ML19331E211
| ML19331E211 | |
| Person / Time | |
|---|---|
| Site: | Columbia  |
| Issue date: | 07/31/1980 |
| From: | WASHINGTON PUBLIC POWER SUPPLY SYSTEM |
| To: | |
| Shared Package | |
| ML17275A572 | List: |
| References | |
| XXXX, NUDOCS 8009090322 | |
| Download: ML19331E211 (400) | |
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AMENDMENT NO. 10 July 1980
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1 AMENDMENT NO. 10 July 1980 lh ( !
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WNP-2 AMENDMENT NO. 10 July 1980 m 7.3 ENGINEERED SAFETY FEATURE SYSTEMS 7.
3.1 DESCRIPTION
Section 7.3 describes the instrumentation and controls of t.he following plant Engineered Safety Features (ESP) systems:
- a. Emergency Core Cooling Systems (ECCS)
- b. Primary Containment and Reactor Vessel Isolation Control Systems (PCRV1CS)
- c. Main Steam Line Isolation Valve - Leakage Control System (MSLC)
- d. RHRS-Containment Spray Cooling Mode (RCSCM)
- e. RHRS-Suppression Pool Cooling Mode (RSPCM)
- f. Standby Service Water System (SSW)
- g. Main Control Room and Critical Switchgear Rooms HVAC System O'(_j h. Reactor Building Ventilation and Pressure Control System
- i. Standby Gas Treatment System (SGTS)
- j. Containment Instrument Air System (CIA)
- k. Containment Atmosphere Control System (CAC)
The sources which supply power to the engineared safety feature systems originate from on-site AC and/or DC safety-related buses or, as in the case of the PCRVICS failsafe logic, from the non-safety-related RPS M-G sets. Refer to Chapter 8 for a complete discussion of the ESF systems power sources. 7.3.1.1 System Description 7.3.1.1.1 Emergency Core Cooling Systems (ECCS) - Instrumentation and Controls
.The emergency core cooling systems are a network of the following subsystems. See 6.3.1 and <6.3.2.
- a. High pressure core spray system (HPCS);
O-Ns 7.3-1
WNP-2 AMENDMENT NO. 10 July 1980 O
- b. Automatic depressurization system (ADS);
- c. Low pressure core spray system (LPCS);
- d. Low pressure coolant injection (LPCI) mode of the residual heat removal system (RHRS).
The following plant variablet are monitored and provide auto- l matic initiation of the ECCS when these variables exceed pre-determined limits:
- 1. Reactor Vessel Water Level A low water level in the reactor vessel could indicate that reactor coolant is being lost through a breach in the reactor coolant pressure boundary and that the core is in danger of becoming overheated as the reactor coolant inventory diminishes. Refer to Figure 7.3-9 (Nuclear Boiler P&ID) for a schematic arrangement of reactor vessel instrumentation.
- 2. Drywell Pressure High pressure in the drywell could indicate a breach of the reactor coolant pressure boundary inside the drywell and that the core is in danger of becoming overheated as reactor coolant inventory diminishes.
7.3.1.1.1.1 High Pressure Core Spray (HPCS) System - Instrumentation and Controls
- a. HPCS Function The purpose of the HPCS is to provide high pressure reactor vessel core spray for small line breaks which do not depressurize the reactor vessel. In addition, HPCS is redun-dant to the RCIC system for mitigation of the consequences of various events listed in Appendix 15A. Refer also to 6.3.2.2.1.
- b. HPCS Operation Schematic Arrangements of system mechanical equipment is shown in Figure 7.3-7 (HPCS P&ID). HPCS system component control logic is shown in Figure 7.3-8 (HPCS FCD) and Figure 7.3-4 (HPCS Power Supply FCD). Instrument specifications are listed in Tables 7.3-1 and 7.3-2. Plant Layout drawings and Electrical Schematics are identified in 1.7. Operator Information Displays are shown in Figure 7.3-7 (HPCS P&ID) and Figure 7.3-8 (HPCS and HPCS Power Supply FCD).
7.3-2
WNP-2 AMENDMENT NO. 10' July 1980 / t i Q/ The.HPCS is initiated automatically by either reactor vessel low water level (Trip Level'2) or drywell high pressure. The system is' designed to operate automatically for at least 10 minutes without any actions required by the control room operator. Once initiated the HPCS logic seals-in and can be reset by the cperator only when the initial conditions return to normal. Refer.to Figure 7.3-8 for a schematic represen-tation c the HPCS System Initiation logic. Reactor vessel water level (Trip Level 2) is monitored by four redundant differential pressure switches. The switch contacts are arranged'in a one-out-of-two twice logic arrangement to assure that no single event can prevent the initiation of the HPCS. Initiation diversity is provided by drywell pressure which is monito' red by four redundant pressure' switches. The switches are electrically connected in a one-out-of-two twice logic arrangement to assure that no single instrument failure can prevent the initiation of the HPCS. The HPCS components respond to an automatic initiation signal as follows (actions are simultaneous unless stated otherwise):
) 1. The HPCS Diesel Generator is signalled to start and its protrs".ive relays are bypassed. Once the Diesel is Utc 'ed it signals the start of its cooling water pump. See 6.3.1.1.8.2.
- 2. The HPCS pump motor is signalled to start.
- 3. The normally open pump suction from the conden-sate storage tank valve M0 F001, is signalled to open.
- 4. The test return valves M0 F010, M0 F011 and M0 F023 are signalled closed.
- 5. The HPCS injection valve M0 F004 is signalled to open.
The HPCS pump discharge flow and pressure are monitored by pressure switches. If pump discharge pressure is normal but discharge flow-is low enough that pump overheating may occur the minimum flow return line valve M0 F012 is signalled open. The valve is automatically closed if flow is normal. If the water level in the condensate storage tanks falls below ('m a' predetermined level, che suppression pool suction valve
\, MO-F015 automatically opens. When M0 F015 is fully open the condensate storage tank suction valve M0 F001 automatically 7.3-3
WNP-2 AMENDMENT NO. 10 July 1980 O closes. Two level switches are used to detect low water level in each of the condensate storage tanks. Either switch can cause automatic suction transfer. The suppression pool suc-tion valve also automatically opens if high water level is detected in the suppression pool. Two level switches monitor suppression pool water level and either switch can initiate opening of the suppression pool suction valve. To prevent losing suction to the pump, the suction valves are interlocked so that one suction path must be open before the other closes. The HPCS provides makeup water to the reactor until the vessel water level reaches the high level trip (Trip Level 8) at which time the injection valve M0F004 is automatically closed. The pump will continue to run on minimum flow recirculation. The injection valve will automatically reopen if vessel level again drops to the low level (Trip Level 2) initiation point. The HPCS pump motor and injection valve are provided with manual override controls. These controls permit the reactor operator to manually control the system following automatic initiation. 7.3.1.1.1.2 Automatic Depressurization System (ADS)- Instrumentation and Controls
- a. ADS System Function The automatic depressurization system is designed to provide automatic depressurization of the reactor vessel by activating seven safety / relief valves. These valves vent steam to the suppression pool in the event that the HPCS cannot maintain the reactor water level following a LOCA. ADS reduces the reactor pressure so that flow from the low pressure ECCS, LPCI system and LPCS, can inject into the reactor vessel in time to cool the core and limit fuel cladding temperature. Refer also to 6.3.2.2.2.
- b. ADS Operation Schematic arrangements of system mechanical equipment is shown in Figure 7.3-9 (Nuclear Boiler P&ID). ADS component control logic is shown in Figure 7.3-10 (Nuclear Boiler FCD).
Instrumentation specifications are listed in Tables 7.3-3 and 7.3-4. Plant Layout Drawings and Electrical Schematics are identified in 1.7. Operator Information Displays are shown in Figure 7.3-9 (Nuclear Boiler P&ID) and Figure 7.3-10 (Nuclear Boiler FCD). O 7.3-4
WNP-2 AMENDMENT NO. 10 July 1980 (~h.
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To prevent. inadvertent actuation ofLthe-ADS two channels of logic for'each ADS trip system (A & B) are used. Both chan-nels must be activated to actuate an ADS trip system. Refer to. Figure _7.3-7 for a schematic representation of the ADS ini-tiation logic. Each' channel-contains a single input from a drywell high pressure sensor. In addition, one channel includes two dif-ferential pressure sensor inputs monitoring reactor vessel low water level (Trip Level 3 and Trip Level 1). The second low
~
water level trip (Trip Level 3) provides confirmation of a reactor vessel low water level condition. The other channel, in addition to drywell high pressure, includes a single reac-tor vessel low water level (Trip Level 1) input. To assure that adequate makeup water is available after the vessel has been depressurized each logic channel includes a
-pump discharge pressure permissive signal indicating LPCI or LPCS system available for vessel water makeup. Any one of the three LPCI pumps or the LPCS pump is sufficient to permit automatic depressurization.
After receipt of the initiation signals and after a delay pro-vided by timers, each of the two solenoid pilot air valves are O energized. This allows pneumatic pressure from the accumula-tor.to act on the air cylinder operator. Each ADS trip system timer can be reset manually to delay system initiation. If reactor vessel water level is restored by HPCS prior to the r end of the time delay, ADS initiation will be prevented. The ADS trip' system A actuates the "A" solenoid pilot valve on each ADS relief valve. Similarly, the ADS trip system B actuates the "B" solenoid pilot valve on each ADS relief valve. Actuation of either solenoid pilot valve causes the ADS valve to open to provide depressurization. Once initiated the ADS logic seals-in and can be reset by the control room operator only when either drywell pressure or vessel water level return to normal. Two control switches (one for_each trip system solenoid) are located in the main control room for each safety / relief valve associated with the ADS. Each switch controls one of the two solenoid pilot valves. t O 7.3-5 i . - - - - - .-
WNP-2 AMENDMENT NO. 10 July 1980 0 7.3.1.1.1.3 Low Pressure Core Spray (LPCS) - Instrumentation and Controls
- a. LPCS Function The purpose of the LPCS is to provide low pressure reactor vessel core spray following a loss-of-coolant accident when the vessel has been depressurized and vessel water level has not been restored by the HPCS. The LPCS is functionally diverse to the LPCI mode of the residual heat removal system.
See 6.3.2.2.3.
- b. LPCS Operaticn Schematic Arrangements of system mechanical equipment is shown in Figure 7.3-11 (LPCS P&ID). LPCS component control logic is shown in Figure 7.3-12 (LPCS FCD). Instrument specifications are listed in Tables 7.3-5 and 7.3-C. Plant Layout Drawings and Electrical Schematics are identif.ied in 1.7. Operator Information Displays are shown in Figure 7.3-11 (LPCS P&ID) and Figure 7.3-12 (LPCS PCD).
The LPCS is initiated automatically by either rea( lor vessel low water level and/or drywell high pressure. The system is designed to operate automatically for at least 10 minutes without any actions required by the control room operator. Once initiated the LPCS logic seals-in and can be reset by the control room operator only when the initial conditions return to normal. Refer to Figure 7.3-12 for a schematic represen-tation of the LPCS system initiation logic. l Reactor vessel water level (Trip Level 1) is monitored by two redundant differential pressure switches. To provide diver-sity drywell pressure is monitored by two redundant pressure switches. The vessel level switch contacts and the drywell pressure switch contacts are connected in a one-out-of-two twice loaic arrangement so that no single instrument failure can prevent initiation of LPCS. The LPCS components respond to an automatic initiation signal simultaneously (or sequentially as noted) as follows: l
- 1. The Division 1 Diesel Generator is signalled to start,
- 2. The normally closed test return line to the suppression pool valve M0 F012 is signalled closed,
- 3. When power (offsite or onsite) is available O at the LPCS pump motor bus the LPCS pump is signalled to start, 7.3-6
c WNP-2 AMENDMENT lNO'. 10 July.1980 [V i
~4. Reactor p'ressure is monitored by a differen-
.tial pressure switch which senses; the-pressure-. difference (vessel.to LPCS) across-
!' the LPCS injection valve M0 F005. When the dif ferential pressure is low'enough to pro-tect the'LPCS from overpressure and power 1is availablef to; the pump motor bus, . the injec ' tion. valve-is signalled to'open. 7: The LPCS pump discharge flow-is. monitored by a differential pressure' switch. When-the pump is running'and discharge flow
'is. low enough~that pump overheating may occur, the' minimum.
flow-return line valve M0 F011 is opened. The valve is auto- ,. matically closed if1 flow is normal. The LPCS. pump suction from~the suppression pool valve M0 F001 is normally open, the control switch is keylocked in the open
- - position, and thus requires'no automatic open signal for l system. initiation.
- The.LPCS pump and injection valve are provided with manual
- override controls. These controls permit the operator to 3 manually control the' system subsequent to automatic initiation.
() 7.3.1.1.1.4 RHRS -~ Low Pressure Coolant-Injection (LPCI) Mode Instrumentation and Controls Y j a. LPCI Function i Low pressure coolant injection (LPCI) is an operating mode of the residual heat removal system (RHRS). .The purpose of the LPCI system is to provide low pressure reactor vessel coolant makeup following a loss-of-coolant accident when the vessel has beenLdepressurized and vessel water level is not. restored by the HPCS. :See 6.3.2.2.4.
- b. LPCI' Operation 4
1
- Schematic' Arrangements of. system mechanical equipment is shown
- r. in~ Figure'7.3-13-(RHR P&ID). LPCI component control logic is-
[ Lshown'in> Figure 7.3-14 (RHR~FCD). Instrument specifications
- are listed in Tables 7.3-7.and'7.3-8. Plant Layout Drawings and' Electrical Schematics.are identified in 1.'s. Operator lInformation Displays are~shown in Figure 7.3-13 (RHR P&ID) and
- Figure;7.3-14T(RHR-FCD).
4The LPCIJsystem iscinitiated automatically by either reactor vessel. low water ' level and/or: by~ drywell high pressure. - The
~ system is l designed. to operate Lautomatically for at least 10 minutes 7without any; actions required by'the control room I
w operator. . 'Once initiated ' the :LPCI logic seals-in and . can be W ii 7.3-7
-m., ~. c, , . - - + .- n u, - .%,. , , , . , m ., .-.-,y . - . - -
ym.,,-y,,,y ., _,, -,,cwe,,~y,e<,m-
WNP-2 AMENDMENT NO. 10 July 1980 9 reset by the control room operator only when initial con-ditions return to normal. Refer to Figures 7.3-13 and 7.3-14 for a schematic representation of the LPCI A and the LPCS B/C initiation logic, respectively. Reactor vessel water level (Trip Level 1) is monitored by two redundant differential pressure switches. To provide diver-sity drywell pressure is monitored by two redundant pressure switcher. To initiate the Division 2 LPCI (Loops B and C) the vessel level switen contacts and the two drywell pressure switch con-tacts are connected in a one-out-of-two-twice arrangement so that no single instrument failure can prevent initiation of LPCI. The Division 1 LPCI (Loop A) receives its initiation signal from the LPCS logic. The LPCI system components respond to an automatic initiation signal simultaneously (or sequentially as noted) as follows (the Loop A components are controlled from the Division 1 logic; the Loop B and C components are controlled.from the Division 2 logic): #
- 1. The Division 2 diesel generator is signalled to start from the Loop B and C initiation logic.
- 2. If normal auxiliary (offsite) power is available at the pump motor busses the LPCI Loop A, B, and C pumps are signalled to start. If offsite power is not uvailable and the diesel generators are providing power to the pump motor busses sequential loading of the diesel generators is required. This is accomplished by delaying the start of the LPCI pumps A aad B by 5 seconds while allowing the LPCS and LPCI C pumps to start immediately.
- 3. Reactor pressure is monitored by differential pressure switches which senses the pressure difference (vessel to LPCI system) across each LPCI injection valve M0 F042 A, B, C.
When the differential is low enough to pro- l tect the LPCI from overpressure and power is available at the associated pump motor bus, the injection valve is signalled to open, 7.3-8
WNP-2 AMENDMENT NO. 10 July 1980
\-' 4. The following normally closed valves are signalled closed to ensure proper system lineup:
a) The RHR heat exchanger discharge to RCIC valves MO F026 A, B, and AO F065 AB, b) The RHR heat exchanger flush to suppression pool valves MO F011 A, B, c) The RHR heat exchanger steam pressure reducing valves AO F051 A, B, d) The RHR heat exchanger steam inlet isola-tion valves MO F052 A, B and MO F087 A, B, e) The test return line to the suppression pool valves MO F024 A, B and MO F021, f) The suppression pool spray valves MO F027 A, B.
- 5. The normally open heat exchanger bypass
/~h valves MO F048 A, B are signaled open. The kl % open signal is automatically removed 10 min-utes after system initiation to allow operator control of the valve for throttling purposes.
Each LPCI pump discharge flow is monitored by a differential pressure switch which, when the pump is running and following an 8-second time delay, opens the minimum flow return line valve MO F064 A, B, C if flow is low enough that pump overheating may occur. The valve is automatically closed.if flow is normal. The 8-second time delay is provided to pre-vent reactor vessel inventory loss during the shutdown cooling mode of the RHRS (see 5.4.7.2.6(a)). The three RHR pump suction from the suppression pool valves MO F004 A, B, C and the RHR heat exchanger inlet and outlet valves MO F047 A, B and MO F003 A, B have their control switches keylocked in the open position, and thus require no automatic open signal for system initiation. The two series service water crosstie valves MO F093 and MO F094 have their control switches keylocked in the close position, and thus require no automatic close signal for system initiation. O l 7.3-9
~
WNP-2 AMENDMENT NO. 10 July 1980 The two series containment spray valves MO F016 A, B and MO 0 F017 A, B, the two series RHR heat exchanger vent valves MO F073 A, B and F074 A, B and the RHR shutdown cooling mode suc-tion valves M0 F006 A, B are all normally closed and thus require no automatic close signal for system initiation. The LPCI pump motors and injection valves are provided with manual override controls. These controls permit the operator to manually control the system subsequent to automatic initiation. 7.3.1.1.2 Primary Containment and Reactor Vessel Isolation Control System (PCRVICS) - Instrumentation and Controls
- a. PCRVICS Function The PCRVICS includes the instrument channels, trip logics and actuation circuits that automatically initiate valve closure providing isolation of the primary containment and/or reactor vessel, and initiation of systems provided to limit the release of readioactive materials.
See 6.2.4 and Table 6.2-16 for a complete description of pri-mary containment and reactor vessel process lines and isola-tion signals applied to each.
- b. PCRVICS Operation Schematic mechanical arrangements of containment isolation valves and other components initiated by PCRVICS are shown in Figures 7.3-13 (RHR P&ID), 7.3-9 (Nuclear Boiler P&ID), 3.2 '1 (RWCU P&ID), 3.2-3 (Reactor REURC P&ID), 11.2-2 (Equip. Drain Flow Diag.), 11.23 (Floor Drain Flow Diag.) and 3.2-16 (SGTS Flow Diag.). PCRVICS component control logic is shown in Figure 7.3-10 (Nuclear Boiler FCD), 7.3-14 (RHR FCD) and 7.3-1 (RWCU FCD). Instrument specification are listed in Tables 7.3-5 and 7.3-7. Plant Layout drawings and Electrical Schematics are identified in 1.7. Operator Information Displays are shown in Pigure 7.3-10 (Nuclear Boiler FCD).
Refer also to Figure 7.3-9 (Nuclear Boiler P&ID). During normal plant operation, the isolation control system sensors and trip logic that are essential to safety are energized. When abnormal conditions are sensed, instrument contacts open and de-energize the trip logic and thereby ini-tiate isolation. Once initiated the PCRVICS trip logics seal-in and may be reset by the operator only when the initial conditions return to normal. 7.3-10
WNP-2 AMENDMENT NO. 10 July 1980 .O Nj , The.PCRVICS trip logic provides isolation signals to the Main Steam Line Isolation Valves AO F022 A, B, C, D and AO F028 A, B, C, D; to the Main Steam Line Drain Valves MO F016, F067 A, B, C, D, and MO F019; to the Reactor Water Sample Valves.MO F019 and F020; to the RHR Shutdown Cooling System Valves MO F008, F009, F023, F040, F049, F053 A, B, F099 A, B; to the RHR Sample Line Valves Solenoid Operated Valve (SOV) F060 A, B and F075 A, B; to the Reactor Water Cleanup System Valves MO F001 and F004; to the Drywell Equipment Drain Valves AO F019 and F020; to the Drywell Floor Drain Valves AO F003 and F004; and to isolate the TIP system valves. Each main steam line isolation valve (MSIV) has two control solenoids. Each solenoid receives inputs from two redundant
.ogics. A signal from either can deenergize the solenoid.
For any one valve to close automatically, both of its sole-noids must be de-energized. The main steam line isolation valve logic has a minimum of four redundant instrument channels for each measured variable. One channel of each variable is connected to one trip logic. One group of redundant logics (A, C) is used to control one solenoid of both inboard and outboard valves of all four main O steam lines, and the other group of redundant logics (B, D) is used to control the other solenoid of both inboard and out-board valves. The four PCRVICS trip logics are arranged in a one-out-of-two twice logic combination (Trip Logic A or C and B or D). Refer to Figure 7.3-2. The main steam line drain valves, drywell equipment and floor drain valves, reactor water sample valves, the reactor water cleanup system, and residual heat removal system isolation valves also' operate in pairs. The inboard valves close if both of the Division 1 isolation logics (A and B) are tripped, and the outboard valves close if both of the Division 2 logics (C and D) are tripped. Refer to Figure 7.3-3. The PCRVICS also provides signals to start the standby gas treatment system; to remove nonessential loads from essential busses; and to isolate the reactor building ventilation system, and the primary containment purge and vent system. The following variables provide inputs to the PCRVICS logics for initiation of reactor vessel and drywell isolation, as well as the initiation or trip of other plant functions when predetermined limits are exceeded. Combinations of these variables, as necessary, provide initiation of various iso-lating and' initiating functions as described in Table 6.2-16 and below: 7.3-11
WNP-2 AMENDMENT NO. 10 July 1980
- 1. Reactor Vessel Low Water Level 0
A low water level in the reactor vessel could indicate that reactor coolant is being lost through a breach in the reactor coolant precsure boundary and that the core is in danger of becoming overheated as the reactor coolant inventory diminishes. Reactor vessel low water level initiates closure of various valves. The closure of these valves is intended to isolate a breach of the pipelines, conserve reactor coolant by closing of f process lines, and limit the escape of radioactive materials from the primary containment through process lines that communicate with the primary coolant boundary or primary containment. Two reactor vessel low water level isolation trip settings are used to complete the isolation of the primary containment and the reactor vessel. The first (and higher) reactor vessel low water level isolation trip (Trip Level 3) initiates closure of all RHR system isolation valves. The main steam lines are lef t open to allow the removal of heat from the reactor core. The second (and lower) reactor vessel low water level isola-tion trip (Trip Level 2) completes the isolation of the pri-mary containment and reactor vessel by initiating closure of all other isolation valves and also provides inputs to logic which trips or initiates other plant equipment. Reactor vessel water level is monitored by four redundant dif-ferential pressure switches. Each provides a low water level input to one of the four PCRVICS trip }ogics. Diversity of trip initiation for pipe breaks inside of primary containment is provided by drywell high pressure.
- 2. Drywell High Pressure High pressure in the drywell could indicate a breach of the reactor coolant pressure boundary inside the drywell and that the core is in danger of becoming overheated as reactor coolant inventory diminishes.
Drywell pressure is monitored by four redundant pressure switches. Each switch provides an input to one of the four trip logics.
- 3. Main Steam Line-High Radiation The main steam line radiation monitoring senses the gross release of fission products from the fuel ano initiates 7.3-12
-WNP-2 AMENDMENT NO. 10 July 1920 /'^'si t
V appropriate actions to limit fuel damage and contain the released. fission products. Four redundant detectors monitor the gross gamma radiation from the main steam lines. Each provides an input to one of the four_PCRVICS trip logics. Each monitoring channel consists of a gamma-sensitive ion chamber and a log radiation monitor. Each log radiation moni-tor has three trip circuits. One upscale trip circuit is used to initiate scram (see 7.2.1) and containment isolation, and alarm. The second circuit is used for an alarm and is set at a level below that of the upscale trip circuit used for scram and isolation. The third circuit is a-downscale or inopera-tive trip that actuates an instrument trouble alarm in the control room and produces an isolation and scram trip signal. When the main steam line radiation level exceeds a predeter-mined value, PCRVICS initiates closure of all main steam line isolation valves, main steam line drain valves and reactor water sample valves. The offgas system mechanical vacuum pump is tripped, and the mechanical vacuum pump lines are isolated.
<- 4. Main Steam Line-Tunnel and Pipe Routing in
( ,w) Turbine Building High Ambient Temperature and Differential Temperature High ambient temperature in the tunnel and pipe routing areas in the turbine building in.which the main steam lines are located outside of the primary containment could indicate a leak in a main steam line. Such a lsak may also be indicated by high differential temperature between the outlet and inlet ventilation air for these areas. The automatic closure of valves prevent the excessive loss of reactor coolant and the release of a significant amount of radioactive material from the reactor coolant pressure boundary. Four redundant main steam line high ambient temperature sen-sors are provided in the main steam tunnel and four in the steam line area of the turbine building. Four redundant dif-ferential' temperature sensors monitor the outlet and inlet ventilation ~ air ducts of the main steam line tunnel. Each
-main steam line trip isolation logic is de-energized by high ambient temperature in the main steam tunnel on the turbine building and by high differential temperature in the tunnel inlet / outlet ventilation air.
When an increase in main steam line tunnel ambient or dif-ferential temperature is detected, trip signals initiate clo- '(f-~) m : sure of all main steam line isolation and drain valves. 7.3-13
WNP-2 AMENDMENT NO. 10 July 1980 Diversity of trip initiation signals for main steam line tun-nel ambient temperature and high differential temperature is provided by main stean. line high flow, and steam line low pressure instrumentation.
- 5. Main Steam Line-High Flow Main steam line high flow could indicate a breach in a main steam line. Automatic closure of isolation valves prevents excessive loss of reactor coolant and release of significant amounts of radioactive material from the reactor coolant pressure boundary.
Sixteen redundant differential pressure switches, four for each main steam line, monitor the main steam line flow. Four differential pressure switches for each main steam line pro-vide inputs to each of the four trip logics. When a significant increase in main steam line flow is detected, trip signals initiate closure of all main steam line isolation and drain valves.
- 6. Main Turbine Inlet - Low Steam Pressure Low steam pressure at the turbine inlet while the reactor is operating could indicate a malfunction of the nuclear system pressure regulator in which the turbine governor valves or turbine bypass valves become fully open, and causes rapid depressurization of the reactor vessel. From reduced power, the rate of decrease of nuclear system saturation temperature could exceed the allowable rate of change of vessel tempera-ture. A rapid depressurization of the reactor vessel while the reactor is near full power could result in undesirable differential pressures across the channels (around some fuel bundles) of sufficient magnitude to cause mechanical defor-mation of channel walls. Such depressurizations, without ade-quate preventive action, could require thorough vessel analysis or core inspection prior to returning the reactor to power operation.
Four redundant pressure sensors, one for each main steam line, monitor main steam line pressure and each provides an input to one of the four trip logics. When a significant decrease in main steam line pressure is detected, the PCRVICS initiates closure of all main steam line isolation and drain valves. The main steam line low pressure trip is bypassed by the reac- l tor mode switch in the Shutdown, Refuel, and Startup modes of l reactor operation. In the Run mode, the low pressure trip ' function is operative. l l l 7.3-14
WNP-2 AMENDMENT NO. 10 July 1980
] '~'
- 7. Reactor Building Ventilation Radiation Monitor -
Instrumentation and Controls The reactor building ventilation monitoring consists of four sensor and trip units. Each channel has two trips. The upscale trip indicates high radiation and the downscale trip indicates instrument trouble. The reactor building ventilation radiation monitor senses reactor building exhaust to the elevated release point. In the event that radiation levels exceed predetermined limits the intake and exhaust dampers are closed.
- 8. Reactor Water Cleanup (RWCU) System-High Differential Flow High differential flow in the reactor water cleanup system could indicate a breach of the reactor coolant pressure boundary of the cleanup system. The flow at the inlet to the system (suction from Recirc. lines) is compared with the flow at the outlets of the system (flow return to feedwater or flow to the main condenser and/or radwaste).
7s Two redundant differential flow sensors compare the reactor
\') water cleanup system inlet-outlet flow. Each of the flow monitoring sensors provides an input to one of the two (inboard or outboard) logic trip channels.
When an increase in reactor water cleanup system differential flow is detected, the PCRVICS initiates closure of all reactor water cleanup system isolation valves. Diversity of trip initiation signals for reactor water cleanup system line break is provided by instrumentation for reactor water level, differential flow, and ambient or differential temperature in RWCU equipment areas. The reactor water cleanup system high differential flow trip is bypassed by an automatic timing circuit during normal reac-tor water cleanup system surges. This time delay bypass pre-vents inadvertent system isolations during system operational changes.
- 9. Reactor Water Cleanup (RWCU) System-Area High Ambient Temperature and Differential Temperature High temperature in the equipment room areas of the reactor s water cleanup system could indicate a breach in the reactor
/ coolant pressure boundary in the cleanup system.
K.s) 7.3-15
WNP-2 AMENDMENT NO. 10 July 1980 Six redundant ambient temperature and six differential tem-0 perature sensors monitor the reactor water cleanup system area temperatures. Three redundant ambient and three redundant differential temperature circuits are associated with each of the two trip logics. Six redundant ambient temperature ele-ments are located in the followin.g locations: Pump Room, Filter /Demin. Room, and Heat Exchanger Room. Six pairs of tem-perature elements are located in the ventilation supply and exhaust areas of the above locations. When a significant increase in reactor water cleanup system area ambient or differential temperature is detected the PCRVICS initiates closure of all reactor water cleanup system isolation valves. The output trip signal of each sensor initiates a logic trip and closure of either the inboard or outboard reactor water cleanup system isolation valve. Diversity of trip initiation signals for high differential temperature is provided by two pair of differential tem-perature elements and associated differential temperature switches (DTS) for each reactor water cleanup system area. Each pair of temperature elements and its differential tem-perature switch are associated with one of two logic channels.
- 10. RHR System-Area High Ambient Temperature and Differential Temperature High temperature in the equipment room areas of the RHR system could indicate a breach in the reactor coolant pressure boundary in the RHR system.
Four redundant ambient temperature and four redundant dif-ferential temperature sensors monitor the RHR system area temperatures. Two ambient and two dif'erential temperature sensors are associated with one trip logic. The remaining temperature channels are associated with the other trip logic. The ambient temperature elements are located in each RHR equipment area. Four pairs of temperature elements are located in the ventilation supply and ventilation exhaust of each RHR equipment area. When an increase in RHR system area ambient temperature or differential temperature is detected the PCRVICS initiates closure of all RHR system isolation valves. The output trip signal of each sensor initiates a trip logic and closure of either the inboard or outboard RHR system iso-lation valve. 7.3-16
WNP-2 AMENDMENT NO. 10 July 1980 m (b Diversity of trip initiation signals - for RHR line break is provided by ambient tetperature, differential temperature, and cooling flow instrumentation. An increase in space tempera-ture, differential temperature, or flow will initiate RHR system isolation.
- 11. RHR System - Flow Rate Monitoring High flow in the RHR system suction line from the reactor vessel could indicate a breach in the reactor coolant pressure boundary in the RHR system.
Two redundant differential pressure switches, one for each trip logic, monitor the RHR shutdown cooling mode suction line. The output trip signal of each sensor initiates a logic trip and closure of either the inboard or outboard RHR system iso-lation valve.
- 12. Main Condenser Vacuum Trip The main turbine condenser low vacuum signal could indicate a leak in the condenser. Initiation of automatic closure of
, various valves will prevent excessive loss of reactor coolant 4
and the release of significant amounts of radioactive material. Four redundant vacuum switches monitor the main condenser
. vacuum. Each switch provides an input to one of the four trip logics.
When a signficant decrease in main condenser vacuum is detected, the PCRVICS initiates closure of all main steam line isolation and drain valves. Main condenser low vacuum trip can be bypassed manually when the turbine stop valve is less than 90% open. 7.3.1.1.3 Main Steamline Isolation Valve Leakage Control System (MSLCS) - Instrumentation and Control
- a. MSLCS Function The MSLCS is designed to minimize the release of fission pro-ducts which could bypass the standby gas treatment system after the postulated LOCA. This is accomplished by directing the leakage through the closed main steamline isolation valves to bleed lines which pass the leakage flow into an area served by the standby gas treatment system. See 6.7.
V(i 7.3-17 l
WNP-2 AMENDMENT NO. 10 July 1980 O
- b. MSLCS Operation Schematic Arrananments of system mechanical equipment is shown in Figure 3.2-25 (MSLC Flow Diag.). MSLC system component control logic is shown in Figure 7.3-15 (MSLC Control Logie Diag.). Instrument specifications are listed in Tables 7.3-27 and 7.3-28. Plant Layout drawings and Electrical Schematics are identified in 1.7. Operator Information Displays are shown in Figure 3.2-25 (MSLC Flow Diag.) and Figure 7.3-15 (MSLC Control Logic Diag.).
The MSLCS is manually actuated after a LOCA has occurred, pro-vide 6 that the reactor and steamline pressures are below the pressure permissive setpoints and the inboard MSIVs are fully closed. The outboard and inboard subsystems are provided with one remote manual initiating switch each. When the inboard system is initiated, the exhaust blower FN-1 is actuated and the bleed valves V-2A,B,C,D and V-3A,B,C,D and the bypass valves V-1A,B,C,D are opened, heaters are actuated and timers are initiated. If the steamline pressure is greater than 5 psig after one minute, the bleed valves will close. If the pressure is not excessive, the bleed valves will remain open. After another minute, the bypass valve is closed. The flow is thus routed through the flow element. Within the next 30 seconds, a third timer allows flow to be monitored and the bleed valves to be closed, if necessary, for high flow. When the outboard system is initiated depressurization valves V-9 and V-10 are opened and the exhaust blower FN-2 is activated. When the steam lines have depressurized to approximately atmo-spheric pressure, the depressurization branch valves V-4 and V-5 are closed and flow is diverted to the blower suction. 7.3.1.1.4 RHRS-Containment Spray Cooling Mode (RCSCM) - Instrumentation and Controls
- a. Containment Spray Cooling Mode Function The containment spray cooling mode is an operating mode of the Residual Heat Removal System. It is designed to condense steam in the suppression chamber air volume and/or the drywell atmosphere following a LOCA. See 6.5.2.
- b. Containment Spray Cooling Mode Operation Schematic Arrangements of system mechanical equipment is shown in Figure 7.3-13 (RilR P&ID). RHR system component control 7.3-18
WNP-2 AMENDMENT NO. 10 July 1980 V logic is shown in Figure 7.3-14 (RHR FCD). Instrument speci-fications are listed un Tables 7.3-25 and 7.3-26. Plant Layout drawinga.and Electrical Schematics are identified in 1.7. Operator Informaulon Displays are shown in Figure 7.3-13 (RHR P&ID) and Figut'e 7.3-14 (RHR-FCD) The containment spray cooling mode i.s initiated by the control room operator by diverting LPCI flow to either the suppression pool or the drywell by opening valves MO F027A,B or MO F016A,B, and MO F017A,B. The following conditions must exist before the operator can initiate a containment spray cooling loop:
- 1. The LOCA signal which automatically initiated LPCI must still exist,
- 2. Drywell high pressure is monitored by two redundant pressure switches. One of the ,
two switches must indicate high pressure,
- 3. The operator must close the LPCI injection valve MO F042 A, B.
7.3.1.1.5 RHRS-Suppression Pool Cooling Mode (RSPCM) Instrumentation and Controls
- c. SPCM Function The suppression pool cooling mode is an operating mode of the residual heat removal system. It is designed to prevent suppression pool temperature from exceeding predetermined limits following a reactor slowdown of the ADS or safety /
relief valves.
- d. SPCM Operation Schematic Arrangements of system mechanical equipment is shown in Figure 7.3-13 (RHR P&ID). Component control logic is shown in Figure 7.3-14 (RHR FCD). Instrument specifications are listed in Tables 7.3-23 and 7.3-24. Plant layout Drawings and Electrical Schematics are identified in 1.7. Operator Information Displays are shown in Figure 7.3-13 (RHR P&ID) and
. Figure 7.3-14 (RHR FCD).
The. suppression pool cooling mode is initiated by the Control Room operator either during normal plant operation or following a LOCA, when the suppression pool temperature moni-(y toring system (see 7.6) indicates that pool temperature may ( ,) exceed a predetermined limit. 7.3-19
WNP-2 AMENDMENT NO. 10 July 1980 0 During normal plant operation the operator initiates the SPCM as follows:
- 1. The RHR Pump (A or B) is started. The standby service water pump is started and the RHR heat exchanger service water discharge valve M0 F068 A, B is opened automatically when the RHR pump starts.
- 2. The RHR test return line valve M0 F024 A, B is opened.
- 3. The RHR heat exchanger inlet and outlet valves M0 F047 A, B and M0 F003A, B are keylocked open. The heat exchanger bypass valve M0 F048 A, B and valve M0 F024 A, B are throttled as necessary.
Subsequent to a LOCA the operator initiates the SPCM as follows:
- 1. Once reactor vessel water level has been restored, the LPCI flow must be terminated by closing the LPCI injection valve M0 F042 A, B. Closing the injection valve causes the LOCA initiation logic to be overridden and h allows operator control of the system.
- 2. The RHR test return line valve M0 F024 A, B control logic also has LOCA signal override provisions. This allows the operator to open the valve.
- 3. The RHR heat exchanger inlet and outlet valves M0 F047 A, B and M0 F003 A, B are keylocked open. The heat exchanger bypass valve M0 F041 A, R, after a time delay (a ten minute timer keeps this valve open following a LOCA) and valve M0 F024 A, B are throttled as necessary.
7.3.1.1.6 Standby Service Water (SSW) System - Instrumentation and Controls
- a. SSW Function The standby service water system provides cooling water to the diesel generators, the RHR heat exchangers, the HPCS, RCIC, LPCI, and LPCS auxiliary equipment (room cooler, pump cooler) and the essential HVAC chillers. See 9.1.
7.3-20
WNP-2 AMENDMENT NO. 10 July 1980 . ,rq V
- b. SSW System Operation Schematic Arrangements of system mechanical equipment is-shown in Figure 9.1-4-(SSW Flow Diag.). SSW component control logic is shown in Figure 7.3-17 (SSW Control Logic Diag.). Instrument specifications are listed in Tables 7.3-21 and 7.3-22. Plant Layout Drawings and Electrical Schematics are identified in 1.7. Operator.Information Displays are shown in Figure 9.1-4 (SSW' Flow Diag.) and Figure 7.3-17 (SSW Control Logic Diag.).
The SSW system is automatically initiated as follows:
- 1. The Division 1 SSW pump P-1A is started auto-matically if water level in the spray pond is suf ficient and when either the RHR 0 pump, the LPCS pump, or the Division 1 diesel generator is started,
- 2. The Division 2 SSW pump P-1B is started auto-matically if water level in the spray pond is sufficient and when either the RHR B pump, RHR C pump, the. Division 2 diesel generator, or the RCIC pump is started,
- 3. The HPCS service water pump C002 is automati-cally started when the HPCS pump is started.
Once the service water pumps are started the following occurs:
- 1. The RHR heat exchanger service water discharge valves MO F068A,B are signalled open,
- 2. After the SSW pumps discharge pressure exceeds a minimum value the pump discharge valves V-2A,2B and V-29 are signalled to open, The SSW pumps are automatically tripped if spray pond water level becomes too low.
7.3.1.1.7 Main Control Room and Critical Switchgear Rooms HVAC System - Instrumentation and Controls Schematic Arrangements of system mechanical equipment is shown in Figure 3.2-19 (HVAC Flow Diag.). Component control logic is shown in Figure 7.3-14 (HVAC Control Logic Diag.). Instrument specifications are listed in Tables 7.3-19 and 7.3-20. Plant
->g Layout Drawings and Electrical Schematics are identified in k ,) 1.7. Operator Information Displays are shown in Figure 3.2-19 7.3-21
WNP-2 AMENDMENT NO. 10 July 1980 O (HVAC Flow Diag.) and Figure 7.3-14 (HVAC Control Logic Diagram). For a complete description of the Main Control Room and Critical Switchgear Rooms HVAC Instrumentation and Controls refer to 9.4.1. 7.3.1.1.8 Containment Atmosphere Control (CAC) System - Instrumentation and Controls Schematic Arrangements of system mechanical equipment is shown in Figure 3.2-17 (CAC Flow Diag.). CAC component control ogic is shown in Figure 7.3-21 (CAC Control Logic Diag.). nstrument specifications are listed in Tables 7.3-17 and
.3-18. Plant Layout Drawings and Electrical Schematics are .dentified in 1.7. Operator Information Displays are shown in Figure 3.2-17 (CAC Flow Diag.) and Figure 7.3-21 (CAC Control Logic Diag.).
For a complete description of the CAC System Instrumentation and Controls refer to 6.2.5. 7.3.1.1.9 Standby Gas Treatment System (SGTS) - Instrumentation and Controls For a complete descrigtion of the SGTS Instrumentation and Controls refer to 6.5.1. Schematic Arrangements of system mechanical equipment is shown in Figure 3.2-16 (SGTS Flow Diag.). SGTS component control logic is shown in Figure 7.3-19 (SGTS Control Logic Diag.). Instrument specitications are listed in Tables 7.3-15 and 7.3-16. Plant Layout Drawings and Electrical Schematics are identified in 1.7. Operator Information Displays are shown in Figure 3.2-16 (SGTS Flow Diag.) and Figure 7.3-19 (SGTS Control Logic Diag.). 7.3.1.1.10 Reactor Building Ventilation and Pressure Control System - Instrumentation and Controls
- a. System Function The reactor building ventilation and pressure control system automatically maintains subatmospheric pressure of 1/4" water gage in the reactor building atmosphere. See 9.4.2.
- b. System Operation Schematic Arrangements of system mechanical equipment is shown in Figure 3.2-18 (HVAC React. Bldg. Flow Diag.). System com- h ponent control logic is shown in Figure 7.3-19 (SGTS Control 7.3-22
WNP-2 AMENDMENT NO. 10 July 1980 Logic Diag.). Instrument specifications are liste'd in Tables 7.3-15 and 7.3-16. Plant Layout Drawings and Electrical Schematics are identified in 1.7. Operator Information Dis-plays-are shown in Figure 3.2-18 (HVAC React. Bldg. Flow Diag.) and Figure 7.3-19 (SGTS Control Logic Diag.). The differential pressure is monitored by eight redundant dif-forential prassure transmitters, four in Division 1 and four in Division 2, wnich measure the differential pressure from the exterior of four sides of the reactor building to the fuel pool area. " The signal indicating the least differential pressure from the four differential pressure transmitters in one division is selected and is used to control the position of a damper of that division in the normal reactor building exhaust fan units or upon the initiation of the standby gas treatment system by containment isolation signals high drywell pressure, low reactor water level, or reactor building exhaust high radiation, the reactor building ventilation and. pressure control system then controls reactor building pressure by controlling the standby gas treatment system fan units. (See 6.5.1). 7.3.1.1.11 Containment Instrument Air (CIA) System (} a. CIA System Function The purpose of the containment instrument air system is to provide uninterruptable divisional instrument air to essential ADS valve accumulators inside primary containment and non-safety-related instrument air supplies to other valves inside containment as shown in Figure 3.2-21. The system consists of a safety-relate' portion, which is compriced of two divisions, and a non-safety-related portion. During normal operation, l the non-safety-related portion of the system provides control
. air to ADS accumulators and other valves.
In the event of failure of the non-safety-related portions of the system, which is indicated by low header pressure and detected by two redundant pressure switches, the safety-related portion automatically maintains header pressure from two nitrogen bottle sources. The non-safety-related portion l of the system is isolated from the safety-related portion upon detection of failure of the non-safety-related portion. See 9.3.1.2.2.
- b. CIA System Operation Schematic Arrangements of system mechanical equipment is shown in Figure 3.2-21 (CIA Flow Diag.). CIA component control
[s^)$ logic is'shown in Figure.7.3-20 (CIA Control Logic Diag.). Instrument' specifications are listed in Tables 7.3-13 and 7.3-23
WNP-2 AMENDMENT NO. 10 July 1980 0 7.3-14. Plant Layout Drawings and Electrical Schematics are identified in 1.7. Operator Information Displays are shown in Figure 3.2-21 (CIA Flow Diag.) and Figure 7.3-20 (CIA Control Logic Diag.). The containment instrument air system is always in operation. The instrumentation and controls of the system perform the following functions:
- 1. Monitor CIA system header pressure
- 2. Monitor CIA system compressor operation
- 3. Isolate the non-safety-related portion of system in the event of failure in this por-tion
- 4. Maintain CIA system header pressure in the event of item 3 by sequentially opening nitrogen bottles.
7.3.1.2 Design Basis The ESP systems are designed to provide timely protection against the onset and consequences of conditions that threaten the integrity of the fuel barrier and the reactor coolant pressure boundary. Chapter 15, " Accident Analysis," iden-tifies and evaluates events that jeopardize the fuel barrier and reactor coolant pressure boundary. The methods.of assessing barrier damage and radioactive material releases, along with the methods by which abnormal events are identified, are presented in that chapter.
- a. Variables Monitored to Provide Protective Action The following variables are monitored in order to provide protective actions to the ESP systems:
- 1. HPCS a) Reactor Vessel Low Water Level (Trip Level 2) b) Drywell High Pressure
- 2. ADS a) Reactor Vessel Low Water Level (Trip Level 3) 7.3-24
'WNP-2 AMENDMENT NO. 10 July 1980 'O l b) Reactor Vessel Low Water Level (Trip i
- Level-1) 4 c) Drywell High Pressure-
- 3. LPCS.and LPCI a) Reactor Vessel Low Water Level (Trip j Level 1) i b) Drywell High Pressure
- 4. PCRVICS a) Reactor Vessel Low Water Level (Trip Level 3) b) Reactor Vessel Low Water Level (Trip Level 2) d c) Main Steam Line High Radiation I d) Main Steam Line Area High Ambient and
, Differential Temperature f
I e) Main Steam Line High Flow I f) Turbine Inlet Low Steam Pressure g) Elevated Release Point High Radiation 4 h) RWCU High Differential Flow ' i) RWCU Area High Ambient' Temperature and Differential Temperature j) RHR Area High Ambient Temperature and Differential Temperature k) RHR Suction High Flow ' ! - 1) Main Condenser Low Vacuum
- 5. MSLCS a) Reactor. Vessel Low Pressure
- 6. RCSCM
- a) Drywell High Pressure 7.3-25 l
l
. , , - - - - - .~g --.,l-,n- ,----4 ,-, A, ,,,,,--.,n._,, , -- - , , _ , . - - , -
n e-- -
WNP-2 AMENDMENT 'w. 10 July 1980 0
- 7. RSPCM a) Suppression Pool Temperature b) Drywell High Pressure c) Reactor Vessel Low Water Level (Trip Level 1)
- 8. SSWS a) RHR, LPCS, RCIC, or Diesel Generator Start
- 9. Main Control Room and Critical Swgr. Room HVAC a) Duct Chlorine High Concentration b) Remote Air Intake High Radiation c) High Room Temperature
- 10. Reactor Bldg. Ventilation and Pressure Control a) Reactor Bldg. to Fuel Pool Area Differential Pressure
- 11. SGTS a) Reactor Vessel Low Water Level (Trip Level 2) b) Drywell High Pressure c) Reactor Building Ventilation High Radiation
- 12. CIAS a) Instrument Air Header Low Pressure
- 13. CACS a) Reactor Vessel Low Water Level (Trip Level 2) b) Drywell High Pressure The plant conditions which require protective action involving the ESP systems are described in Chapter 15 and Appendix 15A. h 7.3-26
WNP-2 AMENDMENT NO. 10 July 1980 [J
~N
- b. Location and Minimum Number of Sensors See' Chapter 16 for the minimum number of sensors required to monitor safety-related variables. There are no sensors in the ESP systems which have a spatial dependence.
- c. Prudent Operational Limits Operational limits for each safety-related variable trip setting are selected with sufficient margin so that a spurious ESF system iniation is avoided. It is then verified by an-alysis that the release of radioactive materials, following postulated gross failures of the fuel or the nuclear system process barrier, is kept within acceptable bounds.
- d. Margin The margin between operational limits and the limiting con-ditions of operation of ESF systems are listed and the bases stated in Chapter 16, " Technical Specifications".
- e. Levels s
Levels requiring protective action are established in Chapter
) 16, " Technical Specifications".
- f. Range of Transient, Steady State, and Envirenmental Conditions Refer to Tables 3.11-1 through 3.11-5 and 3.1.2.1.4.1 for environmental conditions. Refer to 8.2.1 and 8.3.1 for the maximum and minimum range of energy supply to ESP instrumen-tation and controls. All ESP instrumentation and controls are specified and purchased to withstand the effects of energ; supply extremes.
- g. Malfunctions, Accidents, an6 Other Unusual Events Which Could Cause Damage to Safety System Chapter 15, " Accident Analysis" describes the following cre-dible accidents and events; floods, storms, tornados, earthquakes, fires, LOCA, pipe break outside containment.
Each of these events is discussed below for the ESF systems.
- 1. Floods The buildings containing ESF systems components have been designed to meet the.PMF (Probable Maximum Flood) at the site This ensures that the buildings will remain water-()
-s location.
tight under FMF conditions including wind-generated wave 7.3-27
WNP-2 AMENDMENT NO. 10 July 1980 O action and wave runup. For a discussion of internal flooding protection refer to 3.4.1.4.1.2, 3.4.1.5.2, and 3.6.
- 2. Storms and Tornados The buildings containing ESF systems components have been designed to withstand meteorological events described in 3.3.
- 3. Earthquakes The structures containing ESF systems components have been seismically qualified as described in 3.7 and 3.8, and will remain functional during and following a safe shutdown earth-quake (SSE). Seismic qualification of instrumentation and electrical equipment is discussed in 3.10.
- 4. Fires To protect the ESF systema in the event of a postulated fire, the redundant portions of the systems are separated by fire barriers. If a fire were to occur within one of the sections or in the area of one of the panels, the ESP systems functions would not be prevented by the fire. The use of separation and fire barriers ensures that even though some portion of the systems may be affected, the ESF systems will continue to pro-vide the required protective action. A fire detection system using heat detectors and product of combustion detectors 10 provided in PGCC floor sections and in panels containing RSP systems components mounted on these floor sections. A Halon fire suppression system is provided in the same areas.
- 5. LOCA The ESF systems components located inside the drywell and functionally required during and/or following a LOCA have been environmentally qualified to remain functional as discussed in 3.11 and indicated in Table 3.11-1.
6 Pipe Break Outside Secondary Containment This condition will not affect the ESF systems. Refer to 3.6.
- 7. Missiles Protection for safety-related components is described in 3.5.
- h. Minimum Performance Requirements l
~ Minimum performance requirements for ESF instrumentation and controls are provided in Chapter 16, " Technical Specifications". 7.3-28
WNP-2 AMENDMENT NO. 10 July 1980 (D V 7.3.1.3 . Final System Drawings The final system drawings including:
- 1. Process and Instrumentation Diagrams (P&ID)/ Flow Diagrams
- 2. Functional Control Diagrams (FCD) Control Logic Diagrams have been~ provided for the ESF systums in this section.
ESF systems electrical interconnection and scheniatic diagrams are in 1.7. Functional and architectural design difference between the PSAR and FSAR are listed in Table 1.3-8. 7.3.2 ANALYSIS 7.3.2.1 ESP Systems - Instrumentation and Controls Chapter 15, " Accident Analysis," and Chapter 6, " Engineered Safety Feature Systems," evaluate the individual and combined O capabilities of the ESP systems. The ESF systems are designed such that a loss of instrument air, a plant load rejection, or a turbine trip will not pre-vent the completion of the safety function. 7.3.2.1.1 Conformance to 10 CFR 50 Appendix A The following is a discussion of conformance to those General Design Criteria which apply specifically to the ESP systems. Refer to 7.1.2.2 for a discussion of General Design Criteria which apply equally to all safety-related systems.
-a. Criterion 33 See 7.3.1.1.1 (HPCS).
- b. Criterion 34 See 7.3.1.1.6 (SSW).
- c. Criterion 35 See 7.3.1 (ECCS) and 7.3.1.1.6 (SSW).
s_- . 7.3-29 _ , y . , ._, , _ v-. --- - - .
WNP-2 AMENDMENT NO. 10 July 1980 9
- d. Criterion 36, 37, 39, 40, 42, 43, 45, 46, and 61 - Fluid Systems See 7.3.2, Regulatory Guide 1.22.
- e. Criterion 38 See 7.3.1.1.4 (RCSCM), 7.3.1.1.5 (SPCM) and 7.3.1.1.6 (SSW)
- f. Criterion 41 See 7.3.1.1.11 (CAC), and 7.3.1.1.9 (SGTS).
- g. Criterton 44 See 7.3.1.1.6 (SSW).
- h. Criterion 60 and 61 See 7.3.1.1.9 (SGTS) and 7.3.1.1.3 (MSLCS).
- i. Criterion 64 See 7.3.1.1.2 (PCRVICS) and 7.3.1.1.8 (Reactor Bldg. Vent. Radiation).
7.3.2.1.2 Conformance to IERR Standards The following is a discussion of conformance to those IEEE standards which apply specifically to the ESF systems. Refer to 7.1.2.3 for a discussion of IEEE standards which apply equally to all safety-related systems. IEEE 279-1971 Criteria for Protection Systems for Nuclear Power Generating Stations:
- 1. General Functional Requirement (IEEE 279-1971, Paragraph 4.1)
The ESF systems automatically initiate the appropriate protec-tive actions, whenever the parameters described in 7.3.1.2.A reach predetermined limits, with precision and reliability assuming the full range of conditions and performance discussed in 7.3.1.2.
- 2. Single Failure Criterion (IEEE 297-1971, Paragraph 4.2)
ESF systems are not required to meet single failure criteria on an individual system (division) basis. However, on a net-7.3-30 ,
WNP-2 AMENDMENT NO. 10 July 1980 (a
\-
' work basis, the single failure criteria does apply to assure the completion of a protective function. Redundant sensors, wiring,' logic,.and actuated devices are physically and electrically separated such that a single failuru will.not prevent,the protective function. Refer to 8.3.1.4 for a complete description of the WNP-2 separation criteria.
- 3. Quality Components (IEEE-279-1971, Paragraph 4.3)
For a discussion of the quality of ESF system components and modules refer to 3.11.
- 4. Equipment Qualification (IEEE 279-1971, Paragraph 4.4)
Vendor certification requires that the sensors associated with each of the RPS trip variables, manual switches, and trip logic components perform in accordance with the requirements listed on the purchase specification as well as in the intended application. This certification, in conjunction with the existing field experience with these components in this application, will serve to qualify these components. Qualification tests of the relay panels are conducted to con-O' firm their adequacy for this service. In-situ operational testing of these sensors, channels, and the entire protection system will be performed during the preoperational test phase. For a complete discussion of RPS Equipment Qualification refer to 3.5, 3.6, 3.10, and 3.11.
- 5. Channel Integrity (IEEE 279-1971, Paragraph 4.5)
For a discussion of ESF systems channel integrity under all extremes of conditions described in 7.3.1.2 refer to 3.10, 3.11, 8.2.1, and 8.3.1.
- 6. Channel Independence (IEEE 279-19'il, Paragraph 4.6)
ESF systems channel independence is maintained through the application of-the WNP-2 separation criteria as described in 8.3.1.4. i
- 7. Control and Protection Interaction (IEEE 279-1971, Paragraph 4.7)
There are no ESF system and control system interactions. 7.3-31 i l
WNP-2 AMENDMENT NO. 10 July 1980 9
- 8. Derivation of System Inputs (IEEE 279-1971, Paragraph 4.8)
The ESP variables are direct measures of the desired variables requiring protective actions. Refer to 7.3.1.1.1 thru 7.3.1.1.11.
- 9. Capability of Sensor Checks (IEEE 279-1971, Paragraph 4.9)
Refer to 7.3.2.1.3, Regulatory Guide 1.22.
- 10. Capability for Test and Calibration (IEEE I
279-1971, Paragraph 4.10) Refer to 7.3.2.1.3, Regulatory Guide 1.22.
- 11. Channel Bypass or Removal from Operation (IEEE 279-1971, Paragraph 4.11)
During periodic tests of any one ESP system channel, a sensor may be valved out of service and returned to service under the administrative control procedures. Since only one sensor is valved out of service at any given time during the test interval, protective action capability for ESF system automa-tic initiation is maintained through the remaining redundant instrument channels.
- 12. Operating Bypasses (IEEE 279-1971, Paragraph 4.12)
The ESF systems contain the following operating bypasses. The PCRVICS has two bypasses. 1) main steam line low pressure operating bypass which is imposed by means of the mode switch. In all modes except run, the mode switch cannot be left in this position above 10% of rated power without inititating a scram. Therefore the bypass is removed by the normal reactor operating sequence, and 2) the low condenser vacuum bypass which is imposed by means of a manual bypass switch in con-6 junction with closure of the turbine stop valves, the reactor mode switch in any position other than "RUN", and reactor pressure below the low pressure setpoint. Bypass removal is accomplished automatically by the opening of the turbine stop valves or raising reactor pressure above the interlock pressure setpoint and manually by placing the bypass switch in normal position or by placing the mode switch in the "RUN" position. 7.3-32
WNP-2 AMENDMENT NO. 10
, . July 1980
(-< 4
\-s 13. Indication of Bypasses (IEEE 279-1971,
. Paragraph 4.13)
For a discussion of bypass and inoperability indication refer to 7.1.2.4, Regulatory Guide 1.47.
- 14. Access to Means for Bypassing (IEEE 279-1971, Paragraph 4.14)
Access to means of bypassing any safety action or function for the ESF systems.is under the administrative control of the i control room operator. The operator is alerted to bypasses as described in 7.1.2.4, Regulatory Guide 1.47. Control switches which allow safety system bypasses are keylocked. All heylock switches in the control room are designed such that their key can only be removed when the switch is in the " accident" or " safe" position. All keys will normally be removed from their respective switches during operation and maintained under the control of the shift supervisor. Further, the key locker will be audited once per day by the shift supervisor. Should a key be required to change a valve position, it will be obtained from the shift supervisor via approved key control proceduras. eO 15. Multiple Trip Settings (IEEE 279-1971, Paragraph 4.15) There are no multiple cet points within the ESF systems. , 16. Completion of Protective Action Once Initiated (IEEE 279-1971, Paragraph 4.16) . Each of the automatically initiated ESF system control logics seal-in electrically and remain energized after initial con-ditions return to normal. Deliberate operator action is required to return (reset) an ESF system logic to normal.
- 17. Manual Initiation (IEEE 279-1971, Paragraph 4.17)
Refer to the discussion of Regulatory Guide 1.62 in 7.3.2.1.3. 18._ Access to Setpoint Adjustments (IEEE 279-1971, Paragraph 4.18) All access to'ESF_ system set point adjustments, calibration controls, Land test points are under the administrative control 1 ye~s 'of the control room operator. b [ 7.3-33 l
WNP-2 AMENDMENT NO. 10 July 1980
- 19. Identification of Protective Actions (IEEE 279-1971, Paragraph 4.19)
ESP protective actions are directly indicated and identified by annunciators located in the main control room and a typed record is available from the process computer.
- 20. Information Readout (IEEE 279-1971, Paragraph 4.20)
The ESP systems are designed to provide the operator with accurate and timely information pertinent to their status. They do not introduce signals that could cause anomalous indi-cations confusing to the operator.
- 21. System Repair (IEEE 279-1971, Paragraph 4.21)
The ESP systems are designed to permit repair or replacement of components. Recognition and location of a failed component will be accomplished during periodic testing or by annunciation in the main control room.
- 22. Identification (IEEE 279-1971, Paragraph 4.22)
The ESP panels are identified by colored nameplates. The nameplate shows the division tc which each panel or rack is assigned, and also identifies the function in the system of each item of the control panel. The system to which each relay belongs is identified on the relay panels. All wiring and cabling outside of panels are labeled to indi-cate its divisional assignment as well as its system assignment. See 8.3.1.3. 7.3.2.1.3 Conformance to Regulatory Guides The following ie a discussion of conformance to those Regulatory Guf aes which apply specifically to the ESP systems. Refer to 7.1.2.4 for a discussion of Regulatory Guides which apply equally to all safety-related systems,
- a. Regulatory Guide 1.22-1972 The ESP systems instrumentation and controls are capable of being tested during normal plant operatio' unless that testing is detrimental to plant availability to verify the operability of each system component. Testing of safety-related sensors is accomplished by valving out each sensor, one at a time, and 7.3-34
WNP-2 AMENDMENT NO. 10 July 1980 ( \ k- applying a test pressure source or in the case of the main
-steam line radiation sensors, the sensors may be removed and test sources applied. This' verifies ~the operability of the
- sensor contacts,.the sensor set point, and the associated logic components in the control room. Functional operability of temperature sensors may be verified by readout comparisons, applying a heat-source to the locally mounted temperature sensing elements or by continuity testing.
For the HPCS, LPCS, and LPCI, testing for functional operabil- I ity of the control logic relays can be accomplished by use of
~ plug-in test jacks and switches in conjunction with single sensor tests.
Four test-jacks are provided to allow ADS logic testing one for each logic channel. During testing, only one logic should be actuated at a time. Ilowever , when the test plug is plugged into one channel, the complement channel of that trip system is automatically rendered inoperative. Therefore, inadvertent ADS actuation cannot occur even if both channels are impro-perly placed in the test mode simultaneously. An alarm is provided if a test plug is inserted in both channels in a division at the same time. Operation of the test plug switch and the permissive contacts will close one of the two series relay contacts in the. valve solenoid circuit. This will cause O' a panel light to come on indicating proper channel operation. Annunciation is provided'in the main control room whenever a test plug is inserted in a jack to indicate to the operator that an ECCS'is in a test status. Operability of air operated, solenoid operated, and motor operated valves is verified by actuating the valve control switches and monitoring the position change by position indi-cating lights at-the control switch. The ESF systems are provided with indications, status displays, annunciation, and computer printouts which aid the control room operator during periodic system tests to verify component operability.
- b. Regulatory Guide 1.53-1973 Refer'to IEEE 279 Para. 4.2, 7.3.2.1.2:.
- c. ' Regulatory Guide 1.62-1973 - Manual Initiation of
. Protective Actions
-rne HPCS , LPCS, and the Division 2 LPCI system are manually
('T \_,) - initiated at.the system level from the Main Control Room by 7.3-35
WNP-2 AMENDMENT NO. 10 July 1980 O actuation of an armed pushbutton. The LPCS pushbutton also initiates the Division 1 LPCI system. The ADS and the PCRVICS are manually initiated at the system (division) level by actuation of two armed pushbuttons (one for each logic channel). The RHRS-Containment Spray Cooling Mode is manually initiated at the system (division) level by actuation of the RHR pump start control switch and by opening the Containment Spray or Suppression Chamber Spray valves. The RHRS-Suppression Pool Cooling mode is manually initiated from the main control room by actuation of system pump and valve controls. The MSLC system, the SGTS, and the CIA system are manually initiated at the system (division) level by actuation of the system start control switch. The SSW system is manually initiated at the system (division) level by actuation of the pump start control switch. The Main Control Room and Critical Switchgear HVAC is manually initiated at the system (division) level by actuation of indi-vidual fan start control switches. The Containment Atmosphere Control system is manually ini-tiated at the system (division) level by actuation of the Recombiner system start control switch. The actuation of the system level manual initiation switches simulate all the actions of automatic or manual (individual equipment initiation) system actuation,
- d. Regulatory Guide 1.96 - D,esign of Main Steam Isolation Valve Leakage Control System for Boiling Water Reactor Nuclear Power Plants.
The MSLCS is designed to remove leakage from Main steam line Isolation valves by drawing a positive suction on the main steam line downstream of the isolation valves, and trans-porting leakage to the Standby Gas Treatment System. See 5.5.5.4 for a description of valve stem packing leakage removal. O 7.3-36
- . .- . -, , ... ~ - - - ~ . - - . _ , . - . ~ - - - _ . _ . _ . = ... - - . . - _ . . _ . - ...
4 J: y
. TABLE 7.3-1 '
HIGH PRESSURE CORE' SPRAY SYSTEM INSTRUMENTATION SPECIFICATIONS _ Instrument' Required HPCS'Functior; g3) Response , Instrument- Range (2)- Trip Setting Margin (4) Accuracy (5) Time-(5) ;, i Reactor Vessel Low Level Switch -150/0/+60" (1)- -38" .- ~ +7.5"
' Water Level
.(B22-NO31A-D)
- - .(Level.2) l, Drywell' Pressure Switch i High Presscre
.25-12 psig 2 psig -
+0.06 psi' LO.6 sec..
' (B22-N047A-D) -
Reactor Vessel High Level Switch 0-60" +55.5" .
+3" -
. Water Level' .(B22-N024 A,C) '
j -f !; w .
-(Level 8) ,
I 4 g w Pump Discharge Pressure Switch 10-340 psig 120 psig , . 4
+10 psi -
Pressure (E22-N012)
't j' : Pump Minimum Flow'- Flow Switch 0-1190 gpm 640 gpm -
+160 gpm 1 0 milli:sec
} (E22-N006) , t . 6 ] '. Suppression Pool Level Switch 0" -
+0.5" -
- High Water Level (E22-N002 A,B) -
(5" above normal 4 water level) ' I i Condensate Storage Level Switch - 0" -
+0.5" -
f j Tanks Low-Level (E22-N001 A s) (11,500 gal) s j Diesel Fuel Day Level Switch (H) - - i Tank Level Low (DD-LS-21) El. 445'-0"
+0
'4
% h H
_. 5" - ). 1 m Ng , i w w
- - O CD E Mi o3 q w z 4 Q e
.f s
! O i' . 1 < i 2 '
.. ._ _ -, . . , . . ,s < ~..c.. -we y v
WNP-2 AMENDMENT NO. 10 July 1980 Page 2 of 2 NOTES FOR TABLE 7.3-1 (1) Instrument zero equal to 527.5" above Vessel zero. (2) See Chapter 16, " Technical Specifications" for opera-tional limits. The range for safety-related instrumentation is selected so as to exceed the expected range of the process v~ari-able being monitored. (3) Trip settings shown are subject to change to comply with Chapter 16, " Technical Specifications". The trip setpoint is located in that portion of an instrument's range which provides the required accuracy. Initial trip setting values are established from oper-ating experience with similar size plants, and backed up with analysis as necessary. (4) See Chapter 16, " Technical Specifications" for instrument h setpoint margins. (5) values shown are subject to change to comply with Chapter 16, " Technical Specifications". t O 7.3-38
WNP-2 AMENDMT.NT NO. 10 July 1980 O TABLE 7.3-2 CHANNELS REQUIRED FOR PROTECTIVE ACTION COMPLETION Instrument Channels Minimum Channels q Channel Provided Required Reactor vessel low 4 2 water level (Level 2) Drywell high 4 2 , pressure Condensate storage 2 1
, tanks low level Suppression pool 2 1 high level N
- O l
4 .i 'l t 1 'l l I l l J F 7.3-39 i i
TABLE 7.3-3 AUTOMATIC DEPRESSURIZATION SYSTEM INSTRUMENTATION SPECIFICATIONS Instrument Required ADS Function Response Instrument Range (2) Trip Setting (3) Margin (4) Accuracy (5) Tinie (5) Reactor Vessel Low Level Switch 0-60* (1) +12.5" -
+3" -
Water Level {B22-NO38 A, B) (Level 3) Reactor Vessel Low Level Switch -150/0/+60" (1) -149" -
+7.5" -
Water Level (B22-N037 (A-D) (Level 1) Drywell High Pressure Switch 0.25-12 psig 2 psig -
+0.06 psi Pressure (B22-N048 A-D) y LPCI Permissive Pressure Switch 10-240 psig 100 psig -
+9 psi -
w (E12-N016 A-C, 1 O E12-N019 A-C) LPCS Permissive Pressure Switch 10-340 psig 150 psig -
+10 psi (E21-N001,
.E21-N009)
Automatic Depressuri- Timer 0-180 sec. 105 sec. -
+18 sec. -
zation Time Delay m 4 Ds C tQ W . C '< g O 5h CD Z rt O8 N !?:
.O
\ H
,1 O O O O
I WNP-2 AMENDMENT NO. 10 July 1980 /~^ Page 2 of 2 NOTES FOR TABLE 7.3-3 (t) Instrument zero equal to 527.5" above Vessel zero. (2) See Chapter 16, " Technical Specifications" for opera-tional limits. The range for safety-related instrumentation is selected so as to exceed the expected range of the process variable being monitored. (3) Trip settings shown are subject to change to comply with Chapter 16, " Technical Specifications". The trip setpoint is located in that portion of an instrument's range which provides the required accuracy. Initial trip setting values are established from operating experience with similar size plants, and backed up with analysis as necessary. i (4) See Chapter 16, " Technical Specifications" for instrument setpoint margins. , .7 ' O (5) Values 16, " shown are subject to change to comply with Chapter Technical Specifications". O V 7.3-41
-4 s y ---
y - - -- aw -c y , *-r-+-my e- -
WNP-2 AMENDMENT NO. 10 July 1980 0 TABLE 7.3-4 CHANNELS REQUIRED FOR PROTECTIVE FUNCTION COMPLETION OF THE ADS Instrument Channels Minimum Channels Channel Provided Required Reactor vessel low 2 1 water level (Level 3) Reactor vessel low 4 2 water level (Level 1) Drywell high 4 2 pressure LPCI permissive 6 2 l Time delay 2 1 l
- LPCS permissive 2 1 l
l l l l O 7.3-42
TABLE 7.3-5 LOW PRESSURE CORE SPEAY SYSTEM INSTRUMENTATION SPECIFICATIONS i l l Instrument Required LPCS FUNCTION Response (5) l Instrument Range (2) Trip Setting II Margin I Accuracy ( ' Time Reactor Vessel ~ Low Level Switch -150/0/+60" (1) -149" -
+7.5" -
Water Level- (B22-N037 A,C) (Level 1) - Drywell High- Pressure Switch .25-12 psig 2 psig -
+0.06 psi' -
Pressure (B22-N048 A,C) Injection Valve Differential 0-800 psid 747 psid -
+12 psi -
Differential. Pressure Switch f Pressure (E21-N006) . W 1 W Pump Minimum Flow Switch 0-1100 gpm 640 gpm -
+170 gpm -
~ Flow Bypass (E21-N004)
M C4 > W c2 D H tn (D Mg 0 00 Z m o8 tJ Z
?
O
WNP-2 AMENDMENT NO. 10 July 1980 Page 2 of 2 h NOTES FOR TABLE 7.3-5 (1) Instrument zero equal to 527.5" above Vessel zero. (2) See Chapter 16, " Technical Specifications" for opera-tional limits. The range for safety-related instrumentation is selected so as to exceed the expected range of the process vari-able being monitored. (3) Trip settings shown are subject to change to comply with Chapter 16, " Technical Specifications". The trip setpoint is located in that portion of an instrument's range which provides the required accuracy. Initial trip setting values are established from oper-ating experience with similar size plants, and backed up with analysis as necessary. (4) See Chapter 16, " Technical Specifications" for instrument setpoint margins. (5) Values shown are subject to change to comply with Chapter 16, " Technical Specifications". l l I O 7.3-44
WNP-2 AMENDMENT NO. 10 July 1980 O V TABLE 7.3-6 CHANNELS REQUIRED FOR PROTECTIVE ACTION COMPLETION OF LPCS SYSTEM AND LPCI "A" Instrument Channels Minimum Channels Channel Provided Required Reactor vessel 2 1 water level Drywell high 2 1 pressure 4 Pump discharge 2 1
, low flow Valve differential 2 1 pressure
- O d
L b O 7.3-45 i-
TABLE 7.3-7 LOW PRESSURE COOLANT INJECTION INSTRUMENTATION SPECIFICATIONS Instrument Required Response W LPCS FUNCTION Instrument Range (2) Trip Setting (3) Margin (4) Accuracy (5) Time Reactor Vessel Low Level Switch -150/0/+60" (1) -149" - 17.5" - Water Level (B22-NO37 A-D) (Level 1) Drywell High Pressure Switch .25-12 psig 2 psig - 10.06 psi - Pressure (B22-N048 A-D) LPCI Pump Delay (on Timer 0-7.5 sec. 5 sec. Loss of Normal Power) 10.75 sec. - ,4 Injection Valve Differential 0-1000 psid w 700 psid - 120 psi - Differential Pressure Switch [ Pressure (E12-N009 A-C) cs Pump Minimum Flow Switch 0-15" H 2O 3" H2O - 1120 gpm - Flow Bypass (E12-N010 A-C) (700 GPM) M C4 m C tQ H O N u)h O .T t2 ts o vi N Z '
?
O O O O
WNP-2 AMENDMENT NO. 10 July 1980 (~s 4 ) Page 2 of 2
~'
NOTES FOR TABLE 7.3-7 (1) Instrument zero equal to 527.5" above Vessel zero. (2). See Chapter 16, " Technical Specifications' for opera-tional limits. The range'for safety-related instrumentation is selected so as to exceed the expected range of the process vari-able being monitored. This may, as in the case of the neutron monitoring system, require more than one instru-ment to cover the expected range. , (3) Trip settings shown are subject to change to comply with Chapter 16, " Technical Specifications". The trip setpoint is-located in that portion of an instrument's range which provides the required accuracy. Initial trip setting values are established from oper-ating experience with similar size plants, and backed up with analysis as necessary. (4) See Chapter 16, " Technical Specifications" for instrument I 1 setpoint margins. V (5) Values shown are subject to change to comply with Chapter 16, " Technical Specifications". 7.3-47
WNP-2 AMENDMENT NO. 10 July 1980 0 TABLE 7.3-8 CliANNELS REQUIRED FOR PROTECTIVE ACTION COMPLETION OF LPCI "B" AND "C" Instrument Channels Minimum Channels Channel Provided Required Reactor vessel low 2 1 water level Drywell high 2 1 pressure LPCI pumps discharge 2 2 low flow Valve differential 2 2 pressure O I l O 7.3-48
,q A (y r\ '(
V U) TABLE 7.3-9 PRIMARY CONTAINMENT AND REACTOR VESSEL ISOLATION CONTROL SYSTEM INSTRUMENT SPECIFICATIONS I Instrument PCRVICS Function Required Response W Instrument Range (2) Trip Setting (3) Margin (7) Accuracy (8) Time Reactor Vessel Low Level Switch 0-60" (1) +12.5" -
+3" -
Water Level (Level 3) (B22-N024 A-D) Reactor Vessel Low Level Switch -150/0/+60" (1) -38" -
+7.5" -
Water Level (Level 2) (B22-N026 A-D) Main Steam Line Radiation Monitor 1 - 10 CPS 3X normal +30,'000 CPS 1 sec. High Radiation (D17-K610 A-D) Main Steam Line Temperature and 50-350*F (4) Area High Temp
+7'F 2.25 sec.
Differential 0-150*F +3*F y Temperature w-Q Main Steam Line Pressure Switch 50-1200 psig 840 psig -
+15 psi at 1 sec.
;e Low Pressure (B22-N015 A-D) setpoint Drywell Pressure Switch 0.2-6 psig 2 psig
+0.05 psi .6 sec.
High Pressure (C72-N002 A-D) Reactor Bldg. Venti- Radiation Monitor 10 102mR/hr (later) -
+9.5 mR/hr .5 sec.
lation Ext euct High D17-K609 A-D) Radiation Main Condenser Pressure Switch 0-30" hg ABS 23" hg ABS -+0.9" hg Low Vacuum (B22-N056 A-D) h H Main Steam Line Differential High Steam Flow Pressure Switch
-15/0/150 psi 132.5 psi -
+3 psi .1 sec. b e" $
(E31-N008 A-D) (E31-N009 A-D) $$ (E31-N010 A-D) (E31-N011 A-D) N e N H O' O tt W
TABLE 7.3-9 (Continued) Instrument Required Response W PCRVICS Function Instrument Range (2) Trip Setting (3) Margin (7) Accuracy (8) Time RCIC turbine steam Temperature and 50-350*F (4) -
+7'F 2.25 sec.
line space high Differential 0-150*F (4) -
+3*F 2.25 sec.
t mperature Temperature RCIC turbine steam Differential -200/0/+200 198 in. H 2O -
+2% .1 sec.
line high flow pressure in H 2O switch (E31-N007 A,B) (E31-N013 A,B) R;ac::or shutdown Temperature and Thermocouple: cooling system Differential 50-350*F space high Temperature 0-150*F (4) -
+7'F 2.25 sec.
." temperature Switches 50-350*F (4) - [3*F 2.25 sec. w I Ut R; actor cleanup Temperature and 50-350*F (4) -
+7*F 2.25 sec.
system space high Differential 0-150*F (4) - [3*F 2.25 sec. temperature Temperature Switches
- Setpoints will be established based upon operating data.
m ca W c d2 H Q Ng O eh QZ tt O >3 w Z
.O H
O O O O
WNP-2 AMENDMENT NO. 10 July 1980 ['N Page 3 of 3 NOTES FOR TABLE 7.3-9 (1) Instrument zero equal to 527.5" above vessel zero. (2) See Chapter 16, " Technical Specifications" for opera-tional limits. The range for safety-related instrumentation is selected so as to exceed the expected range of the process vari-able being monitored. (3) Trip settings shown are subject to change to comply with Chapter 16, " Technical Specifications". The trip setpoint is located in that portion of an instrument's range which provides the required accuracy. Initial trip setting values are established from oper-ating experience with similar size plants, and backed up with analysis as necessary. The initial values are the trip settings listed in the tables. I (4) Trip settings will be established during preoperational testing (normally 50*F above ambient). (5) Setpoint will be determined from instrument calibration curve. (6) Setpoint will be established after background readings are determined during startup. (7) See Chapter 16, " Technical Specifications" for instrument setpoint margins. (8) Values shown are subject to change to comply with Chapter 16, " Technical Specifications".
'O O
7.3-51
WNP-2 AMENDMENT NO. 10 July 1980 0 TABLE 7.3-10 CHANNELS REQUIRED FOR PROTECTIVE FUNCTION COMPLETION OF THE PCRVICS Instrument Channels Minimum Channels Channel Provided Required Reactor vessel low 4 3 water level (Level 3) Reactor vessel low 4 3 water level (Level 2) Reactor vessel high 4 3 presure Main steamline high 4 3 radiation Main steamline space 4 temp 3 temp hign temperature 4 differen- 4 differen-tial temp tial temp Main steamline high 4 3 flow Main steamline low 4 3 pressure Drywell high pressure 4 3 Reactor Building 2 2 ventilation exhaust high radiation Main condenser low 4 3 vacuum Reactor water cleanup 2 2 system differential flow O 7.3-52
. _ . _ _ _ . -_ . . . _ . ~.
O O O TABLE 7.3-11 REACT. BLDC. VENT. & PRESS. CONT. SYSTEM INSTRUMENTATION SPECIFICATIONS INSTRUMENT FUNCTION REQUIRED- RESPONSE INSTRUMENT RANGE (1) TRIP SETTING (2) MARGIN (3) ACCURACY (4) TIME (4) React. Bldg. Differential -3 to + 7" N/A N/A +3% - to Fuel Pool Press. Trans. HO 2 Area Differen- (REA DPT tial Pressure 1Al-1A4) (REA DPT 1B1-1B4) Differential -3 to + 7" -1/4" H 2O Ap N/A +3% - Press. HO 2 Controller
" (REA DPRC- . 1A, 1B) w I
us W w c . LG H
@ M i
o ao 2 m o8 -l' m = 1 .O i
, H l' O 3
1 1
WNP-2 AMENDMENT NO. 10 July 1980 Page 2 of 2 h NOTES FOR TABLE 7.3-11 (1) See Chapter 16, " Technical Specifications" for opera-tional limits. The range for safety-related instrumentation is selected so as to exceed the expected range of the process variable being monitored. (2) Trip settings shown are subject to change to comply with Chapter 16, " Technical Specifications". The trip setpoint is located in that portion of an instrument's range which provides the required accuracy. Initial trip setting values are established from operating experience with similar size plants, and back up with analysis as necessary. (3) See Chapter 16, " Technical Specifications" for instrument setpoint margins. (4) Values shown are subject to change to comply with Chapter 16, " Technical Specifications". 1 i O 7.3-54
WNP-2 AMENDMENT NO. 10 July 1980 TABLE 7.3-12 CIIANNELS REQUIRED FOR PROTECTIVE ACTION COMPLETION FOR REACT. BLDG. VENT & PRESS CONTROL i i MINIMUM INSTRUMENT CHANNELS CHANNELS CHANNEL PROVIDED REQUIRED React. Bldg. to Fuel Pool Area Differential Pressure 2 1 i 1 !O i f O 7.3-55 : 4
TABLE 7.3-13 CONTAINMENT INSTRUMENT AIR SYSTEM INSTRUMENTATION SPECIFICATIONS INSTRUMENT REQUIRED RESPONSE FUNCTION INSTRUMENT RANGE (1) TRIP SE"ITING (2) MARGIN (3) ACCURACY (4) TIME (4) Reactor Level Switch -150/0/160"(5) -38" -
+7.5" -
Vessel Low (B22-NO26A-D) Water Level (Level 2) Drywell High Press. Switch 0.2 - 6 psig 2 psig -
+.05 psi 0.6 sec.
Press. (C72-NOO2A-D) Header Press. Press. Switch - 150 psig - - - Low (CIA-PS21A,B) W l Ln CD O C4
'su c LO H (D '<: g H Hv Cb O CD Z tt O8 N Z
.O w
O O O - O
WNP-2 AMENDMENT'NO. 10 July 1980 '
- ('N Page 2 of 2 O
NOTES FOR TABLE'7.3-13 (1) See Chapta.- 16, " Technical Specifications" for opera-tional limits. The range for safety-related instrumentation is selected so as roiexceed the expected range of the process var,',oole-being monitored. (* Trip settings shown are subject to change to comply with Chapter 16, " Technical Specifications". The trip setpoint is located in that portion of an instrument's range which provides the required accuracy. Initial trip setting values are established from operating experience with similar size plants, and back up with analysis as necessary. i-(3) See Chapter 16, " Technical Specifications" for instrument setpoint margins. (4) Values shown are subject to change to comply with Chapter 16, " Technical Specifications". (5) Instrument zero equal to 527.5" above vessel zero. I l l 1 l 0 p 7.3-57 t-
WNP-2 AMENDMENT NO. 10 July 1980 TABLE 7.3-14 0 CHANNELS REQUIRED FOR PROTECTIVE FUNCTION COMPLETION FOR CONTAINMENT INSTRUMENT AIR SYSTEM MINIMUM INSTRUMENT CHANNELS CHANNELS CHANNEL PROVIDED REQUIRED Reactor Vessel Low 4 2 Water Level Drywell High Pressure 4 2 Header Pressure Low 2 1 O e ! l l 1 7.3-58
O O TABLE 7.3-15 STANDBY GAS TREATMENT SYSTEM INSTRUMENTATION SPECIFICATIONS INSTRUMENT FUNCTION REQUIRED RESPONSE
. INSTRUMENT- RANGE (1) '-TRIP SETTING (2) MARGIN (3) ACCURACY (4) TIME (4)
Reactor' Level Switch -150/0/+60"(5) -38" -
+7.5" -
Vessel Low (B22-NO26A-D) Water-Level (Level 2) Drywell High Pressure 0.2 - 6 psig 2 psig -
+0.05 psi 0.6 see Press Switch (C72-NOO2A-D)
Reactor Bldg. Radiation 10 102 mR/hr -
. Vent. Exhaust Monitor
+9.5 mR/hr
_ 0.5 sec y High Rad. (D17-K609A-D).
.m W
e W C D H O
$k 00 Z -
n on M Z l P
~
i i A
WNP-2 AMENDMENT NO. 10 July 1980 Page 2 of 2 NOTES FOR TABLE 7.3-15 (1) See Chapter 16, " Technical Specifications" for opera-tional limits. The range for safety-related instrumentation is selected so as to exceed the expected range of the process variable being monitored. (2) Trip settings shown are subject to change to comply with Chapter 16, " Technical Specifications". The trip setpoint is located in that portion of an instrument's range which provides the required accuracy. Initial trip setting values are established from-operating experience with similar size plants, and back up with analysis as necessary. (3) See Chapter 16, " Technical Specifications" for instrument setpoint margins. (4) Values shown are subject to change to comply with Chapter 16, "Techr.ical Specifications". (5) Instrument zero equal to 527.5" above vessel zero. O 7.3-60
WNP-2 AMENDMENT NO. 10 July 1980 TABLE 7.3-16 CHANNELS REQUIRED FOR PROTECTIVE FUNCTION COMPLETION FOR THE STANDBY GAS TREATMENT SYSTEM MINIMUM INSTRUMENT CHANNELS CHANNELS CHANNEL PROVIDED REQUIRED
- - Reactor Vessel Water 4 2 Level Drywell High Pressure 4 2 I
Reactor Bldg. Vent High 4 2 Rad O t 7.3-61 1-
. . . . _ . , _ . , _ , - - - - , . . - . , _ _ _ _ . _ . , ~ . . . . - _ _ _ _ . . . . _ . _ _ _ . . _ _ _ _ - . . . . . ~ . . . _ . . - - .
TABLE 7.3-17 CONTAINMENT ATMOSPHERE CONTROL SYSTEM INSTRUMENTATION SPECIFICATIONS INSTRUMENT REQUIRED RESPONSE FUNCTION INSTRUMENT RANGE (1) TRIP SETTING (2) MARGIN (3) ACCURACY (4) TIME (4) Recombiner Temp. Switch 0 - 340*F 300*F - - 11% Blwr. (CAC TS 1A,B) Discharge Temp. High Preheater Temp. Switch 0 - 1500*F 1150*F - 11% Discharge (CAC TS SA,B) Temp. High y Moisture Temp. Switch 0 - 340*F 200*F - 11% Separator (CAC TS 6A,B)
- Outlet Temp.
O High Recycle Flow Flow Switch 0 - 25 KSCFH 3 KSCFH - 11% Low (CAC FS 6A,B) After Cooler Press. Switch 0 - 50 psig 45 psig - 1 1% - Inlet (CAC PS 68A, Pressure B) High Moisture Level Switch - - - - - Separator (CAC LS 1A,B) y q Level High j$ e *< g O 5h CD Z rs od to Z
.O O O O
WNP-2 AMENDMENT NO. 10 July 1980 g-sg Page 2 of 2 Q~ NOTES FOR TABLE 7.3-17
'(1) See Chapter 16, Technical Specifications" for opera-tional limits.
The range for safety-related instrumentation is selected so as to exceed the expected range of the process variable'being monitored. (2) Trip settings shown are subject to change to comply with Chapter 16, " Technical Specifications". The trip-setpoint is located in that portion of an instrument's range which provides the required accuracy. Initial trip setting values are established from operating experience with similar size plants, and back up with analysis as necessary. , (3) See Chapter 16, " Technical Specifications" for instrument setpoint margins. (4) Values shown are subject to change to comply with Chapter 16, " Technical Specifications". O 7.3-63
WNP-2 AMENDMENT NO. 10 July 1980 TABLE 7.3-18 0 CHANNELS REQUIRED FOR PROTECTIVE ACTION COMPLETION FOR CONTAINMENT ATMOSPHERE CONTROL MINIMUM INSTRUMENT CHANNELS CHANNELS CHANNEL PROVIDED ' REQUIRED Recombiner Blwr. Disch. 2 1 Temp. High Preheater Discharge 2 1 Temp. High Moisture Separator 2 1 Outlet Temp. High Recycle Flow Low 2 1 After Cooler Inlet 2 1 Pressure High Moisture Seperator 2 1 Level High O 7.3-64
, . ~ .
L
\/
. TABLE 7.3-19 MAIN CONTROL ROOM & CRITICAL SWGR. RM. HVAC SYSTEM INSTRUMENTATION SPECIFICATIONS' INSTRUMENT REQUIRED RESPONSE FUNCTION . INSTRUMENT RANGE (1) TRIP JETTING (2) MARGIN (3) ACCURACY (4) TIME (4)
Control Room Temp. .- 78* DB - - - Temp. - Controller (WMA Tic 12A, B) Control Room Humid - 40% RH - - -
-Humid Controller-(WMA M1C 55A, B) 4
- Swgr. Rooms Temp. -
80* DB -
+1% -
Y Temp. Controllers @ (WMA T1C 52A, B) (WMA T1C 53A,
- B)
Reactor . Level Switch -150/0/+60"(5) -38" -
+7.5" -
Vessel Low (B22-NO26A-D) Water Level (Level 2) Drywell High Pressure 0.2 - 6 psig 2 psig -
+0.05 psi 0.6 see Press. . Switch _
m c4 (C72-NOO2A-D) y $ e 5< z Reactor Bldg. Radiation 10 102 mR/hr 's
~+9.5 mR/hr 0.5 sec s Vent. Exhaust Monitor .
High Rad. (D17-K609A-D) O tt'
$z O8 to 2
.O s
O
WNP-2 AMENDMENT NO. 10 July 1980 Page 2 of 2 NOTES FOR TABLE 7.3-19 (1) See Chapter 16, " Technical Specifications" for opera-tional limits. The range for safety-related instrumentation is selected so as to exceed the expected range of the process variable being monitored. (2) Trip settings shown are subject to change to comply with Chapter 16, " Technical Specifications". The trip setpoint is located in that portion of an instrument's range which provides the required accuracy. Initial trip setting values are established from operating experience with similar size plants, and back up with analysis as necessary. (3) See Chapter 16, " Technical Specifications" for instrument setpoint margins. (4) Values shown are subject to change to comply with Chapter 16, " Technical Specifications". (5) Instrument zero equal to 527.5" above vessel zero. O 7.3-66
I WNP-2 AMENDMENT NO. 10 July 1980 O TABLE 7.3-20 CHANNELS REQUIRED FOR PROTECTIVE FUNCTION COMPLETION FOR CONT. RM. & SWGR. RM. HVAC MINIMUM INSTRUMENT CHANNELS CHANNELS _ CHANNEL -PROVIDED REQUIRED 1 Control Room 2 1 Temperature {- i Control-Room Humidity 2 1 Swgr. Rooms' Temperatures 2 1 Reactor Vessel Water 4 2 Level 1 Drywell High Pressure 4 2 I , Reactor Bldg. Vent High 4 2 Rad. l ( 1 1
- O 7.3-67 if w- y,w m, e gg._p-,N- 9vw.u =..-mv+ - .- cm-- ye.ww- , - - = ..-+-e-g- -rew-W--MN-T- '"' Pr -T-*"t
TABLE 7.3-21 STANDBY SERVICE WATER SYSTEM INSTRUMENTATION SPECIFICATIONS INSTRUMENT REQUIRED RESPONSE FUNCTION INSTRUMENT RANGE (1) TRIP SETTING (2) MARGIN (3) ACCURACY (4) TIME (4) Reactor Level Switch -150/0/+60" (5) -38" - 17.5" - Vessel Low (B22-NO26A-D) Water Level (Level 2) . Drywell High Pressure 0.2 - 6 psig 2 psig -
+0.05 psi 0.6 see Press. Switch (C72-NOO2A-D)
Reactor Bldg. Radiation 10 102 mR/hr - - 19.5 mR/hr 0.5 sec . Vent. Exhaust Monitor y High Rad. (D17-K609A-D) as Spray Pond Level Switch - El. 415' - - 11% Water Level (SW LS 1A3, Low 1B3) Spray Pond Level Switch - El . 4 33 ' -6" - 11% Water Level (SW LS 1A2-Low 1D2) Spray Pond Level Switch - El. 433'-9" - - 11% Water Level (SW LS 1A4, Low 1B4) m 4> os c 3: tQ HM SW Discharge Pressure Switch - 50 psig - 11%
- (D
'<g Pressure Low (SW PS 1 A, 1B, 40B) s yg O
m
$Z oe N Z
.O H.
O O O O
WNP-2 AMENDMENT NO. 10 July 1980 NOTES FOR TABLE 7.3-21 (1) See Chapter 16, " Technical Specifications" for opera-tional limits. The range for safety-related instrumentation is selected so as to exceed the expected range of the process variable being-monitored. (2) Trip settings shown are subject to change to comply with Chapter 16, " Technical Specifications". The trip setpoint is located in that portion of an instrument's range which provides the required accuracy. , Initial trip setting values are established from operating experience with similar size plantn, and back up with analysis as necessary. (3) See Chapter 16, " Technical Specifications" for instrument setpoint margins. (4) Values shown are subject to change to comply with Chapter 16, " Technical Specifications". , (5) Instrument zero equal to 527.5" above vessel zero. O 7.3-69
WNP-2 AMENDMENT NO. 10 July 1980 TABLE 7.3-22 0 CHANNELS REQUIRED FOR PROTECTIVE PUNCTION COMPLETION FOR THE STANDBY SERVICE WATER SYSTEM MINIMUM INSTRUMENT CHANNELS CHANNELS CHANNEL PROVIDED REQUIRED Reactor Vessel Water 4 2 Level Drywell High Pressure 4 2 Reactor Bldg. Vent High 4 2 Rad. Spray Pond Water 2 1 Level Low (SW LS 1A3, 1B3) Spray Pond Water 2 1 Level Low (SW LS 1A2, 1B2) Spray Pond Water 2 1 Level Low (SW LS 1C2, 1D2) Spray Pond Water 2 1 Level Low ] (SW LS 1A4, IB4) ) 1 SW Discharge Pressure 3 2 l Low l l l l l l l l 7.3-70
.~..
TABLE 7.3-23 RHRS - SUPPRESSION POOL COOLING MODE INSTRUMENTATION SPECIFICATIONS INSTRUMENT REQUIRED RESPONSE FUNCTION INSTRUMENT RANGE (1) TRIP SETTING (2) MARGIN (3) ACCURACY (4) TIME (4) Reactor Level Si:ch -150/0/+60" (5) -149" -
+7.5" -
Vessel Low (B22-NO37A-D) Water Level (Level 1) Drywell High Pressure 0.25 - 12 psig 2 psig -
+0.06 psi 0.6 see Pressure Switch (B22-NO48A-D)
Suppression Temperature 0 - 212*F - -
+2*F 15 sec
,4 Pool Recorder _
w ' Temperature [ High 6 9 ! m c4 >
- W c
- T
! to rR i O H H O CD m 08 j to !2:
.O N
O
WNP-2 AMENDMENT NO. 10 July 1980 Page 2 of 2 NOTES FOR TABLE 7.3-23 (1) See Chapter 16, " Technical Specifications" for opera-tional limits. The range for safety-related instrumentation is selected so as to exceed the expected range of the process variable being monitored. (2) Trip settings shown are subject to change to comply with Chapter 16, " Technical Specifications". The trip setpoint is located in that portion of an instrument's range which provides the required accuracy. Initial trip setting values are established from operating experience with similar size plants, and back up with analysis as necessary. (3) See Chapter 16, " Technical Specifications" for instrument setpoint margins. (4) Values shown are subject to change to comply with Chapter 16, " Technical Specifications". (5) Instrument zero equal to 527.5" above vessel zero. O 7.3-72
t WNP-2 AMENDMENT NO. 10 July 1980 (m
.G)
TABLE 7.3-24 CHANNELS REQUIRED FO?_ PROTECTIVE ACTION COMPLETION FOR RHRS - SUPPh'L"SION POOL COOLING MODE MINIMUM INSTRUMENT CHANNELS CHANNELS CHANNEL PROVIDED REQUIRED Reactor Vessel Water 4 2 Level
- Drywell High Pressure 4 2 Suppression Pool Temp. 2 1 High l
l O 7.3-73
TABLE 7.3-25 RHRS - CONTAINMENT SPRAY COOLING MODE SYSTEM INSTRUMENTATION SPECIFICA;' IONS INSTRUMENT REQUIRED RESPONSE FUNCTION INSTRUMENT RANGE (1) TRIP SETTING (2) MARGIN (3) ACCURACY (4) TIME (4) Drywell High Pressure 0.25 - 2 psig -
+0.06 psi 0.6 see Press. Switch 12 psig (B22-NO48A-D) w e
W l 4 b m C4>
$. bh
<Z t2 w ..t*
Q ha O CD Z tt o8 N Z H O O O O
WNP-2 AMENDMENT NO. 10 July 1980 () NOTES FOR TABLE 7.3 Page 2 of 2 (1) See Chapter 16, " Technical Specifications" for opera-tional limits. The range for safety-related instrumentation.is selected so as to exceed the expect'd range of the process variable being monitored. (2) Trip settings shown are subject to change to comply with Chapter 16, " Technical Specifications". The trip setpoint is located in that portion of an 4 instrument's range which provides the required accuracy. 4 j Initial trip setting values are established from 4 operating experience with similar size plants, and back up with analysis as necessary. (3) See Chapter 16, " Technical Specifications" for instrument , setpoint margins. (4) values shown are sucract to change to comply with Chapter O 16, " Technical Specifics r s" . 2 J l I' t j 7.3-75 i
,_ - , _ - _ , , ,,,..._m. . . _ . . - , , _ _ . . . , , , . . . _, - - , _ , _ , . , _ . . - , ,
WNP-2 AMENDMENT NO. 10 July 1980 TABLE 7.3-26 0 CHANNELS REQUIRED FOR PROTECTIVE ACTION COMPLETION FOR RHRS - CONTAINMENT SPRAY COOLING MODE MINIMUM INSTRUMENT CHANNELS CHANNELS CifANNEL PROVIDED REQUIRED Drywell High Pressure 4 2 9 i 1 i l O 7.3-76 l l
w TABLE 7.3-27 MAIN STEAM LINE LEAKAGE CONTROL SYS'.'.EM INSTRUMENTATION SPECIFICATIONS INSTRUMENT REQUIRED RESPONSE FUNCTION INSTRUMENT RANGE (1) TRIP SETTING (2) MARGIN (3) ACCURACY (4) TIME (4) Reactor Press. Switch - 35 psig - - - Pressure (MSLC-PS20) Low (MSLC-PS24) (MSLC-tS8A-D) (MSLC-FS7A-D) MSLC Header Press. Switch - O psig - - - Pressure Low (MSLC-PS 25)
,4 MSLC Header Press.' Switch -
5 psig - - - w Pressure Low (MSLC-PS 70A-D) 1 4 J MSLC High Flow Switch - 505 CFH - - - Flow (MSLC-FS 3A-D) , i i ) M h ' W c D H
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Uh CD E i m 08 4 M Z ' , .O b O
WNP-2 AMENDMENT NO. 10 July 1980 Page 2 of 2 NOTES FOR TABLE 7.3-27 (1) See Chapter 16, " Technical Specifications" for opera-tional limits. ' The range for safety-related instrumentation is selected so as to exceed the expected range of the process variable being monitored. (2) Trip settings shown are subject to change to comply with Chapter 16, " Technical Specifications". The trip setpoint is located in that portion of an instrument's range which provides the required accuracy. Initial trip setting values are established from operating experience with similar size plants, and back up with analysis as necessary. (3) See Chapter 16, " Technical Specifications" for instrument setpoint margins. (4) Values shown are subject to change to comply with Chapter 16, " Technical Specifications". O 7.3-78
WNP-2 AMENDMENT NO. 10 July 1980 0 TABLE 7.3-28 CHANNELS REQUIRED FOR PROTECTIVE-FUNCTION COMPLETION FOR MAIN STEAM LINE LEAKAGE CONTROL SYSTEM MINIMUM INSTRUMENT CHANNELS CHANNELS CHANNEL PROVIDED REQUIRED Reactor Pressure Low 5 1 (MSLC-PS 20) (MSLC-PS 8A-D) MSLC Header Press. Low 1 0 (MSLC-PS 25) MSLC Header Press. Low 4 0
- (MSLC-PS 70A-D) l MSLC High Flow 4 0 '
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AMENDMENT NO. 10 July 1980 .s l ISOLATION TRIP SYSTEM A l ISOLATION TRIP SYSTEM B i SENSOR SENSOR SENSOR SENSOR A C B D j l l l
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u_<'_" a l CHANNEL A CHANNELC CHANNEL 8 CHANNEL D A-C POWER (REACTOR PROTECTION SYSTEM A-C POWER (REACTOR CHANNE LS PROTECTION SYSTEM M-G SET A OR A-C POWER) M-G SET B OR A-C POWER) A C B D
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ISOLATION LOGICS
] LOGIC Al LOGIC A2 l LOGIC B1 LOGIC B2 C~J LOGIC LOGIC LOGIC LOGIC A1 A2 81 82 A1 A2 B1 82 ISOLATION ACTUATORS FROM A-C POWER FROM A-C POWER RPS M-G SET A FROM A-C POWER FROM A-C POWER RPS M-G SET B RPS M-G SET A RPS M-G SET B A1 == == B1 A2 == B2 TRIP ACTUATOR A2 == == B2 A1 == LOGICS B1 INBOARD VALVES OUTBOARD VALVES (v3 WASHINGTON PUBLIC POWER SUPPLY SYSTEM FIGURE ISOLATION CONTROL SYSTEM FOR MAIN STEAM NUCLEAR PROJECT N0. 2 LINE ISOLATION VALVES. 7.3-2
AMENDMENT NO. 10 July 1980 m ISOLATION TRIP SYSTEM A ISOLATION T'g.r $ysygg 8 SENSOR SENSOR l SENSOR h-
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NUCLEAR PROJECT NO. 2 1 3 ~~j 10f { .
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l c l l lcl l c l 0 I i SUPPORilNG DOCUMENTS: ( A ### 2
- 1. PIPING & INSTRUMENTATION SYMBOLE
%% 2. PRES $URE INTEGRilf 0F PlPING &
--) SN2, c A2) EQUIPMENT PRES $URE PARTS
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- 2. R4 SYSTEM P&lo g12 3010
- 3. LPCS SYSTEM FC0 " '$ '
E21 1030 4 PROCESS INSTRtMENTAil04 A62-4070 1. PlPING HIGH PolNT VENTS & LOW PolNT ORAINS ARE TO
- 5. EMERG. EQUIP. COOLING WATER A62-4230 BE ADDED AS NECESSARY.
- 6. NUCLE AR BOILER SYSTEM FC0 822 1030 2. CHEMICAL CIEANING CONNECTIONS. VALVES. ETC.. IF
+ REQUIRED, A% PROVIDED AS NECESSARY.
7 LPCS SYSTEM PROC DIAG E21-1020
- 8. LPCS SYSTEM OESIGN SPEC 3. INSTRUMENT L t.8E DESIGN AND VALVING MUST CCMPLY E21-4010 9 LE AK DE1Eti10N SYS.1ED WITH INSTRUMiNT PIPING SPECIFICATION. SEE REF. 4 E31-1010 4.
THE METHOD OF MOUNTING LOCAL INSTRlMENTS IS TO BE 4 10. LPCS SYSTEM INST. DATA SH E21 3050 CETERMINED BY OTHERS.
- p4V 11. CLEANING OF PIPING & EQUIP A62 -4140 5 fHl5 LINE MAY BE RETURNED TO SUPPRES$10N CHAMBER THROUGH A SEPARATE LINE OR AliACHED TO THE RHL
- SYSTEM TEST RETURN LINES. IF RETURNED THROUGH THE RHR Lf1ES. A CHECK VALVE MUST BE PROVIDE 0 DOWNSTREAM T OF PO 0001 TO PREVENT BACK FLOW. SEE REF. 2. SH 2 -
ZONE (F.7).
- 6. VENT, DRAIN,&RE. LIEF VALVE DISCHARGE SYSTEMS TO CRW AND ORW OR SUPPRESSION POOL ARE BY PIPING DESIGNER.
- 7. FOR LOCATION ANO l'OfNilFICAT10N OF INSTRUMENTS $EE, REF. 10.
- 8. FLUSHING CONNECil0NS SHALL BE PROVIDED lM AC' CORDANCE '
O WITH REF 11. TEMPORARY STRAINER SCREENS SHALL BE PROVIDEO ON THE SUCTION SIDE OF ALL PUMPS IN ACCORDANCE WlIH REf 11.
- 9. THE PACKtNG GLAND OF VALVE F011. & F012 SHALL BE LOCATED ON THE UPSTREAM SIDE OF VALVE olSK.
N 10.
/ TN C 11. ALLOW ADEQt: ATE PlPING SURFACE AREA FOR COOLING OF Pm P C M2.
I - -L- ME y#O p* 12. FOR ADDITIONAL CONTROL ROOM LIGHTS. SYSTEM ALARMS, M W lWU AND REMOTE MANUAL SWITCHES SEE REF. 3
@E J M N g
D -
- 13. PROVISIONS FOR CONTAINMENT ISOLATION BY OTHERS TO BE IN ACCORDANCE WITH CURRENT LICENSING REQUjREMENTS.
i j T* G r-- t ' i I e LEGEND.
# I ep
'O D C EECWS - EMERGENCY EQUIPMENT COOLING WATER SYSTEM sY, % l "O 7004l~f003 hE @ L F0ll Rff L l l.
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- 6. IsoCLE AR SoeLER Sfl FC0 tav522-iO50
- 7. LOG 1C Stut0LS A 4 2 -+0 50
- 8. ELECT #iCAL EQUrPmEnf SEPan&fs04 FOR CW AFEGua4D SYSTEMS
- A62-40SO REVER$ isle Conf ACTOS 42e CORE SPtaf 14JECf t04 valve NO FOOS L E EE 40*
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't~~T AMENDMENT NO. 10 July 1980 seL ETEM N' EI'21010 NQfES:
- 19. A SEPARATC 5WPPRE55I04 POOL PEmE- e. twSTRUMENT LINE DESIGN 4 UALVlmG MUST T A AT Bou EIF USEC) F oe 'C ' LOOP SM AL L COMPLY ulTN Ie5TRUMEmi PIPING SPECIF-TERMINAff UNDER WATER. ICATIOu REFERENCE 9.
4 6. L INE FROM SELIEF WALTES TO SLOPE 00 # WARD C0af tuuov5LT TO SuPPRE55104 THE METHOD OF Moutf tnG LOC AL In5TRu-2' MENT 5 35 to SE DETERMINED av OTHERS. g y, 1. PIEING NIGN P0luf VENTS a Low POInf DaAl#5 AaE To at ADDED A5 4ECEb5&Rf. 4 FLUSNING C0mMECTIONS SMALL GE PROVIDED >
- 10. ALLOW ADE0VATE PIPING SURF ACE ARE A Im ACCompasCE utTN pf F 2. TEMPORAAT FOR COOLING OF PUMP Cool. STRAINEe SCptEW5 laeALL DE PROVIDED On THE SUCTI0s SIDE OF ALL PWMPS IN 49 FOR LOCaf!0m Ale 0 IDEuf!FIC ATION OF ACCORDANCE WITN REF 2.
Iutf puMENf 5 SEE th5tRUMENT Cata 5HEET
*'* 9. DI5CMa#GE List $ FOR COOLING WATER
- i. 4 20. E ECEPT Af POInf 5 0F C04*ECT!0m WITH TO et ROUTED UP5T4EAM OF SERV!CE gwtSO SUPPLIE0 EQUIPMENT 02 PIPING. WATER aa0I ATIOie M0mIT025.
THE PIPl4G OE51GNER $NALL SIZE ALL PIPES I4 CONFOAmassCE ut fN THE SYSTEM 6 EQUIPMENT 4 IuSTRUMEufS ARE PRE-DESIGN SPEC A40 PROCESS DIAGRAM, FIEED OV 575f EM NO. (EQ2) UNLESS OTHERWISE 40 FED. T. ALL MOTOR OPERATED VALVES ARE AC
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s 8. RE ACTOR NE AD 5PRAV L INE CChu-n 8 ECTION TO Malm LINE SMALL DE Ft A5 (LOSE Al PRACTICIBLE T0 tel 5 **---) es - e 3 vaLvf F006. p, Jo ,
, , , 9. RECOMMENDEO. suf NOT REQUIRE 0.
5 " I. " *** e--.--- [ t t, p to. VALuf SHOULO SE INSTALLE0 wifN e PACEI4G GL Ase0C5),UP5f aE AM 510E g & 4 J . .'. 4 . F0#6 0
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I. 4UCLEAR 80!LER STS FCD-02E-iO30
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. 'TT. .ioo.c ..o-iI h!!?!!' V M-i 'g
>>a
" 'a n ___ __.._______ g g '_;___ieZ
==s
..o.c,.v o. __ _____ _____________;
,,/g, i
, ,,,,,T
'"=.", ,,,
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.a --- .i
.s
~'wg ,:
i i i i i i 9.s. f, 4 aaa go, p__ .______.,_________aa' __ % Us : u.....
'g2 (ORS, ' 25
}s,,, ,
g4 { _ L d>,@i# -
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~
...m .:t;'
BUILDING VENTIL ATION EXH AUST S AP,4PLER
~=:'t:::, :';;rt..itytte.$rt.mt?:,mn" i
WASHINGTON PUBLIC POWER SUPPLY SYSTEM PROCESS RADIATION MONITORING SYSTEM FIGURE (RECOMB /CHARC0AL BED) IED SHEET 3 7.3-NUCLEAR PROJECT N0. 2 16c
IR-21 BPRAY PDPC A ~ LEVEL SH. SM-LS-IA3 LEVEL s 415' i E DIESEL 1A = tD
#h RUNNING EMERGENCY l PDMER gng A W
BD-A 20 SEC. AVAILABLE : j _ START RMS LPCS START PUMP IA RCIC START SPRING RUN
^
RETURN NEUTRAL DFF [ = BD-A i ELEC. TRIP SHORTCKT OR
=
- ca d TRIP l LJNDERVOLTAGE l BD-A RMS SH-V-2A SPRING DPEN ,
RETURN AUTD CLOSE E IR-21 l l PRESS SH. l l SH-PS-1A ,
^
SOPSIGsPRESS P1TR.DVERLDAD DR GROUND 00~^ M VDLTAGE $ so-A PHP.SN-P-1A CIRCUIT BKR INDPERATIVE hBD-S i
.j-
.n_
i. AMENDMENT NO.10 July 1980
-# RHR-V-68A h OPEN CLoSE --- ( g SHITCH LIGHTS & DTHER ,
LOGIC BY G.E. LOCAL SHSD STANDBY SERV SM-7 m HTR. PUMP SH-P-1A gg AND ? m
" * 'sl go-g 9
@]w .
# TI CONT.DN a X SHEET I 524-8
. l E LOCAL SH-V-2A PUMP 1A DISCH. VALVE
> > DPEN h BD-A 8 = CLoSE g i & C DWG. NO.
M- 0 SHEET 524-1 STANDBY SERVICE HATER PUNP 1 A SHT.! CONTROL LOGIC DIAGRAM . WASHINGTON PUBLIC POWER SUPPLY SYSTEM CONTROL LOGIC DIAGRAM - SSW SYSTEM y _ NUCLEAR PROJECT N0. 2
IR-22 SPRAY POND B SWITCH, LIGHTS AND OTHE WEL M LOGIC BY G.E. SH-LS-1BS [
, LVL s 415' -
l l DIESEL 18 - SM-8 BUS (27X1
*n TD 5 l
RUNNING UNDERVOLTAGE ON m (DG 2-8) 20 SEC SPRING RET.TO RHRBSTART{ RHR C START NEUTRAL BD-B RMS I m PUMP 1B _
> l RUN - -
m AND NEUTRAL E ! WF - d BD-RS - -Y RSTS-2 XFR SN FOR SW-P-1B MAIN CONT RM - BD-RS - BD-RS RMS STDBY SW PUNP 1B 5 RUN - NEUTRAL BD-B BD-RS OFF - >
- dh h %
M m , p TRIP ELECT. TRIP O& ' SHORT CUT OR UNDERVOLTAGE ( (
l AMENDMENT NO. 10 July 1980 i RHR-V-688 O > OPEN ~~* - - -> CLOSED LOCAL
--h(
-~ '
L STDBY SVCE gg.g HATER PUMP g m SH-P-1B 7 '
= RUN'
~
--* OFF B -RS X M CONT DN SHEET 524-9 MATCH LINE-SEE_ DHG.524-2A ~~
M-620 i n c OWE. SHEET 524-2 STANDBY SERVICE HATER PUMP 1B CONTROL LOGIC DIAGRAM , 1 WASHINGTON PUBLIC POWER SUPPLY SYSTEM CONTROL LOGIC DIAGRAM - SSW SYSTEM NUCLEAR PROJECT N0. 2 g3-
l l l l l SPRING RET.TD [ AUTO BD-B l m
? - l A RMS m O
I SN-v-28 I O OPEN -
=
m AND O @ AUTO ----* O CLOSE O ' m FRON l l RSTS-4 SHEET I 5 AND m h I ! 524-17
- l BD-RS RMS O SH-V-28 _
j OPEN O net' TRAL O CLOSE O IR-22 PRESS.SH. SH-PS-1B 50 PSIG 4 PRESS. g BD-RS g NDTOR OVLD hBD-B OR ,,OuND A LJNDERVOLTAGE hBD-B PNP SN-P-1B CIRCUIT SKR INOPERATIVE hBD-S I (
a t AMENDMENT N0. 10 July 1980 MATCH LINE - SEE DMG_. _524-2 __ l 1 f L LOCAL SH-V-28 PUMP 1B DISCH.VA. O TEN CLDSE
, h g BD-B WA DR y BD-RS -
M - REF.DHGS. STANDBY SERVICE WATER SYS, FLON DIAG,N-524 s a c DWG. NO. M-620 l SHEET 524-2A STANDBY SERV 1CE NATER PUNP 1B , CONTROL LOGIC DIAGRAN WASHINGTON PUBLIC POWER SUPPLY SYSTEM CONTROL LOGIC DIAGRAM - SSW SYSTEM 3 NUCLEAR PROJECT NO. 2 g
l H22-P028 FRON G.E. [ ESS LOGIC INPUT (DIESEL - - - - - - - - - - - * { ------> START) > --- l r--* '- I H13-P601 I RM8 I PUNP g i RUN -* SPRING l
-- J AEU RET 0FF f 1 l
MOTOR A AA DVERLDAD V H13-P601 l f BD-A RMS I Sw-y-2s I opgy --- a . SPRING l l
~~~--->
RET AUTO CLOSE
, g {
gl l L IR-24 1 O PRESS.SH, 3 V i SH-PS-40B > SOPSIGsPRESS ! 4 I l I I
f AMENDMENT NO. 10 July 1980 LOCAL PCS SERV. p HATER PUMP E-22-C002 '
- RUN H13-P601
- wr '
x i SIGNAL A T 4 SHT.524-13 ) CONT.ON s SHT.524-10 REF.DWGS. STANDBY SERVICE HATER SYSTEM FLON DIAG. M-524 LOCAL SH-V-29 PUNP DISCH.VA,
---> DPEN h BD-A CLDSE g l & C DWG. NO. M-620 SHEET 524-3 DIV. III WCS SERV 1Ct* NATER PUNP E-22-C002 CONTROL LOGIC DIAGRAM WASHINGTON PUBLIC POWER SUPPLY SYSTEM NUCLEAR PROJECT N0. 2 CONTROL LOGIC DIAGRAM - SSW SYSTEM 9 f[
17d l
F > IR-21 g_ y + _ SPRAY POND A 2 >
- j LEVEL SH. O SH-LS-1A2 -
I # LEVELS 433'-I' - :- O BD-A RMS - SH-V-12A SPRING I OPEN # RETURN # AUTO TO AUTO O m. CLOSE 6 ' l I - BD-A i 1 ! RMS SH-V-6SA -
=
f (_ SPRING OPEN
- I q RETURN AUTO # I
> TO AUTO g CLOSE
' ' F #'
i A I - d i BD-B A - v - 1 l C= RM Z # #
> SH-V-70A 5 g
3 SPRING OPEN RETURN # AUTO TO AUTO O m i CLOSE 6 - l i 1R-26 j l BPRAY POND A > 1 LEVEL SW. l 1 SH-LS-1D2 LEVEL 4 433'-I' 1
, 1
AMENDMENT N0. 10 July 1980 LOCAL ' l SPRAY POND B } SHUT-OFF VLV $30-A f SH-V-12A m
) >
O OPEN CLOSE g BD-A n LOCAL n SPRAY POND A REF.DHG. V CLG. THR. STANDBY SERV.HTR.FL.D. M-524 SHUT-OFF f SH-V-6SA
> OPEN BD-A
> CLOSE -\
'mV I
LOCAL
> SPRAY POND A ) CLG, THR.
SHUT-OFF SH-V-70A
> 0 OPEN .- J BD-B I
p
) > CLOSE
% I & C DWG. NO. M-620 4
, SHEET 524-4 SPRAY POND A LEVEL CONTROL LOGIC DIAGRAM ,
l WASHINGTON PUBLIC POWER SUPPLY SYSTEM CONTROL LOGIC DIAGRAM - SSW SYSTEM f NUCLEAR PROJECT N0. 2
IR-22 f # SPRAY POND B ( LEVEL SW. A 7 t h [ SW-LS-182 I LVL # 43 3'- 6" BD-RS RSTS-3 XFR.SW.FOR SW-V-128 & SW-V-698 MAIN CONT RM BD-RS BO-RS . RMS . SW-V-128 1 OPEN - CLOSE SD-RS . RMS [ SW-V-698 f DPEN - O CLOSE : 1 I 1 BD-B - RMS ' SW-V-128 - OPEN - C _ ' AUTO l (
~
CLOSE BD-B - = RMS E I SW-V-69B >_ ' l 7 E J ( OPEN - . i
~ ~
f 3 AMD I l ! O ' CLOSE u __ MATCH L_IN_E-SEE DWG 524-5A 7 1 1 l l l l l
AMENDMENT N0. 10 July 1980 LOCAL BD-B l' r- m SPRAY POND 0 A SHUT LYF - - O C VALVE SH-V-128 BD-B
> > OPEN , h OR
> CLOSE gBD-RS LOCAL SPRAY POND B CLG THR M
SHUT OFF SH-V-69B BD-B l
> TEN m
5 CLOSE hBD-RS _ _ _g 5 1i m A
\
5J i
*~ =
0 i i a c DWG. NO. M-62 0 SHEET 524-5 SPRAY POND B LEVEL CONTROL LOGIC DIAGRAM J l WASHINGTON PUBLIC POWER SUPPLY SYSTEM FIGURE CONTROL LOGIC DIAGRAM - SSW SYSTEM 7.3-NUCLEAR PROJECT N0. 2 17f
g - MATCH _L._INE-SEE DMG 5?4-5 BP RS {
> RSTS-1
( ** O XFR SH.FOR SH-V-70B MAIN CONT RM g BD-RS BD-RS = RMS = SH-V-708 OPEN - 5 F O CLOSE : 2 A O 0 h
- =
m BD-A 5 l [ RMS . . SH-V-708 - [ E - l wm ( AUTO
=
C CLOSE
- 2 -
IR-25
)
SPRAY POND E , LEVEL SW, 5 SH-LS-1C2 5 l q LVL 4 433'-( 1-
AMENDMENT N0. 10 July 1980 i L LOCAL 9 PRAY POND B CLG.THR. SHUT OFF SH-V-70B M BD-A O OPEN OR
- CLOSE hBD-RS N
V l l REF DHG. STANDBY SERV.HTR.FL.D.N-524 N J l l I l l l I & C DWG. NO. H-520 l l SHEET 524-5A SPRAY POPO 5 LEVEL ' CONTROL LOGIC DIAGRAM J WASHINGTON PUBLIC POWER SUPPLY SYSTEM FIGURE d CONTROL LOGIC DIAGRAM - SSW SYSTEM 7.3-NUCLEAR PROJECT NO. 2 77g
RMS 7 ,--- CONTROL; A qNE B IAIR Co TMU l LSF I t (LS3 d TMU SPRAY POND A y IR-21 ' I SPRAY q' POND A LEVEL SWITCH i TMU-LS-1A4 l LVL 4433'-9" > m I - RMS BD-A ;, C > DE TMU-LCV-1A l c-> l SPRING OPEN i -E
- l I
RETURN TO AUTO AUTO CLOSE ----------------* hd \ I n-I
AMENDMENT NO. 10 July 1980 V-1A NC f3 0 _4 k f 30 " TNU( 21- 1 THU-LCV-1A REF DHG, STANDBY SERVICE HATER FL,-D N-524 IR-21 LOCAL PRAY POND SPRAY A HATER POND A NAKE UP HATER NAME SOLENDID UP SHUT VA 0FF VA. FMU-LSPV- THU-LCV 1A -IA = ENERGIZE : OPEN ENERGIZE : CLOSE g DIV. I i & C DWG. NO. M-620 SHEET 524-6 SPRAY POND A HATER NAKE-UP CONTROL SCHENATIC AND LOGIC DIAGRAM .,, WASHINGTON PUBLIC POWER SUPPLY SYSTEM FIGURE NUCLEAR PROJECT N0. 2 7.3-17h
l ( RMS ____ gCONTROL_A LAIR Co l l I (LS3 - QB4) TNU SPRAY POND B U 1 1R-22 SPRAY POND B LEVEL SHITCH TNU-LS-184 LVL 4433'-9" > g
> n BD-B v :
RMS > M TNU-LCV-1B l 3pggg OPEN -l = _------* I RETURN TO AUTO AUTO O hJ CLOSE ( T
)
I AMENDMENT N0.10 ! July 1980 l NE h TMU n' LSPV-10 l NC O O l
---O i k f30" THU(21-1 THU-LCV-I B l
IR-22 LOCAL l SPRAY POND SPRAY A HATER POND B NAKE UP WATER MAKE REF DNG, SOLENGID UP f3 HUT STANDBY SERVICE WATER FL,D. M-524 VA 0FF VA. THU-LSPV- TMU-LCV IB -18
- DE ENERGIZE : OPEN
- ENERGIZE : CLOSE g BD-B DIV. II I & C DWG. NO.
H- 0 l SHEET 524-7 SPRAY POND 8 HATER MAME-UP CONTROL SCHENATIC AND LOGIC DIAGRAM l WASHINGTON PUBLIC POWER SUPPLY SYSTEM FIGURE A - 7.3 7 NUCLEAR PROJECT N0. 2 172.
I ll
~
_ A 2
- 0 D 3 B O
,' 6 C
2 f A4 - 3 E ,' 0 S 3 R 0 O C g aN* C gl N P8 O D 7 s n M1 L U - F PS t 9 1 S
- W<
o I M T c SFP - Ay C - G PN LS 5 WF S 0 2 3 1 2 N C A - A O A P M - P7 L 7 E A D M1 F 1 T - B
- D U -
PS B < B I M T *F - RF P W 0 H - G S 2 m ' ,' RW 8 1 S
i' ,'
= D = :
D N 5 C T 1 E S L L L - A A W A A W A 8 H C 1A 2 O O C 12 1 D C 1 1 O - 4 L O L L 1
- 1 O - 1 F L C - F3 L - F L C -
- C
- T1 C- 5 C- 15 C- 15 M
O E4E- A F-1 M P A F- P M A F- PG M _ R F H2S5 R PW G 0 M DH G M DH S 5 S 57 5 3 0
m
.CLG COIL E
m RRA-CC-11 g SH-F15-40 g go-g
,S- 10GPM DFLON M.E
>8 o C LOCAL x n g LOCAL CLG COIL U6 FAN COIL RRA-CC-5
= v, (LPCS) SH-TS-25
'P c RRA-CC-5 -
" ~
SN'F15-25
$ 40GPM tFLON O M
9 8 i Ei LOCAL x ' l2 Hq RECOMB g SCRUBBER gCO243 S SH-FIS-69A - n o 38GPM DFLON O $ BD-K1 ', b E a g A MATCH LINE-SEE DNG. 524-8A
=
g .
=
e a I
=
bh 2 s "5 -e z 9 a E9
, . em 2
O g ,M CD qg - w U. Ih (- ' O
f f_. s ~,
' 1 \
MATCH LINE-SEE -- DHG. 524-8 LOCAL LOCAL CLG COIL CLG COIL RRA-CC-2 SH-TS-23A BD-R C-2 SW-FIS-23A 120*Fn N e 23 GPM DFLOW = LOCAL l CLG CGIL i RRA-CC-12 SH-FIS-41 y BD-R i 15 GPM *LCW > l i H g REC 0t2 i RRA-CC-13 ~ I SH-FIS-60 5 j 8 GPM 3 FLOW >- & , I LOCAL l LOCAL CONTROL CONTROL RM , ROOM - AIR MANDL. AIR MAFDL . COIL CGIL ! HMA-CC- SH-TS-35A 51A-1 BD-PCI) SH-FIS-35A > OODN 70GPM 3FLON i
LOCAL CABLE ROOM AIR HANDL. COIL AIR HANDL g wng.CC- COIL E 52A-1 8H~TS-35A BD-P(I) SH-F15-36A 120,FsTEMP g RS 30GPM 3FLDH '
>8
=r AE 23 m x LOCAL LOCAL 0g SH. GEAR SH. GEAR
= vi AIR HANDL. AIR HANDL
- P$
m COIL COIL Q HMA-CC- SH-TS-37A y 53A-1 BD-PCI) 120 F4 TEMP q 4 SH-FIS-37A - v i E 45GPM 3 FLOW D n LOCAL l 8 ANALYSER ! $ R00M 4 r RRA-CC-15 g., j M SH-FIS-43 SGPM 3 FLOW g * - 0 E i
@ E E REF DHG.
3 STANDBY SERVICE HATER FL.D. M-524 l 3 @ ._ N un b E %' $ EE v, -
- ~
G E Y -E g g STANDBY SERVICE HATER A LON FLON ALARN gm a CONTROL LOGIC DIAGRAM o -' y O we a
# YE m . ,
. . ~ .
.va ~n l
CO298 LOCAL RHR PUMP B h SH-FIS-17B-2 y 0 8 f i FROM SHEET 4 TD ON 8GPM > FLOW LOCAL CO298 D [ 524-2 15 SH-TS-17B SEC 120*FsTEMP LOCAL PRA-CC-1B C0300 BD-B SH-FIS-428 > 6OGPM > FLOW L h LOCAL RHR PUMP C h BH-FIS-17C-2 BD-B 8GPM >FLON LOCAL SH-TIS-17C DNA-CC-21 . SH-FIS-14A > m 75GPM 3FLON > d CO301 BD-C LOCAL DMA-CC-22 SH-FIS-149 + 30GPM 3 FLOW ? LOCAL
RRA-CC-6 LOCAL SH-TC-29 RRA-CC-6 BD-R 120 FsTEMP SH-F13-29 h 8GPM > FLOW = j ;;; z 4
- F-i 9 LOCAL
,E g
" " CLG COIL 3E o,
LOCAL RMt [rn g CLG COIL RRA-CC-3
- 9 (RHR) SH-TS-23B 2 m RRA-CC-3 BD-R 120.PsTEM ,
3 SH-FIS-238
-< 23GPM tFLDH 4
4 9 LOCAL n e DIESEL , g CABLE CLG CO281 r DNA-CC-51 5 SH-FIS-38 BD-C 50 30GPM >,FLDH i g 5 g g - __ _ _ MATCH LINE-SEE DHG. 524-SA z = n .
, O
=
sn
~ m, g
a t - s= a = 2 wg
$M
= 8'i z
o M . O LE o e im
n , i nm I i
~~
MATCH LINE-SEE DWG. 524-9 LOCAL RRA-CC-14 SH-F15-61 - BD-K-II LOCAL SGPM 3FLON G COIL LOCAL RHR CLG COIL RRA-CC-1 tRNR) SH-TS-23C
~
SH-FIS-23C 120 FsTEMP r 0--@ J 23GPM tFLOW LOCAL CLG CDIL RRA-CC-10 BD-R SH-FIS-39 10GPM > FLOW > LOCAL H g RECOMB CO244 SCRUBBER SH-FIS-69B > [ BD-K(II) 38GPM ?FLDH & LOCAL LOCAL CO.W R . CONTROL RM ROM AIR HAND AIR HANDL COIL DIL 4 NNA-CC- ~ 518-1 SH-TS-35B II' SH-FIS-35B 100 FsTEMP
, _ .~ _
LOCAL CABLE RM E AIR 'EA'
$ HANDLING CABLE ROOM E COIL AIR HANDL -
S I' gE n WNA-CC-SH-TS-35B r 528-1 Bo-p(!!) O, SH-FIS-368 y 120* FsTEMP
;8 n 30GPM 3 FLOW = '
83 RE H := 2 m LOCAL P 55 LOCAL
" SN, GEAR 3 SN. GEAR m AIR HANDL. AIR HANDL y COIL CDIL g HMA-CC- SM-TS-37B 538-1 o BD-PCII1 SH-FIS-378 ?
M 8 45GPM tFLOW ?
"i E
r LOCAL
, 5 ANALYSER j p; ROOM
, RRA-CC-17 i S BD-R o SH-FIS-44 ? e g A
- 8GPM tFLOW ?
l [ ,,, , "I o > un h * , e m g c., 3 l 4 i - 5* t i a m 2 8
*=
m2 E STANDBY SERVICE HATER S LON FLOW ALARM g m
" CONTROL LOGIC DIAGRAM o* '
i
= .
[ H .M) m 4 WC - - t
- 3 3$ O
_ - - ~ l l 1 t
=_= =__-__- = = 2-___ 2 _ _ __ ,. - < ,,
1 FROM ( BD-GI SHEET X tD ( 524-3 DN HPCS PUNP 15 SEC, DISCH.HDR. SH-PS-40A l
- I 50 PSIG m i
> PRESS LOCAL DNA-CC-31 SH-FIS-8A 75 GPN m 4 FLOW LOCAL DNA-CC-32 SH-FIS-8B O
30 GPM m 4 FLOW LOCAL FAN COIL (HPCS) RRA-FC-4 SH-FIS-27 5 38 GPM _ 4 FLOW , LOCAL FAN COIL (HPCS) RRA-TC-4 DISCH. TEMP. SH-TS-27 120 F4 TEMP LOCAL CO306 SH-FIS-9 l 500 GPM >
>f FLON l
k
-i
I f i AMENDMENT NO.10 f July 1980 ! [ ED-A e
@BD-A l
l l BD-A h
~
o g^ ao-REF.DNG. STANDBY SERVICE WATER FL. D. M-524 f
)
/ M-620 I i & C DWG. NO.
BD-A ' MER 524-10 HPCS SERVICE WATER ' LOW FLON' ALARN
?
CONTROL LOGIC DIAGRAM WASHINGTON PUBLIC POWER SUPPLY SYSTEM FIGURE ) CONTROL LOGIC DIAGRAM - SSW SYSTEM 7.3-NUCLEAR PROJECT NO. 2 17n
S TW3 ND* LOCA 1 SH-LS-1A1 SDI [ SH-TS-1A BD 2 l SH-LS-1C1 BDI l SH-TS-1C BDI l l SH-LS-181 BDH I SH-LS-1D1 BDH SH. LOCATION SH-TS-1B BD-TAG NO.& DESCRIPTION SH-TS-10 BDH SET POINT h { TION l l 1 l 1 l l }
)
1 i [jg DESCRIPTION SET POINT LOC ION RSAM JI SPRAY POND *A"HI/LO HTR. LEY. 434'-6"/433'-6" BD-A I D 'I SPRAY POND *A"HI/LO HTR. TEMP. 85 F/35 F ' BD-A l L]I SPRAY POND *B"HI/LO HTR. LEY. 434'-6"/433'-6" BD-A l OI SPRAY POND *B"HI/LO HTR. TEMP. 85 F/35 F BD-A III SPRAY POND"B"HI/LO HTR. LEV. 434'-6"/433'-6" BD-B i!! SPRAY POND *A"HI/LO HTR. LEY. 434'6"/433'-6" BD-B 11 I SPRAY POND *B"HI/LO HTR. TEMP. 85 F/35 F BD-B III SPRAY PONDA"HI/LO HTR. TEMP. 85'F/35"F BD-B i a c DWG. NO.
"~
SHEET 524-11 ALARM ANNUNCIATOR & COMPUTER INPUT AMENDMENT N0. 10
' '^ ^
July 1980 - WASHINGTON PUBLIC POWER SUPPLY SYSTEM FIGURE
^ ~
~
NUCLEAR PROJECT N0. 2 i 17o I
\
C0h f f i $ \ 1 CD l CD l I S I GNAL *A* LOCAL 16 SEC. [ SEE TABLE tD FLOH SHITCH
$L SEE TAKE ON GPM > FLOH >~
(
PoINPUT PTO ANN. FLON SHITCH I S I EAL ,,A,, SET POINT 3g LOCATION NUMBER l 304 BD-C DIESEL 1A START SN'-FIS-12 FLDH < 1400 GPM 305 BD-C DIESEL 1B START SH-FIS-15 FLDH < 1400 GPM I & C DWG. NO. M-620 SHEET 524-12 , STANDBY SERVICE HATER
- LON FLDH ALARM" CONTROL LOGIC DIAGRAM AMEN 0 MENT N0. 10 July 1980 WASHINGTON PUBLIC POWER SUPPLY SYSTEM FIGURE NUCLEAR PROJECT N0. 2 17p l
{ i SIGN / i 1 RHR PUNP' RHR PUNPl RCIC TURBI LPCS PUN HPCS PUN DIESEL'il DIESEL ll HPCSDIE$ OR E22-C@ (SH.51 SIGNAL *A" l l BD-N
=
RMS > RET,TO [ AUTO AUTO CLOSE l
/
AMENDMENT h?. 10 July 1980 RMS FOR RMS & LIGHT LA VALVE LOCATION SEWICE A START SH-V-24A BD-N RHR A BRG.& ROOM CLG.HTR. C START SH-V-24C dC ir 1r NE START SH-V-34 RCIC PUMP ROOM CLG.HTR.
- START SH-V-44 NOTOR BRG.& ROOM CLG.HTR,
> START SH-V-54 PUMP ROOM CLG.HTR, 4 START SH-V-4A DIESEL 1A CLG.HTR. $ START SM-V-48 1r 1B [L STftRT SH-V-4C HPCS DIESEL $2 START O-3) 1r 1r VALVE
- OPEN h BD-N O CLOSED h M- 0 i a e owG. NO.
l SHEET 524-13 STANDSY SERVICE HATER SYSTEM SMEET 13 < CONTROL LOGIC DIAGRAM L l WASHINGTON PUBLIC POWER SUPPLY SYSTEM CONTROL LOGIC DIAGRAM - SSW SYSTEM 7. NUCLEAR PROJECT N0. 2 , 17q
f i BD-RS RS-IP-1 INSTR.PHR, SUPPLY SW, ON l OFF J
AMENDMENT N(1. 10 July 1980
> 125V DC ANNUNCIATOR
> P.S. ON BD-RS
> I20V AC INSTR.
> P.S. ON BD-RS ENERGISE ~ ~
DE-ENERGISE
> DN IR-22 (120V AC INSTR.P.S.)
nacowG.W. M-620 SHEET. 5 2 4-14 1 POWER SUPPLY LOGIC DIAGRAM 1 WASHINGTON PUBLIC POWER SUPPLY SYSTEM FIGURE A - 7.3-NUCLEAR PROJECT N0. 2 17r
( ( RCIC PUNP START BO-N RMS - SH-V-245 i { ( em AUTO I z 2 CLOSE BD-RS y ) RSTS-4 TRANSFER SH. FOR SH-V-2B TO 524-2
& SH-V-24B
)
TO BD-N , CONTR.RN.
- BD-RS #
l BD-RS TO 524-2 RMS SH-V-248 OPEN CLOSE # 1
AMENDMENT N0. 10 July 1980 P LCL , l RHR-B BRG.1 ROOM CLG.HTR SM-V-248 h A m BD-N v - OPEN R og 5 CL0sE h BD-R5 =n M =J i a c DWG. NO. M-520 SHEET 524-17 ! RHR-B BEARING & PUT ROOM CLG.HTR. ; CONTROL LOGIC DIAGRAM l
. . l WASHINGTON PUBLIC POWER SUPPLY SYSTEM FIGURE CONTROL LOGIC DIAGRAM - SSW SYSTEM 7.3-NUCLEAR PROJECT NO. 2 :
17s :
NOTE LOGIC FOR D! VISION II (B SUFFIX 1 I.. UNITS SIMILAReEXCEPT AS 'NOTED. f RE FLOW DIAG.M-548 1 d PIS-51 A ( A P> 1" HG -#
\ AP< . 4" HG -->
km > TO SHEET 3 v
)HMA-FN-52A y STARTER 4 POSITION /
PULL TO TURN AT RMS RMS \ 51A 52A
\
- STARTER h
f 51A & 52A OFF f b h ENERGIZED
- DEENERGIZED f
TIS-503 CALLING FOR ELECTRICAL h IST EQUIP. ROOM
#f G HRA-EHC-51 1ST STAGE MS-55A RH <90K FOR IST - -
STAGE HEAT r ENERGIZED RH >90K g CALLING FOR m m IST STAGE 2ND _ -W ' DEENERGIZED STAGE HEAT ~ 2ND STAGE NOT CALLING ,-{_ > LEVEL NORM. - ENERGIZED FOR 2ND STAGE HEAT i "% m 2ND STAGE LOW LEVEL
-->V DEENERGIZED BC FS-503 RH <40K _
RH >40K ( l l I 1
AMENDMENT N0. 10 July 1980 TS-12A CALLING FOR - HMA-EHC-SIA IST STAGE HT.
~
BD-P CALLING FOR l . 1ST STAGE ON k b A 2ND STAGE HT. 3 2ND STAGE ON m NOT CALLING m
+-
FOR HEAT OFF $
? _ b TCV-12A m s i
PDHER 5 5 TO CHILLED m m HTR. VALVE
> ON
-> OFF HMA-FN-51A ON h
gg,
? 0FF h PS ON sw.TCV-11A (EHO)
STANDBY 10 SEC. SERV.HTR.LINE " PONER TO TD - STANDBY P >100 PSI >
- g SERV HTR VA P <60 PSI -
- ON
[ m 5 TD h 0FF M PHR. ON ALLOWS VALVE b 10 SEC. TO CLOSE VIA THERMOSTAT _ IN CONTROL ROOM l l OV MUMIDIFIER HTR
- -> HU-55A
> ENERGIZED y DEENERGIZED
\ I & C DWG. NO. M-620
, SHEET 548-1 I
L l WASHINGTON PUBLIC POWER SUPPLY SYSTEM CONTROL LOGIC DIAGRAM - HVAC FIGURE MAIN CONTROL ROOM & 73-NUCLEAR PROJECT NO. 2 CRITICAL SWITCHGEAR i8a
i l l INSTR. Tb HMA-TS-206 l LOCATION -207l
-208l INSTR TAG NO. -209 !
-210i CONDITION h C TION -211l
-212l
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HMA-SMD-51 l HMA-SMD-520! HMA-SMD-5381 HMA-SMD-531 NEA-SMD-51 NEA-SMD-53Ol NEA-SMD-53B l NEA-d PS -7 3 Al HEA clPS -7 3B CL 2 ANALYSEi CL 2 ANALYSEI I l I. i
i t AMEN 0 MENT N0. 10 July 1980 ND* L y CONDITION LOCAT DN ' LOCAL TEMP > 125'T FIRE PNL TEMP > l TEP > TEMP > TEMP > TET > TEMP > TEMP > TEMP > TEMP > TEMP > TEP > TEMP > U U DP < 0 . 5" H . G . BD-P1 DP < 0 . 5" H . G. BD-P2 t SR 15 BD-P1 t SR-16 y BD-P2 1 & C DWG. NO. M- 62 0 SHEET 548-2 r HVAC ALARMS CONTROL LOGIC DIAGRAM J WASHINGTON PUBLIC POWER SUPPLY SYSTEM CONTROL LOGIC DIAGRAM - HVAC FIGURE PAIN CONTROL ROOM & 7.3- ! NUCLEAR PROJECT N0. 2 CRITICAL SWITCHGEAR 18b
(. RE, FLOW DIA, M-548 ( HEA-FN-52 RMS(LOCAL PANEL 1 ._ -
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> ENERG12ED
"
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' *n HNA-AD-52 OPEN TO h aE
^E DPEN x m to w CABLE CHASE z x .
AUTO f . r - l CLOSE f Jh CLE h SMD-52B NORMAL - TRIPPED ? h BD-FP t i
AMENDMENT N0. 10 July 1980 TO CHILLED WATER VALVE. TS-52A CALLING FOR I ST STAGE ECH-52A HEAT 5 5 1ST STAGE ON l _ _ CALLING FDP, , E O2ND STAGE ON 2ND STAGE ' g6 d : DEENERGIZED HEAT NOT CALLING FOR HEAT L FILTER CIRCUIT
=
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g yV - o I & C DWG. NO. M-620 SHEET 548-3 WASHINGTON PUBLIC POWER SUPPLY SYSTEM CONTROL LOGIC DIAGRAM - HVAC FIGURE MAIN CONTROL ROOM & 3-NUCLEAR PROJECT N0. 2 CRITICAL SWITCHGEAR 7{8c l
RE. FLOW DIA. M-548 ( l dl SPV-53A
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AMENDMENT NO. 10 July 1980 _m TO SERVICE WATER
~ '
VALVE TCV-168 TS-53A CALLING FOR 1 ST STAGE EHC-53A
^ b 0 1ST STAGE ON g _ _
CALLING FOR , 2 O2ND STAGE ON 2ND STAGE [ - DEENERGIZED HEAT C NOT CALLING FOR HEAT CIRCUIT TACTOR) WMA-FN-53A g O N O OFF l BD-P 1 & C DWG. NO. M-620 NEA-FN-53A g SHEET 548-4 ! OrF g ; WASHINGTON PUBLIC POWER SUPPLY SYSTEM CONTROL LOGIC DIAGRAM - HVAC FIGURE MAIN CONTROL ROOM & 3-NUCLEAR PROJECT N0. 2 CRITICAL SWITCHGEAR 7[8d
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LCL LCL MOISTURE SH. HEATER HOA-MS- HOA-EHC-SOA -> 54A 7 0 % <RH > > ON C0458 b C0457 -> - 0FF i m AIR FLOH OK l % -P (INTERNAL) BD-RAD-1 -
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i' AMENDMENT N0. 10 July 1980 hZ .- -. -* 3@ q f VBD-Ic l , RMS I ' " " t_ SOLEHotD% CONTRDLVA AUTO --- b1 SPV V-2A \ CLOSE OPEN -- -- Jh DEENEEl2ED e e # = -OPEN $ ; ENERGlZED e # /= .CLOSE $ IbD-jEEE8- , " ' VBD- T<II I I LOCAL RMS L necova c m r u .v4 AUTO ---+b SPV V- 2 B \ CLOSE OPEri - - - - - -- @- -
- - - - - - - +
DEENEElZED i c = OPEN -% ENERGlZED 2 CLOSE $ k N O T E S *. ~AL TO
- 1. ISOLATION SIGNAL CODES: G OSE OPEN A- REACTOR VESSEL LOW WATER LEVEL (DfV. I d.IL)
F- HIGH DRYWELL . PRESSURE (DlV. I h H) Z- HIGH RADIATION, REACTOR BUILDING VENTILATION SYSTENI (DIV.I4 I)
- 2. ALL RMS ARE SPelNG RETURN To AUTO FDSFFION (SEE FIG, I )
- 3. FOR REFERENCES SEE DWG, + M 544 ;
W
-VAC SG N ,_i S 1 & C DWG. NO.
T SHEET F . f WASHINGTON PUBLIC POWER SUPPLY SYSTEM
~
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*4& S D - K$. !MotSD 70*/oR 6A ED-KI ' TEMP')250 F 6 P2 %D-R5 TEMP) 250 F 7A SD - xi -
TGMP/2p' F Ta e n-x3, - TEM P)25b*F 6A-1 ab-KI TEMP / 2 to' F , [661 BD-K5 - TEMP)2W'F M (p10 f7A-l sD-x1 - TE MP) 2 ld F ' l a C DWG. NO. )76-l Eb-K3 - TE MP) 210 F l 2A-i so-x2 now<3 q, sneer 544-8 og 9-1 so-xii nowsoo* I WASHINGTON PUBLIC POWZR SUPPLY SYSTEM I E CONTROL LOGIC DIAGRAM - SGTS F,y G_ NUCLEAR PROJECT N0. 2 19h i
m = - ~ - -- k . ( l.o d AL
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66 C T^ rsts G Fer< taxTico Locac J $[\ mr uusT Iv4TRTAc ut (Y) C.o n b , 4
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AMENDMENT NO.10 July 1980 b lusta, TAG doub trioN 1N ST R- TAG (o~n it io n 9Nunctate% Mi (x) ut (Y) Lo CATIO N d4T- TS IAS TEMP < 60 F 46T-TS- 1 A t TEMN bo *F Bb- U 6&T-Tf, l%-3 ITEHf(hg of %T-T S -1%-l TC Hr(80 #F 4 D - R 3, 64T- Ts- 1A 4 itHr> l@ % S4b TS-1A1 TEMP)l10*i B D ' t'- I-GGT-TS- 1% 4 TEMr> l'LO 'F 64T- T6- 18 2 Tp4P)l20*F BD-K5 1 & C DWG. NO. MA10 " SHEET M OF WASHINGTON PUBLIC POWER SUPPLY SYSTEM CONTROL LOGIC DIAGRAM - SGTS NUCLEAR PROJECT NO. 2 7{g
i l ( i l l BD- Lor AL. RMS i.. DE. LUGE. A SOLE.MolO " valve. " B O PE.y : EME.RG11ED 0 0 0 CLoSE : DE.tMERGi?.ED -0 0 0 ( i ( .
AMENDMENT NO. 10 i
; July 1980 i -
I RMS BD- DELUGE sol. ELECJ. DELUGE.
' A" vA
- B Div. VALVE "C"
% T e spv-Fi KI %T-SPv-S i A %T-Pc.v- F i l 52 K1 F2 A F2 ;
! F3 KI F3 A FS I l 54 KE F4 g F4 , i F5 xY F5 & FL . 1 i 9 1:6 KE F o FG B C V FG { i LeCAt. DEL.O CmL VALVE "C " 0 0 CPEN O O c.LO MD i I & C DWG. NO. M- 62O j SHEETMF STAMDBY GAS TREATMEMT DELUGE VALVE cokiTRot LOGnc blAGRAM . WASHINGTON PlTdLIC POWER SUPPLY SYSTEM FIGURE A M-NUCLEAR PROJECT N0. 2 7.37-19L
FROM SMT.556-2 { =
~
1
= :
RMS CIA-C-1A l DN STANOBY I l Drr i REMOTE l
)
LOCAL PB CIA-C-1A DN OFF u.- IR-72 AIR RECEIVER PRESS.5W, CIA-PS-15 I40PSIGSPRESS (
AMENDMENT N0.10 July 1980 FROM !MT.555-3 LOCAL 2 CONTAINMENT INSTRUMENT AIR COMP. CIA-C-1A
- ;) #
O _ RUN h BD-A
+ OFF I f CONT ON I SNT.556-2
; R 8
~
. SHEET . 5 5 6 - 1 CONTAINMENT INSTRUNENT AIR COMPRESSOR CIA-C-1A CONTROL LOGIC DIAGRAM 4
WASHINGTON PUBLIC POWER SUPPLY SYSTEM FIGURE
" ^ "- ^ 7 3~'
NUCLEAR PROJECT NO. 2 20a
II FROM SHT.555-1 ( i - 80-B > _
=
RMS CIA-C-IB DN I STANDBY : OFF REMME : l l LOCAL PS CIA-C-1B ( 1 0FF ?
=
IR-72 AIR RECEIVER 1 PRESS.SW. l ClA-PS-1E 140PSIG> PRESS FRC l 1
'J l
! H
fAMENDMENTNO.10 July 1930 IOM SHT.55E-4 , LOCAL l CONTAINNENT ! INSTRUNENT AIR COMP.
% CIA-C-18 ;
)
- RUN h 80-8 4 WF h P
CONT. ON OSHT.556-1 O N SHT 556- 4 a c OWG, NO. N"E20 SHEET 556-2 CONTAINNENT INSTRUNENT AIR COMPRESSOR CIA-C-15 - CONTROL LOGIC DIAGRAM WASHINGTON PUBLIC POWER SUPPLY SYSTEM NUCLEAR PROJECT NO. 2 CONTROL LOGIC DIAGRAM - CIAS 7 f_ ' 20b
LOCAL
, LUBE
( DIL PRESS CIA-PS-2A d C0488 PRESS LOW LOCAL COMPRESSOR A^ v ~# DUTLET AIR CIA-PS-1A PRESS HIGH
~ LOCAL y C0487 COMPRESSOR RANGE UUTLET AIR 0-500 F CI A-TIS- 2 A C0490
( TEMP 350*F LOCAL h BD-A 4 CLG WTR DUT CI A-TIS- I A d C0330 ~ t TEtt* HIGH ! LOCAL t l j INTCLR WTR l CIA-LS-1A h BD-J j LEVEL LOW C0489 s
.}
AMENDMENT N0. 10 July 1980 i 1
-g TO SHT O 556-1 m TO SHT N 3 556-1 1 & C DWG. NO.
M-620 SHEET 556-3 iCIAS LOGIC DIAGRAM WASHINGTON DUBLIC POWER SUPPLY SYSTEM FIGURE CONTROL LOGIC DIAGRAM - CIAS 7.3-NUCLEAR PROJECT N0, 2 > 20e.
LOCAL - f LUBE
, DIL PRESS
' CI A-PS- 2B
[ C0492 PRESS LON LOCAL COMPRESSCR OUTLET AIR CIA-PS-IB PRESS HIGH LOCAL k C0491 COMPRESSOR RANGE OUTLET AIR 0-500*F CIA-TIS-28 C0494 TEMP 350 F LOCAL h BD-B l l CLG HTR DUT CIA-TIS-18 d C0331 TEMP HIGH LOCAL l INTCLR HTR H CIA-LS-1B ! LEVEL LON y C0493 3 (d v 1 l t .
AMENDMENT NO. 10 July 1980 hBD-B g TO SHT Q 556-2 a TG SHT v s 556-2 hBD-B a c DWG NO. M-620 SHEET 556-4 CIAS LOGIC DIAGRAM , WASHINGTON PUBLIC POWER SUPPLY SYSTEM FIGURE CONTROL LOGIC DIAGRAM - CIAS 7.3-NUCLEAR PROJECT N0. 2 20d l
~
NOTE IR-67 NHEN A NITROGEN BOTTLE EHPTIES ITS CONTENTeTHE LOW PRESSURE 18 STEP
- SIGNAL ENERGI2ES THE HOTOR. STEPPING ADVANCES THE PROGRAMMER TO PROGRAMMER THE NEXT STEP IN SEQUENCE' CIA-PROGR-THUS CAUSES AN OUTPUT IA HNICH DEENERGIZES (OPENS) THE SOLENDID VALVE AT THIS NEW STEP.
ALL OTHER SOLENDID VALVES BEFORE TYPICE FM l AND AFTER THIS NEW STEP SHALL BE ALL SOL. IN ENERGIZED (CLOSED) POSITION. VALVES LOCAL hBD-A BD-A - SOLENDID j RMS p LOCAL VALVE CIA-V-39A > CIA-V-39A CIA-SPV-IA DPEN - -- OPEN(ENERG) g DEENERGIZE AUTO CLOSE C W EN)
- T- m CLOSE -
Jh- (DEENERG) 2 C [-ENERG(CLOSE) 10 CIA-SPV-2A IR-71 3 0 CIA- M -3A I CIA-PS-39A PRESS.< 140
-[- 4 O
PSIG. MS h l 3 MLN, 6 O
o CONTAIN.1SD. SIGNALS 8 E E F --+ R - h~ ~ g g v "
+ -
o
, S A4 +-
89 10 r-9, o "5 1R-67 r-m -n CIA-PS-21A 11 x o 2g PRESS.<150 Rm
~ w PSIG 12 z m a Pg m
hED-A g3 r-
- o
, STEPPING PROGRAMMER 3 MOTOR DRIVE 14 y POWER SUPPLY 0 3 15
-4 O <
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T ' ** , f l l NOTE am-sa i NHEN A NITROGEN BOTTLE EHPTIES IT5 CONTENTeTHE LON PRESSURE 20 STEP SIGNAL ENERGIZE 5 THE HOTORe ADVANCES THE PROGRAHMER TO THE NEXT STEP IN SEQUENCC' CIA-PROGR-THUS CAUSES AN DUTPUT 15 NHICH DEENERGIZE5 (OPENS) THE SOLENDID VALVE AT THIS NEN STEP. ALL DTHER SOLENOID VALVES BEFORE TYPICE FM AND AFTER THIS NEN STEP SHALL BE ALL SOL. IN ENERGIZED (CLDSED) POSITION. VALVES LOCAL h50-5 SOLENDID 50-5 - RMS CIA-V-3SB O h LOCAL CIA-V-395 VALVE CIA-SPV-IB
]
DPEN - -- DPENCENERG) g DNME C (DPEN) AUTO r' m CLOSE CLOSE - (DEENERG) ENERG(CLOSE) f,j) -- 2 C TO CIA-SPV-25
~
3,, _ h50-5 C TO CIA-SPV-35 4 PRESS.< I40 .O PSIG. td5 o 3 M: N 6 0
g r, = z "i = . " _ p * ;8 5 5 8 5 5 . 5 7 8 9 0 1 1 1 1 2 V V V V v - P P P P P S S S S S - A A A A A I I I I I . C C C C C O O O O O T T T T T 4 5 5 7 8 9 0 O. go e0 0o go I2g 3 1 g 1 O 1 O 1 O 1 C 1 C 1 C 1 C 2 C R l E
- "O M M
A
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. O O i, o O$ g- 2h=
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IL S .I AA P 5S TN - 5P NG A E OI I R Q2N 2* ~ $4- >3 CS C P Z^ M IQ g nG'N'3 o-h* r8 U o ~ g,2 EM5SS o5rM 3y w$q wdM* 3a= oE*% 52 SRg2 i aa JLi 2M;*=;g2 6 zP N 8m _ 1
(~ ( ( LOCAL CIA HDR. PRESS. CIA-PIS-29 0 PRESS, < 130 PSIG g' l l Q:'
AMENDMENT N0. 10 July 1980 l l i 9 i l i l l 3 i i y l & C DWG. NO. M-620 SHEET 556-7 CIAS LOGIC DIAGRAM WASHINGTON PtBLIC POWER SUPPLY SYSTEM FIGLNtE CONTROL LOGIC DIAGRAM - CIAS - 7.3-NUCLEAR PROJECT NO. 2 20g i l
l 3 FOR LOCATION SEE TABLE M.D. ! RMS ISQLATION j VA. i SPRING -l RETURN DPEN > OPEN TO NEUTRAL ? CLOSE f Ul CLOSE 4 I s 6 1
)
{ L 1
\
RMS S RMS IND LT, VALVE NO. REMARK LOC. BD-A CIA-V-20 DIV I BD-A CIA-V-30A DIY I . BD-B CIA-V-308 DIV !! . l hBD-S FOR LOCATION SEE TAILE MBD-S I A C DWG. NO. M-520 SHEET . 55 5- 8 CONTAINMENT ISDLATI(34 VALVE f CONTROL LOGIC DIAGRAM , AMENDMENT N0.10 July 1980 WASHINGTON PUBLIC POWER SUPPLY SYSTEM FIGURE
^ ~
~
NUCLEAR PROJECT NO. 2 20h
i l SH. TAG NO. LOCATION 1 CIA-dPIS-1A LCL FILTD CI AM PIS-1B LCL FILTEE CIA-dPIS-2A LCL FILTD SW. LOCATION CIAM PIS-28 LCL FILTD TAG NO.& [ COMPUTER I DESCRIPTION SET PDINT hDCkTION -(
AMENDMENT NO.10 July 1980 . DESCRIPTION SET POINT LOCAT!DN REMARK t CIA-F-1A DIFF. PRESS.HI 120"He0 INCR. BD-A C03it2 t CIA-F-1B DIFF. PRESS.HI 120"Hs0 INCR. BD-B CO323 HICIA-F-2A DIFF. PRESS.HI 120"Ha0 INCR, BD-A CO320 b CIA-F-3A DIFF. PRESS.HI 120"H2O INCR, BD-B CO32.1 i I & C DWG. NO. M-520
\
l SHEET . 5 5 5-9 ALARM ANNUNCIATOR & COMPUTER INPUT CONTROL LOGIC DIAGRAM ; WASHINGTON PUBLIC POWER SUPPLY SYSTEM FIGURE N A - A f NUCLEAR PROJECT NO. 2 7.37 201
i k i ~ Sw. TAG NO. LOCATION CIA-PS-3 IR-71 CIA SW. LOCATION TAG NO.& ( DESCRIPT10N td S ANN. SET POINT OCATION 2 MIN l k 0
AMENDMENT NO.10 July 1980 DESCRIPTION SET POINT LO T ON REMARK AIR RECEIVER PRESSURE 140 PSIG DECR. BD-A 1 l 1 l l I&CDWG.NO. M-620 SHERT 556-10 ALARM ANNUNCIATOR INPUT CONTROL LOGIC DIAGRAM 3 WASHINGTON PUBLIC POWER SUPPLY SYSTEM FIGURE CONTROL LOGIC DIAGRAM - CIAS 7,3_ NUCLEAR PROJECT NO. 2 l 20i
SHEET 554-3 SHEET 554-3 LCL ISO VALVE -
, BD-KI 5-- - --*-
--> CAC-V-6 l
I ISO VA RMS L _ . , _ _ _ , CAC-V-6 I I _ , OPEN l ( y _ g ..j l l : CLOSE I AUTO > , o CLOSE - I - J l La ISC VALVE - BD-KI ---> - CAC-V-2 ISO VA RMS : ~~~~ CAC-V-2 y_ l I
, OPEN CL E OPEN -- -
I AUTO > ' l CLOSE - M i y I LCL ISO VALVE b - ->- --> BD-KI CAC-V-8 l & ISO VA RMS ____, CAC-V-8 I I
, OPEN e
open
~
.) I I
I i CLOSE AUTO ; -- CLOSE - - O I y l La L _ _, ISO VALVE - BD-KI _ CAC-V-4 ISO VA RMS : ____, CAC-V-4 I , OPEN
-5 EN OPEN AUTO m l n CLGSE -
O
. J
\
l l h l
NOTE - ISO VALVES ARE gBD-S 125 VDC OPERATED CAC-FCV- 1 A AMENDMENT N0.10 July 1980 BD-S
] BD-K I (FEN E a- E @ BD-S CAC-FCV- 2 A BD-KI (FEN 00~8 E a= E @ W-S CAC-FCV- 3 A BD,-K I OPEN BD-S E aosc M @ W-5 CAC-FCV- 4A BD-S
] BD-KI OPEN E a= E ia cOWG.NO. M-62O SHEET 554-1 1 i Hq RECOMBINER CAC-HR-IA CENTROL LOGIC DIAGRAH l WASHINGTON PUBLIC POWER SUPPLY SYSTEM CONTROL LOGIC DIAGRAM FIGURE CONTAINMENT ATMOSPHERE CONTROL SYSTEM 7.3-NUCLEAR PROJECT NO. 2 21a l__-__________ , -- -- --- -- - -
- - - - - - - - - - - ~ ~ ' - - ~ ~ ~ ~ ~ ~ ~
; ( SHEET 554-4 SHEET 554-4 W - -> ISO VALVE -
BD-KII - ( I I w
--> CAC-V-15 ISO VA RMS ~~-~>
I CAC-V-15 I r--------__, I OPEN
-S I > CLOSE CPEN
^
I AUTO - CLOSE , l ; w I I lJ l LCL ISO VALVE - BD-KII ---> -
--> CAC-V-11 ISO VA RMS ~~~~>
CAC-V-11 r-- -
!I , OPEN
-S ' > CLOSE OPEN I -
AUTO - m l CLOSE - y 7q \ l v v \
! LCL BD-K11 b - -> _
ISO VALVE CAC-V-17 I 3 ISO VA RMS -
~~~~>
I I OPEN CAC-V-17 ----_, OPEN -3r---I I CLOSE AUTO g- - I CLOSE - ;- l l J
! LCL BD-KII b-p -
ISO VALVE -
--> CAC-V-13 ISO VA RMS --
~~~~>
CAC-V-13 I OPEN _ ,r-----.----->
- CLOSE OPEN - - .8 AUTO p -
CLOSE - I w
- J l l t I
1 l
AMENDMENT N0. 10 July 1980 NOTE - 150 VALVES ARE BD-S 125 VDC OPERATED CAC-FCV-1B y BD-K II OPEN hBD-S l h CLOSE h 4 15)-S CAC-FCV- 2B 90-K I I OPEN h BD-S h CLOSE h BD,-S CAC-FCV- 3B g BD-K II OPEN BD-S h CLOSE h -@BD-S CAC-FCV- 48
~
BD-K II OPEN g am h @BD-S l a c owG. M. M-620 SHEET 554-2 Hq RECONBINER CAC-HR-IB CONTROL LOGIC DIAGRAM WASHINGTON PUBLIC POWER SUPPLY SYSTEM CONTROL LOGIC DIAGRAM FIGURE NUCLEAR PROJECT NO. 2 CONTAINMENT ATMOSPHERE CONTROL SYSTEM 7.3-21b
~
IHgREC0fiBINER ! ( l LOCAL PANEL l
} lBY CONTRACT 471
) BD-KI MASTER RELAY CAC-NR-1A LOCAL TEST FOR ALL ITEMS CONTROLLED PERMISSIVE FROM MAIN / LOCAL BOARD CAC-SS-2 A THRU THESE CONTACTS MAIN BD. KI O DEENERGIZED(MAIN BOARD)
LOCAL TEST O ENERG12ED(LOCAL TEST) l INSTR.PHR.FOR BD-KI BD-KI ON/DFF (24VDC & 20VAC) In INSTRUMENT HARM-UP INSTRU. A ^ MASTER RELAY CAC-MR-2A CAC-CR-SA HARM-UP DN BD-KI ALL LOCAL INSTRU. N . CAC-RMS-9 A ~~ THRU THESE CONTACTS R DN I~ O l y y DEENERGIZED(POWER ON) 0FF O ENERGIZED (POWER OFF) l l BD-K! l . (, GAS RECIRC. --> _ GAS RECIRC. SYSTEM PHASE % BLONER/PREHEATER CAC-RMS-10A .__, - - - - - - > ON START --
~~-> OFF AUTO l l OFF HeRECOMBINER REACHED OPERATING TEMP. 500 F BD-KI H q RECOMB. !
SYSTEM TO SHEET REST OF hq RECOMB. CAC-RMS-11A G(5/521-1 5 SYSTEM oN ____________, ou AUTO O OFF l TO SHEET ~~ ~~ 521-1
\
l b I
AMENDMENT NO.10 July 1980 i l BD-Ki HARM-UP BTRUMENT LIGHT TEST M -UP 8 CAC-PS-1A PUSH TO TEST = BD-KI C- t S-1 A A INSTRUMENT I m
- t. D 8 gWARM
.ON (RED ON) 15 NIN.
I g BD-K I l Q CO473 H 2 RECOMBINER i
- N TEFF. OK OPERATOR TO START
. TO SEET 554-1 gi BD-K I I a c DWG. NO.
M-620 l _ TO SEET MEET 554-3 554-1 Hg RECOMBINER CAC-HR-1A ! CONTROL LOGIC DIAGRAM a WASHINGTON PUBLIC POWER SUPPLY SYSTEM CONTROL LOGIC DIAGRAM FIGURE CONTAINMENT ATMOSPHERE CONTROL SYSTEM 7.3- l NUCLIAR PROJECT N0. 2 21c l ^' -' s w-m w w ww -- m - . - , , _ , - -
,__,,,__a
l lHq RECOMBINER l LOCAL PANEL (x I'BY CONTRACT #71
- BD-KII MASTER RELAY CAC-NR-1B I LOCAL TEST FOR ALL.ITCMS CONTROLLED 3 PERNISSIVE FROM MAIN / LOCAL BOARD CAC-SS-2B THRU THESE CONTACTS MAIN BD.KII : DEENERG12ED(NAIN BOARD)
LOCAL TEST
- ENERGIZED (LOCAL TEST)
INSTR.PHR.FOR BD-KII BD-KII ON/0FF (24V & 20VAC) , g O i INSTRU. CAC- 31-9 B NASTER RELAY CAC-PR-28 NARM-UP DN BD-K I I A LOCAL INSTRU. N . CAC-RMS-SS -- THRU THESE CONTACTS @ F-
- DEENERGIZED(P0HER ON) -
l ON , g 0FF 5 ENERGI2ED(P0HER OFF1 g y l BD-KII l GAS RECIRC. -> _ . GAS RECIRC.SYSTEN PHASE y BLOWER /PREHEATER l CAC-MMS-10B _, ----* ON - START -- ---*
- OFF -
AUTO I ll OFF He RECOMBINER REACHED - l DPERATING TEMP. 500 F BD-KII H 2 RECDPE. ! ! SYSTEH TO SHEET REST OF Hq RECOMB. CAC-RMS-118 X 7 521-1 SYSTEM i DN - - - - - - - - - - - - - > ON AUTO 5 0FF - g _ 0FF 8 8 AxTO SHEET ! -- ~~
! i t y s21-1 i t
I. g
)
'k i .
h
AMENDMENT NO. 10 July 1980 I e I i ID-KII NARM-lF GSTRurENT LIGHT TEST HARM-UP e CAC-PB-IB TI M PUSH TO TEST 5 BD-K I I W-$-IB j
& INSTRUNENT tD HARM ON (RED ON) 15 MIN.
. I l C0474
- EiD-KII 1 4 Hg 9IECtHBIPER
.J V TDF. OK 5 (FCRATER TO START
. .. TO 5 TEE"Y 554-2 llD-KI1 I&CDWG.W. M-620 To SHEET
'~
SHEET 554-4 H g RECDPEBIPER CAC-tR-IS CENTREL LOGIC DIAINtAM WASHINGTON PUBLIC POWER SUPPLY S'/ STEM CONTROL LOGIC DIAGRAM FIGURE CONTAINMENT ATMOSPHERE CONTROL SYSTEM 7.3-NUCLEAR PROJECT NO. 2 21d l
WNP-2 AMENDMENT NO. 10 July 1980 A
)
7.4 SYSTEMS REQUIRED FOR SAFE SHUTDOWN 7.
4.1 DESCRIPTION
This section discusses the instrumentation and controls of the following systems required for safe plant shutdown:
- a. Reactor Core Isolation Cooling (RCIC) System
- b. Standby Liquid Control System (SLCS)
- c. RHR Shutdown Cooling Mode (RSCM)
- d. Remote Shutdown Gystem (RSS)
The sources which supply power to the safe shutdown systems originate from on-site AC and/or DC safety-related busses. Refer to Chapter 8 for a complete discussion of the safety-related power sources. 7.4.1.1 Reactor Core Isolation Cooling (RCIC) System
- a. RCIC System Function
) The reactor core isolation cooling system (see 5.4.6.2) instrumentation is designed to maintain or supplement reactor vessel water inventory during the following conditions:
- 1. Normal Operation. When the reactor vessel is isolated from its primary heat sink (the main condenser) and maintained in the hot standby condition;
- 2. Normal Operation. When the reactor vessel is isolated and accompanied by a loss of normal coolant flow from the reactor feedwater system;
- 3. When the plant is being shutdown and normal coolant flow from the feedwater system is-started before the reactor is depressurized to a level where the reactor shutdown cooling mode of the RHR system can be placed into operation.
- 4. When required as a backup to the High Pressure Core Spray System to mitigate the consequences of the rod drop accident by S automatically supplying cooling water to the
() reactor if vessel low water level is sensed. 7.4-1
WNP-2 AMENDMENT NO. 10 July 1980
- b. RCIC System Operation}}