Subsection 7.3, "Primary Containment Isolation System"). Adequate net positive suction head (NPSH) is provided for the ECCS throughout the loss-of-coolant accident. The ECCS response and resulting peak cladding temperatures have been evaluated. These evaluations were performed for a complete spectrum of break sizes and postulated single failures. The results of these evaluations demonstrate 6.5-2
BFN-28 that adequate core cooling capability exists over the entire spectrum of break sizes even with a concurrent loss of normal auxiliary power and postulated single failures.30 6.5.2.1 Analysis Methodology ECCS performance was analyzed using the approved EXEM BWR-2000 methodology13 for AREVA fuel analyses. Browns Ferry specific modifications to the EXEM BWR-2000 methodology are documented in Reference 30. These modifications were approved by NRC for application to Browns Ferry Units 1, 2, and 3 in Reference 32. These methodologies evaluate the short-term and long-term reactor vessel blowdown response to a pipe rupture, the subsequent core reflooding by ECCS, and the final rod heatup. The reactor vessel pressure and liquid levels, ECCS performance, and other primary thermal-hydraulic phenomena are predicted as a function of time. These predictions are then used to determine the cladding temperature under various accident conditions. The Code of Federal Regulations (10 CFR 50.46) outlines the acceptance criteria for ECCS analysis. A summary of the acceptance criteria is provided below. Criterion 1 - Peak Cladding Temperature - The calculated maximum fuel element cladding temperature shall not exceed 2200F. Criterion 2 - Maximum Cladding Oxidation - The calculated total local oxidation shall not exceed 0.17 times the total cladding thickness before oxidation. Criterion 3 - Maximum Hydrogen Generation - The calculated total amount of hydrogen generated from the chemical reaction of the cladding with water or steam shall not exceed 0.01 times the hypothetical amount that would be generated if all the metal in the cladding cylinder surrounding the fuel, excluding the cladding surrounding the plenum volume, were to react. Criterion 4 - Coolable Geometry - Calculated changes in core geometry shall be such that the core remains amendable to cooling. Criterion 5 - Long-Term Cooling - After any calculated successful initial operation of the ECCS, the calculated core temperature shall be maintained at an acceptably low value and decay heat shall be removed for the extended period of time required by the long-lived radioactivity remaining in the core. 6.5-3
BFN-28 The conformance with Criteria 1 through 3 is demonstrated by performing LOCA analyses and the conformance with Criterion 4 is demonstrated by conformance to Criteria 1 and 2. Compliance with Criterion 5 is addressed per Reference 29. AREVA Fuel LOCA Analyses The evaluation model used for LOCA analyses of AREVA fuel is the EXEM BWR-2000 LOCA analysis methodology described in Reference 13. Browns Ferry specific modifications to the EXEM BWR-2000 methodology are documented in Reference 30. These modifications were approved by NRC for application to Browns Ferry Units 1, 2, and 3 in Reference 32. This methodology employs the following three major computer codes to evaluate the LOCA system and fuel response: RELAX, HUXY, and RODEX2. RELAX is used to calculate the system and hot channel response during the blowdown, refill and reflood phases of the LOCA. The HUXY code is used to perform heatup calculations for the entire LOCA, calculating the PCT and local clad oxidation at the axial plane of interest. RODEX2 is used to determine fuel parameters (such as stored energy) for input to the other codes. The analysis uses models, inputs and assumptions consistent with all the required features of 10 CFR 50 Appendix K to calculate the Licensing Basis PCT. Once the limiting LOCA break size, type, location, axial power shape, and ECCS single failure are established, HUXY heatup calculations versus exposure are made for the reload nuclear fuel design to determine the maximum average planar linear heat generation rate (MAPLHGR) limits, PCT, and local maximum and core average cladding oxidation fractions. The reload fuel-specific analysis results and MAPLHGR limits are documented in the Reload Licensing Analysis (Appendix N of this UFSAR). To conform with 10 CFR 50.46 and the LOCA analysis licensing methodologies, the Licensing Basis PCT must be less than 2200F. 6.5-4
BFN-27 To demonstrate the conformance with the ECCS acceptance criteria, the LOCA analysis is performed for the full break spectrum. Various single failures are also investigated to identify the worst cases. Table 6.5-1 shows the plant operating conditions assumed in the analysis. Table 6.5-2 shows the ECCS equipment capacity assumed for the analysis. Table 6.5-3 identifies systems available in core cooling for given combinations of break locations and single failures. One active single failure within the plant is postulated to occur concurrent with the pipe break. Additional details about the AREVA LOCA analysis bases, plant operating conditions, ECCS equipment performance and timing can be found in Reference 24. A seismic event is neither postulated to occur concurrently with the LOCA nor as an initiator of the pipe break. 6.5.2.2 High Pressure Coolant Injection System The HPCI is designed to provide adequate reactor core cooling for small breaks and to depressurize the reactor primary system such that the LPCS and LPCI can be initiated. A detailed discussion of the performance of the HPCI, in conjunction with the LPCS and LPCI, is given in paragraph 6.5.3. Table 6.5-3 lists the ECCS which are available for both recirculation suction and discharge breaks following an assumed single failure. When the HPCI begins operation, the reactor depressurizes more rapidly than would occur if HPCI was not initiated due to the condensation of steam by the cold fluid pumped into the reactor vessel by the HPCI. As the reactor vessel pressure continues to decrease, the HPCI flow momentarily reaches equilibrium with the flow through the break. Continued depressurization causes the break flow to decrease below the HPCI flow and the liquid inventory begins to rise. This type of response is typical of the small breaks. The core never uncovers and is continuously cooled throughout the transient so that no core damage of any kind occurs for breaks that lie within the range of the HPCI. An analysis has been made to determine if any carryover occurs in the steam supply to the HPCI turbine which could have a detrimental effect on turbine operation. In the case of a break in a liquid line, when the HPCI is energized, the level in the reactor vessel is low enough to prevent carryover in the steam which leaves the reactor vessel. In the case of a small break in the reactor steam region simultaneously with loss of normal AC power, reactor scram, recirculation pump coast-down, and loss of feedwater, analysis shows that the initial decrease of pressure in the reactor results in no significant level swell and no carryover of water into the steam supply to the HPCI turbine. HPCI cold water quenches any steam formation in the downcomer region. After the HPCI has been operating, and as the level rises in the reactor vessel, natural circulation within the vessel becomes 6.5-5
BFN-27 established and any steam to the HPCI turbine passes through the steam separators and dryers eliminating any moisture carryover. It is concluded that a mechanism to cause bypassing of the steam separators, by the swelling steam water mixture, is not available. Therefore, gross moisture carryover to the HPCI turbine should not occur over the range of steamline breaks of interest in this system. The HPCI turbine has been designed for high reliability under its design requirements of quick starting. Moreover, the turbine has adequate capacity to accept the small losses in efficiency due to any credible moisture carryover, since HPCI turbine efficiency is not of paramount importance. The feedwater spargers are utilized in the reactor for HPCI injection. Each sparger is mounted to the inside reactor vessel surface. The thermal sleeve is attached to the sparger midpoint; however, the sleeve is not welded to the vessel nozzle. Therefore, the feedwater sparger is removable. The spargers are mounted in the vessel at one elevation to distribute the feedwater in a symmetric pattern about the vessel axis. Each sparger is supported by the thermal sleeve and a bracket mounted to each end of the sparger. Provision is made for the differential expansion between the stainless steel sparger and carbon steel vessel. Radial differential expansion is taken up by the slip fit of the thermal sleeve into the vessel nozzle. Tangential differential expansion is taken up by tangential slots cut in the bracket mounted to each end of the feedwater sparger bracket. The sparger is analyzed assuming the thermal sleeve is welded into the nozzle. Additionally, pressure differentials, jet reactions, and earthquake loadings are all added; these stresses within the sparger are all within ASME Code Section III allowables for Class A Vessels. The resultant bracket loads meet the loading criteria given in Appendix C. It is concluded that design basis 8 is satisfied. 6.5.2.3 Automatic Depressurization System When the ADS is actuated, the flow of steam through the valves provides a maximum energy removal rate while minimizing the corresponding fluid mass loss from the reactor vessel. Thus, the specific internal energy of the saturated fluid in the reactor vessel is rapidly decreased causing pressure reduction. Once the reactor vessel pressure has decreased below the shut off head of the LPCS and LPCI pumps, these low pressure ECCS will provide adequate core cooling flow. Therefore, the ADS provides the backup for the HCPI. Table 6.5-3 lists the ECCS available for both recirculation suction and discharge breaks following an assumed single failure. Actuation of the ADS function does not require any source of offsite power. The main steam relief valves require DC power from the unit batteries for control and air 6.5-6
BFN-28 power from accumulators for operation. This satisfies safety design basis 5. The accumulators and the nuclear system main steam relief valves are within the primary containment and this satisfies the containment isolation requirements of safety design basis 6. 6.5.2.4 Low Pressure Core Spray System The LPCS, in conjunction with other ECCS Systems, is designed to maintain continuity of reactor core cooling for a large spectrum of loss-of-coolant accidents up to and including the design basis double-ended recirculation line break. The integrated performance of the LPCS, in conjunction with other ECCS, is given in paragraph 6.5.3. Performance analyses of the reactor LPCS are based on an analytic prediction of the reactor vessel pressure and mass inventory as a function of time following a postulated rupture of the coolant system piping. The LPCS alone cannot protect the core below certain break sizes. This is because vessel pressure does not drop rapidly enough to allow sufficient injection before the cladding hot spot reaches excessively high temperature. Below this break size either the HPCI or the ADS extend the range of the LPCS to breaks of insignificant magnitude. For the limiting design basis accident, a combination of the LPCS and LPCI provide sufficient cooling to meet peak clad temperature limits.30 Table 6.5-3 lists the ECCS available for both recirculation suction and discharge breaks following an assumed single failure. Experimental tests have shown that the quantity of flow currently being provided for core spray is greatly in excess of the minimum actually required for satisfactory core cooling3. The tests showed that more than the minimum flow required is readily attained for every fuel assembly. Other tests include evaluation of the effects of updraft caused by steam flow through the core or evaporation of the water that enters the fuel assembly. The effects of updraft are minor. A series of tests were performed to obtain design data relating to distribution of core spray coolant over the top surface of the reactor core. The topical report3 contains a description of the test facility and plots of the significant results from the tests. The core spray tests also provided experimental effective heat transfer coefficients, thus enabling correlation of the core heatup model with the actual test data. Data from tests on an exact prototype at power resulted in volume percentile temperature distributions. The close correlation between the peak temperature and general trend demonstrates the adequacy of the analytical models employed. 6.5-7
BFN-28 To assure continuity of core cooling, signals to isolate the primary or secondary containments do not operate any LPCS valves. This arrangement satisfies safety design basis 6. The check valve is the only core spray equipment in the primary containment required to actuate during a loss-of-coolant accident which requires consideration for the high temperature and humidity environment in the containment from the accident. The selected valve actuates on flow through the pipeline, independent of any external signal. Thus, neither the normal nor accident environment in the containment affects the operability of the core spray equipment for the accident. It is concluded that safety design basis 9 is satisfied. Taking the core spray water from the pressure suppression pool establishes a closed loop for recirculation of the spray water escaping from the break. It is concluded that safety design basis 11 is satisfied. The core spray spargers and piping are designed as Class I (see Appendix C) so that they meet design basis 10. 6.5.2.5 Low Pressure Coolant Injection System LPCI is provided to automatically reflood the reactor core in time to limit cladding temperatures after a nuclear system LOCA when the reactor vessel pressure is below the shutoff head of the pumps. The LPCI provides cooling by flooding which differs from the LPCS which provides cooling by spraying. For the limiting design basis accident, a combination of the LPCS and LPCI is necessary to supply sufficient cooling to meet peak clad temperature limits.30 Table 6.5-3 lists the Emergency Core Cooling Systems which are available for both recirculation suction and discharge breaks following an assumed single failure. The maximum vessel pressure against which the LPCI pumps must deliver some flow is determined by the required overlap with HPCI which has a low pressure cutoff for the HPCI turbine. The LPCIS pumps are capable of delivering flow above the pressure due to the pump head flow characteristics. LPCI cooling capability is analyzed using approved analytical methods based on the mass and energy flows to and from the reactor. The LPCI control system responds as described in Section 7.4.3. When the nuclear system pressure decreases to the pumping head of LPCI, the check valve in the injection line opens and LPCI water is pumped into the reactor vessel to reflood the core. 6.5-8
BFN-27 These actions provide an integral flow path for the injection of the LPCI flow into the bottom plenum of the reactor vessel. As the LPCI flow accumulates, the level rises inside the shroud. When the level reaches the top of the jet pumps, spillover occurs for a time raising the level outside the shroud. As the subcooled LPCI flow begins spilling into the region outside the shroud, the depressurization effect of the break is reduced since the subcooled water is now flowing out the break. As the pressure begins to rise, the LPCI flow is reduced until a quasi-equilibrium pressure is reached. At this point the break is partially covered by subcooled water which has spilled over the top of the jet pumps and the equivalent area of the break available for steam blowdown is maintained at the required equilibrium value by the LPCI spillage. If pressure were to rise, the LPCI flow would be reduced, the equivalent break size for steam blowdown would increase, and pressure would drop. Complete equilibrium will be reached when the rate of saturating the LPCI water becomes equal to the boil-off rate. It is noted that this condition will not actually be attained because of the HPCI and ADS effects on the transient. Although HPCI flow will be lost when pressure is reduced sufficiently, the auto depressurization valves would open as level continues to drop. To assure continuity of core cooling, signals to isolate the primary or secondary containments do not operate any LPCI valves when aligned for power operation. This arrangement satisfies safety design basis 6. The two check valves in Units 1, 2, and 3 are the only LPCI equipment in the primary containment required to actuate during a LOCA which require consideration for the high temperature and humidity environment in the containment from the accident. The type of valve chosen actuates on flow through the pipeline, independent of any external signal. Note: For Units 1, 2, and 3, Unit 3, the RHR check valve actuator, controls and indication functions have been deleted. The actuator is provided only for test. Thus, neither the normal nor accident environment in the containment affects the operability of the LPCI equipment for the accident. It is concluded that safety design basis 9 is satisfied. Using the pressure suppression pool as the source of water for LPCI establishes a closed loop for recirculation of LPCI water escaping from the break. It is concluded that safety design basis 11 is satisfied. 6.5-9
BFN-28 The LPCI and appropriate portions of the recirculation loops are designed as Class I (see Appendix C) so that they meet design basis 10. 6.5.3 Integrated Operation of the Emergency Core Cooling Systems Sections 6.4 and 6.5.2 described the performance and operation of each of the EECS individually. Section 6.5.3 is directed toward the integrated performance of the ECCS, i.e., how the entire ECCS operates together to provide core cooling for the entire spectrum of LOCA concurrent with an assumed active single failure. The ECCS integrated performance under LOCA conditions is evaluated according to the analysis procedure described in Section 6.5.2. The primary emphasis is placed on the liquid line break since for a given break size the consequences of a break in a liquid line are more severe than for a break in a steam line. As summarized in the next paragraphs, it is demonstrated that the core cooling capability is available even concurrently with the improbable loss of all offsite AC power. In actual cases, more core cooling capability than assumed in the analysis is available since the reliability of plant AC power is extremely high. The containment response is discussed in Section 14, "Plant Safety Analysis." 6.5.3.1 Recirculation Line Breaks Large Breaks The analyses assume a loss of normal auxiliary AC power concurrent with a line break. This satisfies safety design basis 5. The recirculation suction line is the largest liquid line inside the drywell. A double-ended break in this line would result in the most rapid depressurization and inventory loss. For the limiting case, the reactor is assumed to be operating at 102% of 3952 MWt. The analyses assume the reactor water level is at the normal level when the break occurs. Immediately after the break, critical flow would be established at the break. The large increase in core void fraction that would be caused by the decreasing vessel pressure would be sufficient to render the core subcritical. High drywell pressure would initiate mechanical scram of the control rod system in less than one second. A reactor vessel low-low-low water level signal initiates closure of the main steam isolation valves very rapidly. A double-ended break in the recirculation line will result in rapid depressurization of the reactor vessel and loss of reactor coolant inventory. Either the low-low-low 6.5-10
BFN-27 water level sensor in the reactor vessel or high drywell pressure sensor plus low reactor vessel pressure would generate a signal to initiate the ECCS (i.e., LPCS and LPCI mode of the RHR System). Intermediate or Small Breaks For break sizes below the lower limit of the unassisted LPCS or LPCI, the HPCI or ADS acts to depressurize the nuclear boiler, allowing either the LPCS or the LPCI to provide core cooling. The intermediate break range here is defined between the limits of the unassisted HPCI and the unassisted LPCS or LPCI, in which one of the high pressure systems (HPCI or ADS) works in conjunction with one of the low pressure systems (LPCS or LPCI) to provide core cooling. The integrated performance of the ADS, in conjunction with the LPCS or LPCS for the intermediate break sizes, is very similar to the performance of these same combinations for the small break sizes, except depressurization is more rapidly accomplished with increasing break size. For this reason, the integrated performance of these system combinations is discussed below for small breaks. For small breaks, the core is protected by the HPCI and by the LPCS or LPCI in conjunction with the ADS. The performance of the HPCI for these small break sizes is described in paragraph 6.5.2.2. Generally, the ADS/LPCS or ADS/LPCI combinations will not prevent partial core uncovery as the HPCI does. However, the ADS is designed to allow the LPCS or LPCI to prevent excessive cladding temperatures for all break sizes down to and including a zero break size. The sequence of events during the initial phase of the small break transient is similar to that discussed earlier for the HPCI System performance. The important difference in the system response for the ADS case is that the vessel pressure hovers near rated for a much longer period of time since the HPCI is assumed to be inoperable and the ADS is not actuated until 120 seconds after the low-low-low trip level is reached (and confirmation of ECCS system operation). An additional obvious difference is the more rapid loss of liquid inventory due to the higher vessel pressure and absence of HPCI makeup. When the ADS is actuated, a marked increase in the vessel pressure reduction rate occurs due to the critical steam flow through the main steam relief valves. This increased depressurization rate causes a corresponding increase in water levels inside and outside the shroud. When the water level inside the shroud drops to the top of the jet pumps, the water level is sustained except for flashing due to depressurization and boil-off. The level outside the shroud continues to drop rapidly until the LPCS or LPCI have restored the water level inside the shroud to the top of the jet pumps. At this level, the water from inside the shroud begins spilling over into the region outside the shroud, thereby raising the water level outside the shroud. 6.5-11
BFN-28 The performance cases illustrated above are typical for the ADS/LPCS or ADS/LPCI combinations in the small break region. As much larger breaks are considered, the performance of these system combinations becomes the same as the LPCS or LPCI alone discussed in paragraphs 6.5.2.4 and 6.5.2.5, respectively. Many different cases of recirculation line breaks in combination with various single failures were analyzed using nominal, Appendix K, and licensing basis assumptions 30. The AREVA LOCA break spectrum analysis for ATRIUM-10XM fuel is provided in Reference 30. The limiting break is determined to be a 0.20 ft2 split break in the recirculation discharge line with top-peaked axial power shape and battery failure affecting HPCI and LPCI. Using the Appendix K assumptions for the maximum recirculation suction line break coincident with battery failure results in a PCT well below the 2200°F limit. The licensing basis PCT is determined by summing the nominal PCT with an adder that is based on the nominal and Appendix K results. The analyses demonstrate the licensing basis PCT is well below the 2200°F limit 30. This satisfies safety design bases 1, 2, and 3. The cycle specific values are provided in Appendix N. It is also demonstrated30 that the maximum cladding oxidation and maximum hydrogen generation for all cases are well below the ECCS acceptance criteria discussed in paragraph 6.5.2.1. Typical operational sequences of the Emergency Core Cooling Systems are shown in Table 6.5-4 (for a recirculation discharge line break).30 6.5-12
BFN-28 6.5.3.2 Non-Recirculation Line Breaks Non-recirculation line breaks were analyzed.30 The non-recirculation breaks include: core spray line break, feedwater line break, and steamline breaks inside and outside the containment. The results show that these postulated non-recirculation line breaks are less limiting than the most severe break in the recirculation line. Discussion and illustration of the ECCS performance capability has purposely been directed toward the liquid breaks below the core. In general, the ECCS design criterion of limiting cladding temperature to less than 2200F is more easily satisfied for steam breaks than for liquid breaks because the reactor primary system depressurizes more rapidly with less mass loss for steam breaks than for liquid breaks. Thus the ECCS performance for a given break size improves with increasing break flow quality. The most severe steam pipe break would be one which occurs inside the drywell, upstream of the flow limiters. A break in this location would permit the pressure vessel to continue to depressurize to the drywell. As serious as this accident could be, it does not result in thermal-hydraulic consequences as severe as the rupture of a coolant recirculation pipe. The accident sequence would proceed as follows:
- a. An instantaneous guillotine severance of the steam pipe upstream of the steam flow restrictors would permit flow to accelerate to its limiting critical flow value in the break at the pressure vessel end, and at the flow-limiter end.
- b. The steam loss in excess of the generation rate would result in rapid depressurization of the pressure vessel and the steam pipes.
- c. The first 10 seconds of this accident are similar to the break outside the drywell discussed in Section 14, "Plant Safety Analysis."
However, for the break inside the drywell, closure of the isolation valves reduces the blowdown rate but does not prevent the vessel from depressurizing; the vessel would continue depressurizing through the pipe from the pressure vessel to the break. The reactor would be shut down immediately during a steamline break by the voids which would result from the depressurization. A mechanical scram would be initiated by a position switch in each isolation valve. Low water level and high drywell pressure which would occur later in the blowdown, would also initiate a scram. 6.5-13
BFN-27 The reactor coolant loss through blowdown consists of three intervals: first steam blowdown, then mixture blowdown, and finally steam blowdown again. After the accident, steam in each end of the break is accelerated to critical flow. As the reactor vessel depressurizes, the water level rises due to flashing. When the level reaches the steam pipes, the break flow changes from a steam blowdown to a steam-water mixture blowdown. Closure of the isolation valves reduces the blowdown rate. As coolant is expelled and the pressure decreases, the water level in the pressure vessel will drop below the steam pipe elevation and steam blowdown will begin again. The system depressurization rate would increase, but the mass flow rate through the break would decrease. 6.5.3.3 Effect of Fuel Cladding Failure on Core Cooling The ECCS is designed to prevent clad melting and fragmentation upon cooldown for any loss of coolant within the design basis spectrum. The purpose is to keep the cladding and fuel from distortion to a degree that subsequent cooling is ineffective. Satisfying these criteria does allow for the tolerance of cladding perforation. Even though the cooling equipment is successful in keeping cladding temperature well below the criterion for the design maximum temperature (2200F), a small percentage of the fuel may perforate. However, the occurrence of even a large number of perforations does not prevent core cooling from being successful. Cladding perforation occurs when the gas pressure within the rod exceeds the pressure the cladding can withstand for that particular cladding temperature. The mode of this failure is known. The perforation is local, in that a given fuel rod perforates at a particular location the extent of which is on the order of an inch in axial length. The perforation is random in that it may occur and will be localized at a weak point along the fuel rod length-probably at a point of cladding flaw, pellet oversize, pellet chip, or point of slightly increased cladding oxidation. The locations of such weak points are randomly distributed among the fuel rods within the fuel assembly. The conclusion that perforation is random and local is based upon confirmed observations of irradiated fuel.9 Such random failures have also been demonstrated in test loops by placing Zircaloy tubing, filled with UO2 pellets and pressurized with gas, in an induction heating facility and observing the failure mode. The observed failures were always localized and random along the length of the heated rod. Only a slight change in diameter was observed to occur, except within an inch or two on either side of the perforation. This is characteristic of burst failure at high temperatures in general. From these tests, the major conclusions are that perforations are indeed local and that the fuel rods do not distort grossly over the length of the fuel rod. 6.5-14
BFN-28 A separate study10 has been made of metal-water reactions during a loss-of-coolant accident and the report of the study shows how ECCS design is influenced by metal water reaction. In addition, this study covers the effects of hydrogen generation during a loss-of-coolant accident. The temperatures resulting from the postulated accident are determined using conservative models and conservative assumptions. Since steamline breaks are not as severe as liquid line breaks, it is concluded that safety design bases 2 and 3 are satisfied. 6.5.3.4 Alternate Operating Mode Considerations Operation with feedwater heaters out-of-service (FHOOS) or final feedwater temperature reduction (FFWTR) results in slightly more initial coolant mass which leads to improved cooling, a faster reflood time, and a slightly lower Peak Cladding Temperature (PCT) as confirmed by a FHOOS/FFWTR analysis of a typical BWR. The conclusions of the Reference 30 report are applicable for operation with a recirculation flow mismatch in accordance with Technical Specification Surveillance Requirement (SR) 3.4.1.1, FHOOS, FFWTR, and single-loop operation (SLO) 6.5.4 Emergency Core Cooling Systems Redundancy The design criterion of preventing peak cladding temperatures greater than 2200F is satisfied across the entire spectrum of possible liquid or steamline break sizes even in the event of the loss of normal auxiliary power combined with a single failure. It is concluded that the redundant capability of the ECCS is sufficient for all size line breaks up to and including the design basis break. It is assumed that one of the postulated single failures is a spurious accident signal initiated from a non-accident unit. This opposite unit false LOCA signal is assumed to occur either coincident with the valid LOCA signal prior to the valid LOCA signal or after the valid LOCA signal. There is sufficient redundancy in the ECCS initiation and preferred pump logic that the ECCS will be able to satisfy the design criteria for any postulated combinations of spurious and real accident signals. The individual functions of the ECCS also meet the design criteria over various ranges of break sizes in the nuclear system. Their integrated performance provides adequate and timely core cooling over the entire spectrum of LOCAs up to and including the design basis LOCA even with concurrent loss of offsite AC power. It is concluded that safety design basis 1 is satisfied. 6.5-15
BFN-28 6.5.5 Net Positive Suction Head (NPSH) 6.5.5.1 Description An important design attribute for the operation of the ECCS pumps is the assurance of adequate NPSH. NPSH is a function of individual pump flow, total ECCS flow, friction loss, ECCS strainer blockage, suppression pool water level, suppression pool temperature, and wetwell pressure. NPSH can be challenged for the ECCS pumps during events involving large increases in suppression pool temperature, which includes LOCA, Anticipated Transient Without Scram (FSAR Section 7.19), Station Blackout (FSAR Section 8.10), and NFPA 805 fire events (Fire Protection Plan). The HPCI is designed to provide reactor inventory makeup water and assist in depressurizing the reactor during an intermediate or small break LOCA. During the first 10 minutes after initiation of the accident, the suppression pool temperature remains below 140ºF. If the HPCI pump operates beyond the first 10 minutes following the event and the HPCI pump suction is aligned to the suppression pool, then manual operation would terminate HPCI pump operation when the suppression pool temperature reaches approximately 140°F due to system operating restrictions, which is below pool temperatures where NPSH margin would be of concern. This approach ensures that the HPCI pump NPSH remains adequate for operation following a postulated accident. The determination of the available NPSH (NPSHa) to the RHR and LPCS pumps as a function of time after a postulated DBA-LOCA short-term and DBA-LOCA long-term has been accomplished in accordance with the guidelines in Regulatory Guide (RG) 1.82, Revision 3, Water Sources for Long-term Recirculation Cooling Following a Loss-of-Coolant Accident. 6.5-16
BFN-28 6.5.5.2 NPSH Available (NPSHa) NPSHa is a function of the system design and operation and is determined from the following standard equation for pumps with flooded suctions: NPSHa = hatm + hstatic - hloss - h vapor where hatm = the head on the surface of the suppression pool due to the pressure in the wetwell atmosphere hstatic = the head due to the difference in elevation between the suppression pool surface and the centerline of the pump suction hloss = the head loss due to fluid friction, fittings in the flow path from the suppression pool to the pump, and the ECCS suction strainers which prevent ingestion of debris into the pumps hvapor = the head due to the vapor pressure of the suppression pool water at the suppression pool water temperature NPSHa, has been determined using specified design input parameters including suppression pool temperature, suppression pool water level, and ECCS pump flow rates 31. 6.5.5.3 Potential Plugging of Emergency Core Cooling System Suction Strainers NRC Bulletin 96-03 concerns the potential for inadequate NPSH for ECCS pumps resulting from accumulated debris on the ECCS suction strainers during the recirculation phase of a LOCA. As part of the resolution of this issue, high capacity passive stacked disk strainers are installed on the ring header. The strainer design is governed by the debris generation calculations in accordance with the Boiling Water Utility Resolution Guidance (BWROG) (URG) (GE NEDO-32686, R0, Dated November 1, 1996) for ECCS Suction Strainer Blockage. The primary insulation type on the drywell piping systems is reflective metal insulation (RMI) of both aluminum and stainless steel. The suction strainers provide adequate NPSH margin with a saturated thickness of foils of either aluminum or stainless steel insulation. The quantification of debris loading used in the strainer design is provided in GENE E12-00148-02, Debris Load Report for Sizing of Browns Ferry ECCS Pump Suction Strainers, July 1997. Debris captured by the strainer during a postulated LOCA will increase the pressure drop across the strainers. 6.5-17
BFN-28 6.5.5.4 NPSH Required (NPSHr) NPSHr is a function of the pump design. The pump vendor testing results for the BFN RHR and LPCS pumps was evaluated to determine values for NPSHr. The evaluation resulted in the generation of tables and charts that provide RHR and LPCS pump NPSHr as a function of the flow rate. 6.5.5.5 Analysis Methods The containment temperature and pressure conditions are determined using the M3CPT Code for the DBA-LOCA short-term and the SHEX Code for the DBA-LOCA long-term. Design inputs that maximize the suppression pool water temperature are utilized for the evaluation. The MULTIFLOW computer code is used to calculate the flow losses for the piping system configuration associated with each RHR and LPCS pump for the NPSH calculations. This code produces steady-state hydraulic flow results. Design inputs for the evaluations include decay heat, maximum pump flow rates, minimum suppression pool water level, maximum suppression pool temperature, containment spray operation, and operation of the drywell coolers. For the short-term analysis, the RHR and LPCS pumps are assumed to be at their maximum flow rate for the break location. For the long-term analysis, the RHR and LPCS pumps are throttled. The long-term analyses assume drywell and wetwell spray operation for the duration of the LOCA. The NPSHa is compared to the NPSHr for each of the pumps to demonstrate that adequate NPSH margin exists to ensure that the RHR and LPCS pumps perform their intended design safety functions. 6.5.5.6 Results Adequate NPSH is provided to ECCS pumps for all DBA-LOCA, accidents, abnormal operational transients and special events. The NPSH analysis does not take credit for any containment pressure greater than that assumed to exist at the start of the postulated events 31. 6.5-18
BFN-28 REFERENCES
- 1. Dresden 2 - Docket No. 50-237, Amendment 7, response to question 5-1.
- 2. Polomik, E. C., Levy, S. S., Sawochka, S. G. Film Boiling of Steam-Water Mixtures in Annular Flow at 800, 1100, and 1400 psi, ASME Paper No.
62-WA-136, November 30, 1962.
- 3. Effectiveness of Core Standby Cooling Systems for General Electric Boiling Water Reactors, General Electric Co., Atomic Power Equipment Division, APED-5458, March 1968.
- 4. Moody, F. J., "Maximum Flowrate of a Single Component Two-Phase Mixture,"
Journal
- 5. Moody, F. J., Liquid/Vapor Action in a Vessel During Blowdown, APED-5177, June 1966.
- 6. Zuber and Hench, Steady-State and Transient Void Fraction of Bubbling Systems and Their Operating Limits, Part I, Steady State Operation, General Electric Report 62GL 100 (1962).
- 7. Haberman and Morton, An Experimental Investigation of the Drag and Shape of Air Bubbles Rising in Various Liquids, Navy Dept., David Taylor Model Basic, Report 802, (September 1963).
- 8. Wilson, J. F., et al., The Velocity of Rising Steam in a Bubbling Two Phase Mixture, ANS Transactions, Vol. 5, No. 1, pg. 151 (1962).
- 9. Schraub, F. A., and Bock, R. G., Fuel Rod Failures During Simulated Loss-of-Coolant Conditions, APED-5479, March 1968.
- 10. Ianni, P. W., Metal Water Reactions--Effect on Core Cooling and Containment, General Electric Co., APED-5454, March 1968.
- 11. MCHFR (Minimum Critical Heat Flux Ratio) This is a fuel thermal limit parameter which was used in analyzing and operating the early core loadings in Browns Ferry. MCHFR has been replaced by a similar fuel thermal parameter called MCPR (Minimum Critical Power Ratio).
- 12. DELETED 6.5-19
BFN-28
- 13. EMF-2361(P)(A) Revision 0, EXEM BWR-2000 ECCS Evaluation Model, Framatome ANP, May 2001, as supplemented by the site-specific approval in NRC safety evaluation dated April 27, 2012.
- 14. DELETED
- 15. DELETED
- 16. DELETED
- 17. Calculation MDQ0999970046, NPSH Evaluation of Browns Ferry RHR and CS Pumps, Revision 12.
- 18. DELETED
- 19. DELETED
- 20. DELETED
- 21. DELETED
- 22. DELETED
- 23. DELETED
- 24. DELETED
- 25. DELETED
- 26. DELETED
- 27. DELETED
- 28. DELETED
- 29. NEDO-20566A, Analytical Model for Loss-of-Coolant Analysis in Accordance with 10 CFR Appendix K, General Electric Company, September 1986
- 30. ANP-3377P Revision 3, Browns Ferry Units 1, 2, and 3 LOCA Break Spectrum Analysis for ATRIUM 10XM Fuel (EPU), AREVA NP Inc., August 2015.
6.5-20
BFN-28
- 31. Calculation MDQ099920060011, Transient NPSH/Containment Pressure Evaluation of RHR and Core Spray Pumps, Revision 8.
- 32. NRC Letter, Browns Ferry Nuclear Plant, Units 1, 2, and 3 - Issuance of Amendments Regarding the Extended Power Uprate (EPU) (CAC Nos.
MF6741, MF6742, and MF6743), August 14, 2017. 6.5-21
BFN-28 TABLE 6.5-1 PLANT PARAMETERS USED LOCA ANALYSIS AREVA Analysis Plant Parameters Appendix K Core Thermal Power (MWt) 4031 Corresponding Power 102% of 3952 MWt Vessel Steam Output (lbm/hr) 16.85 X 106 Rated Core Flow (lbm/hr) 102.5 X 106* Vessel Steam Dome Pressure (psia) 1053 Maximum Recirculation Suction Line Break Area (ft2) 7.013 Maximum Recirculation Discharge Line Break Area (ft2) 7.013 Low Pressure Core Injection Valve Pressure Permissive for 335 Opening (psig) Core Spray Injection Valve Pressure Permissive for 335 Opening (psig) Recirculation Pump Discharge Valve Pressure Permissive 200 for Closure (psig) Maximum delay time from EDG start signal until bus at 13 rated voltage (seconds)
BFN-28 TABLE 6.5-2 ECCS EQUIPMENT CAPACITY ASSUMED IN LOCA ANALYSIS Function Number Installed Flow HPCI 1 3,600 gpm at 1174 psid, linearly increasing to 5,000 gpm at 1120 psid; 5,000 gpm between 1120 psid and 150 psid within 35 seconds. ADS Valves 6 800,000 lb/hr per valve at 1125 psig Core Spray System 2 (two pumps for each system) 5600 gpm at 105 psid per two CS pumps
- 5285 gpm at 105 psid per two CS pumps***
LPCI Pump 4 9400 gpm at 20 psid per pump on one loop; 17,300 gpm at 20 psid per two pumps on one loop ** NOTE: psid - pounds per square inch differential between reactor vessel and primary containment (torus)
- The FANP Analyses assume 5,435 gpm sprays inside of shroud after accounting for 40 gpm total leakage for LPCS as-built openings, core spray thermal sleeve-safe end connection and quarter inch high point vent hole in core spray vent hole. Also included are 125 gpm for leakages due to downcomer repairs.
- The FANP Analyses assume 8,640 gpm and 16,540 gpm (1 pump and 2 pumps in a loop) injected into the shroud after accounting for 600 gpm leakage principally around the jet pumps and 160 gpm access hole cover leakage.
- For core spray, the relaxed flow rate per loop is reduced by leakages associated with the 1/4-inch vent hole in the core spray T-joint (specified as 40 gpm) and repairs at the T-joint and the lower section replacement (specified as a total of 125 gpm). Flow rates are further reduced by 150 gpm to account for leakages associated with the Lower Sectional Hardware modification. This further reduction applies only to the configuration at Unit 3 and can be conservatively applied to Units 1 and 2.
- The FANP Analyses assume 8,640 gpm and 16,540 gpm (1 pump and 2 pumps in a loop) injected into the shroud after accounting for 600 gpm leakage principally around the jet pumps and 160 gpm access hole cover leakage.
BFN-26 TABLE 6.5-3 SINGLE FAILURE EVALUATION USED FOR LOCA ANALYSIS Remaining Systems*, Assumed Failure Recirculation Suction Recirculation Disposition Break Discharge Break Battery 6 ADS, 1 LPCS, 2 6 ADS, 1 LPCS Analyze SF-BATTlBA§ LPCI Battery 4 ADS**, HPCI, 1 4 ADS**, HPCI, 1 Analyze SF-BATTlBB§ LPCS, 2 LPCI LPCS Battery 4 ADS, HPCI, 1 LPCS, 4 ADS, HPCI, Bounded by SF-BATTlBB SF-BATTlBC§, 3 LPCI 1 LPCS, 1 LPCI SF-LOCA 6 ADS, HPCI, 1 LPCS, 6 ADS, HPCI, 1 Bounded by SF-BATTlBA (U1 & 2 only) 2 LPCI LPCS and SF-BATTlBB SF-LPCI 6 ADS, HPCI, 2 LPCS, 6 ADS, HPCI, 2 Bounded by SF-BATTlBA (Injection Valve) 2 LPCI LPCS and SF-BATTlBB SF-DGEN 6 ADS, HPCI, 1 LPCS, 6 ADS, HPCI, 1 Bounded by SF-BATTlBA (Diesel Generator) 2 LPCI LPCS and SF-BATTlBB SF-HPCI 6 ADS, 2 LPCS, 4 6 ADS, 2 LPCS, 2 Bounded by SF-BATTlBA LPCI LPCI SF-ADSlIL 4 ADS, HPCI, 2 4 ADS, HPCI, Bounded by SF-BATTlBB (Initiation Logic) LPCS,4 LPCI 2 LPCS, 2 LPCI SF-ADSlSV 5 ADS, HPCI, 2 LPCS, 5 ADS, HPCI, Bounded by SF-BATTlBB (Single Valve) 4 LPCI 2 LPCS, 2 LPCI
- Each LPCS means operation of two core spray pumps in a system. It is assumed that both pumps in a system must operate to take credit for core spray cooling or inventory makeup. Furthermore, 2 LPCI refers to two LPCI pumps into one loop, 3 LPCI refers to two LPCI pumps into one loop and one LPCI pump into one loop. 4 LPCI refers to four LPCI pumps into two loops, two per loop.
4 ADS, 5 ADS, and 6 ADS means the number of ADS valves available for automatic activation. Systems remaining, as identified in this table for recirculation suction line breaks, are applicable to other non-ECCS line breaks. For a LOCA from an ECCS line break, the systems remaining are those listed for recirculation suction breaks, less the ECCS in which the break is assumed.
§ SF-BATTlBX (X=A, B, or C) is the single failure of a Unit Battery that results in the loss of a downstream unitized 250V DC RMOV Board A, B, or C.
** A single failure of 250V RMOV Board B results in the loss of the normal power supply to ADS logic. Units 1, 2, and 3 are designed with an automatic transfer scheme for ADS logic power to provide four ADS valves for this board failure.
Unit 3 systems remaining. Conservative for Units 1 and 2. A special case of ADS failure involves failure of initiation logic. However, 4 ADS valves are single failure proof, and will still function automatically.
BFN-28 Table 6.5-4 SEQUENCE OF EVENTS FOR LIMITING RECIRCULATION DISCHARGE LINE BREAK BATTERY FAILURE (SF-BATTlBA) (EXEM BWR-2000 LOCA METHODOLOGY) (at 3952 MWt) Events Time (sec) Initiate break 0.0 Initiate scram 0.6 Low-low liquid level, L2 (448 in) 30.4 Low-low-low liquid level, L1 (372.5 in) 49.8 Jet pump uncovers 75.0 Recirculation suction uncovers 124.4 Lower plenum flashes 167.6 LPCS high-pressure cutoff 258.4 LPCS valve pressure permissive 248.0 LPCS valve starts to open 250.0 LPCS valve fully open 283.0 LPCS permissive for ADS timer 78.8 LPCS pump at rated speed 81.8 LPCS flow starts 258.4 RDIV pressure permissive 287.7 RDIV starts to close 289.7 RDIV fully closed 325.7 Rated LPCS flow 353.2 ADS valves open 171.8 Blowdown ends 353.2 Bypass reflood 486.9 Core reflood 423.1 PCT 423.1
BFN-26 TABLE 6.5-5 DELETED BY AMENDMENT 25
BFN-26 Table 6.5-6 DELETED BY AMENDMENT 26
BFN-26 Table 6.5-7 DELETED BY AMENDMENT 25
BFN-24 Figure 6.5-1 Deleted by Amendment 15
BFN-24 Figure 6.5-2 Deleted by Amendment 15
BFN-24 Figure 6.5-3 Deleted by Amendment 24
BFN-24 Figure 6.5-4 Deleted by Amendment 24
BFN-24 Figure 6.5-5 Deleted by Amendment 24
BFN-24 Figure 6.5-6 Deleted by Amendment 15
BFN-24 Figure 6.5-7 Deleted by Amendment 15
BFN-24 Figures 6.5-8 through 6.5-43 Deleted by Amendment 9
BFN-28 Figure 6.5-44 Deleted by Amendment 28
BFN-25 6.6 INSPECTION AND TESTING Each active component of the Emergency Core Cooling Systems provided to operate in a design basis accident is designed to be tested during normal operation of the nuclear system. The inboard isolation check valves can only be tested during cold shutdown (MODE 4 or 5). ASME Section XI, "Inservice Pump and Valve Test and System Pressure Test Requirements" are discussed in Subsection 4.12, "Inservice Inspection and Testing." The High Pressure Coolant Injection System (HPCIS), Automatic Depressurization, and Core Spray System have no normal process uses and, therefore, are tested periodically to provide assurance that the Emergency Core Cooling Systems will operate to effectively cool the reactor core in an accident. The four LPCI pumps may be placed in use as part of the Residual Heat Removal System during shutdown cooling, thus their status is known from normal process uses. However, the LPCI pumps are tested no less frequently than the rest of the ECCS. Other parts of the LPCI, such as the two check valves inside the primary containment drywell and the four shutoff valves outside the drywell, are also tested periodically. Note: For Units 1, 2, and 3, the RHR check valve actuator, controls and indication functions have been deleted. Preoperational tests of the Emergency Core Cooling Systems were conducted during the final stages of plant construction prior to initial startup. Testing of the HPCI turbine may be done using auxiliary boiler steam prior to nuclear system heatup (see Section 13, "Conduct of Operations"). These tests assured the proper functioning of all controls and instrumentation, pumps, piping, and valves. System reference characteristics such as pressure differentials and flow rates were documented during the preoperational tests and were used as the initial base points for measurements obtained in the subsequent operational tests. During plant operations, the pumps, valves, piping, instrumentation, wiring and, other components outside the primary containment can be visually inspected at any time. Components inside the primary containment can be inspected when the drywell is open for access. When the reactor vessel is open, for refueling or other purposes, the spargers and other internals can be inspected. The testing frequencies of most components, and its associated controls and instrumentation of the Emergency Core Cooling Systems are contained in the technical specifications and tested per approved site procedures. When the system is tested, the operation of most of the components are indicated in the control room. There are exceptions which require local observation at the component and may require special tests for which there are special provisions and methods. 6.6-1
BFN-25 Pressure operated relief valves may be removed for maintenance, bench checks, and setting adjustments during normal plant shutdown. Bench tests of the Automatic Depressurization valves are discussed in Subsection 4.4, "Nuclear System Pressure Relief System." A pressure operated control valve such as the one upstream of the HPCIS gland seal condenser can be functionally tested and adjusted in place. Flow operated check valves for reverse flow or excess flow are tested periodically in place by simulating the functional conditions. The proper position of manual valves for the accident mode is verified every 31 days by the technical specifications. Test lines are provided between pairs of containment isolation valves in the Emergency Core Cooling Systems to perform containment leak rate tests as specified in Table 5.2-2. Pumps for the Emergency Core Cooling Systems are equipped with face-type mechanical shaft seals. Normal leakage for these seals are 0 to 0.1 gph at operating conditions. The portions of the Emergency Core Cooling Systems requiring pressure integrity are designed to specifications for inservice inspection to detect defects which might affect the cooling performance. The reactor vessel nozzles, the core spray, and feedwater spargers receive particular attention. A design flow functional test of the HPCI, up to the normally closed pump discharge valve, is performed during normal plant operation by pumping water from the condensate supply header and back through the full flow test return line to the condensate tanks. The HPCIS turbine-pump is driven by steam from the reactor. The suction valves from the pressure suppression pool and discharge valves to the reactor feedwater line remain closed. HPCI test requirements are given in the Technical Specifications, Section 3.5.1. During a HPCIS flow test, if an initiation signal occurs, the system returns to the automatic startup mode; however, operator action may be required to adjust the flow demand setting. The HPCI may be tested at full flow with condensate at any time except when the reactor vessel water level is low, the drywell pressure is high, the condensate supply header level is low, or the valves from the pressure suppression pool to the pump are open. 6.6-2
BFN-25 To conduct the full flow test, the test return line valves to the condensate storage tank are opened. The turbine steam valves are open. The pump delivers bypass flow to the pressure suppression pool to provide minimum flow protection. To test the pump discharge valve, it is operated with the remote control switch, observing the valve position lights. The containment isolation check valve must be tested during cold shutdown (MODE 4 or 5). If the differential pressure is too high, equalization of the pressure across the valve may be accomplished by pressurizing upstream of the valve using the test connections. Each loop of the Core Spray System may be tested during reactor operation. The test conditions are tabulated in the Technical Specifications, Section 3.5.1. The normal system test does not inject cold water into the reactor because the inboard injection check valve is held closed by the reactor pressure which is higher than core spray pump pressure, and the inboard injection valve remains closed. To test the core spray pumps at rated flow, the pump suction valves from the pressure suppression pool remain open, the pumps are started using the remote manual switches in the control room, then the test bypass valve to the pressure suppression pool is opened. Proper operation is determined by observing the instruments in the control room. If an initiation signal occurs during the test, the Core Spray System is signaled to start and the system returns to the automatic startup mode. The two motor-operated injection valves are tested by cycling and observing the position indicator lights. The inboard injection check valve inside the drywell is tested during periods of cold shutdown (MODE 4). Similarly, LPCI pumps and valves are tested periodically during reactor operations. With the injection valves closed and the return line open to the pressure suppression pool, full flow pumping capability is demonstrated. The injection valves and the check valves are operated, as described previously for the core spray valves. Controls and instrumentation are tested as described in Subsection 7.4, "Emergency Core Cooling Systems Control and Instrumentation." Upon receipt of an LPCI initiation signal during tests, the valves in the test bypass lines are closed automatically to assure that the LPCI pump discharge is routed properly to the reactor vessel. 6.6-3
BFN-25 The ECCS piping and components in the Reactor Building are monitored by routine inspections, general housekeeping practices, and system operability testing which maintains system leakage to an as-low-as-possible level. It is concluded that safety design basis 7 is satisfied. 6.6-4}}