ML21133A091

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0 to Updated Final Safety Analysis Report, Chapter 7, Section 7.7, Control Systems Not Required for Safety
ML21133A091
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Issue date: 04/29/2021
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LGS UFSAR 7.7 CONTROL SYSTEMS NOT REQUIRED FOR SAFETY 7.

7.1 DESCRIPTION

This section discusses instrumentation and controls of systems whose functions are not essential to the safety of the plant. The systems include:

a. Reactor vessel instrumentation
b. Reactor manual control system
c. Recirculation flow control system
d. Feedwater control system
e. Pressure regulator and turbine-generator system
f. Neutron monitoring system - instrumentation and controls
1. Source range monitor system
2. Rod block monitor system
3. Traversing incore probe system
g. DELETED
h. Reactor water cleanup system
i. Process radiation monitoring systems
1. South stack effluent radiation monitoring system
2. Radwaste equipment rooms ventilation exhaust radiation monitoring system
3. Charcoal treatment system process exhaust radiation monitoring system
4. Recombiner rooms and hydrogen analyzer compartments exhaust radiation monitoring system
5. Steam exhauster discharge and vacuum pump exhaust radiation monitoring system
6. Radwaste enclosure and chem. lab ventilation exhaust radiation monitoring system
7. Air ejector offgas effluent radiation monitoring system
8. Primary containment leak detection radiation monitoring system CHAPTER 07 7.7-1 REV. 20, SEPTEMBER 2020

LGS UFSAR

9. Hot maintenance shop ventilation exhaust radiation monitoring system
10. Liquid radwaste discharge radiation monitoring system
11. Service water radiation monitoring system
12. Reactor enclosure cooling water radiation monitoring system
j. Area radiation monitoring system
k. Gaseous radwaste system
l. Liquid radwaste system
m. Solid radwaste system
n. Fuel pool cooling and cleanup system
o. Refueling interlocks
p. Leak detection system
q. Emergency Response Facility Data System
r. Containment instrument gas system
s. Fire protection and suppression system
t. Nonsafety-related equipment area cooling ventilation systems
u. Rod worth minimizer
v. Plant monitoring system
w. Hydrogen water chemistry system 7.7.1.1 Reactor Vessel Instrumentation Drawings M-41, M-43 and M-44 show the instrument numbers, arrangements of sensors, and sensing equipment used to monitor the reactor vessel conditions. Because the reactor vessel sensors used for safety-related systems are described and evaluated in other portions of this document, only the sensors not required for safety are described in this section.

7.7.1.1.1 RVI Identification 7.7.1.1.1.1 RVI General The purpose of the reactor vessel instrumentation is to monitor the key reactor vessel operating variables during plant operation.

CHAPTER 07 7.7-2 REV. 20, SEPTEMBER 2020

LGS UFSAR These instruments and systems are used to provide the operator with information during normal plant operation, startup, and shutdown. They are monitoring devices and provide no active power control or safety functions.

7.7.1.1.1.2 RVI Classification The systems and instruments discussed in this section are designed to operate under normal and peak operating conditions of system and ambient pressures and temperatures, and are classified as not related to safety.

7.7.1.1.1.3 RVI Reference Design Table 7.1-2 lists the reference design information. The RVI is an operational system and has no safety function. Therefore, there are no safety design differences between this system and those of the referenced design facilities. This system is functionally identical to the referenced system, except for RPV level instrumentation used during refueling.

7.7.1.1.2 RVI Power Sources Nonsafeguard instrumentation is powered from a non-Class 1E 120 V ac 60 Hz instrument bus.

Safeguard pressure, differential pressure, and level transmitter and trip unit channels are powered from divisional 24 V dc buses that are energized by Class 1E 120 V ac / 24 V dc power supplies.

Some of these power supplies are energized from the appropriate Class 1E 120 V ac divisional instrument bus, and the others are energized from the Class 1E 125 V dc instrument bus for the division via a 125 V dc/120 V ac inverter. See Sections 7.2, 7.3, and 7.4 for more discussion of the Class 1E safeguard (divisional) power sources.

7.7.1.1.3 RVI Equipment Design For safeguard and nonsafeguard sensing instruments located below the process tap, the sensing lines slope downward from the process tap to the instrument about 1/2 in/ft (the design minimum is 1/4 in/ft, which is allowed for those lines that cannot be sloped a minimum 1/2 in/ft because of obstructions) so that air traps are not formed.

Where it is impractical to locate the instruments below the process tap, the sensing lines ascend vertically for at least 36 inches of process connection up to a high point vent located at an accessible elevation above the instrument, before sloping a minimum of 1/4 in/ft to the instrument.

The purpose of this is to prevent entrapment of a noncondensable gas in the sensing line.

7.7.1.1.3.1 RVI Circuit Description 7.7.1.1.3.1.1 RVI Jet Pump Flow Chapter 5 gives a description of the reactor jet pump flow controls.

7.7.1.1.3.1.2 RVI Temperature The reactor vessel temperature is indicated by several thermocouples placed at selected locations on the vessel shell and flange, as well as one on the vessel bottom head drain line as shown in CHAPTER 07 7.7-3 REV. 20, SEPTEMBER 2020

LGS UFSAR drawing M-41. These thermocouple temperatures are recorded at a local panel in the reactor enclosure and some selected thermocouples are input into the plant and ERFDS computers.

The reactor vessel temperature can also be determined on the basis of reactor coolant temperature. Temperatures needed for operation and compliance with the Technical Specification operating limits are obtained from one of several sources, depending on the operating condition.

During normal operation, either reactor pressure (Section 7.7.1.1.3.1.5) or the inlet temperature of the coolant in the recirculation loops (drawing M-43) can be used to determine the vessel temperature. Below the operating span of the resistance temperature detectors in the recirculation loop, and above 212F, the vessel pressure is used to determine the temperature. Below 212F, the vessel coolant, and thus the vessel temperature, is shown by the RWCU system inlet temperature (drawing M-44). These three sources of input are available from the process computer and from indicators or recorders in the control room. During normal operation, vessel thermal transients are limited by operational constraints on parameters other than temperature.

7.7.1.1.3.1.3 RVI Water Level Figure 7.7-1 shows the water level ranges and the vessel penetration for each range. The instruments that sense water level are differential pressure devices calibrated to be accurate at a specific vessel pressure and liquid temperature condition. Additional gauge pressure devices are used during refueling. Water level instrument line failure is addressed in Section 7.7.1.1.6. The following is a description of each water level range shown in Figure 7.7-1:

a. Shutdown water level range:

This range is used to monitor the reactor water level during the shutdown condition when the reactor system is flooded for maintenance and head removal. The water level measurement design is of the condensate reference chamber-type when the vessel head is in place. The vessel temperature and pressure conditions used for the calibration are 0 psig and 120F water in the vessel. The zero of the instrument is the bottom of the dryer skirt. The two vessel instrument penetrations used for this water level measurement are located at the top of the RPV head and just below the bottom of the dryer skirt. When the vessel head is removed for refueling, the condensate reference chamber and associated wet leg is not functional, and a gauge pressure transmitter is switched in, to measure shutdown range level. One indicator is provided in the control room to indicate reactor water level in the shutdown range.

b. Upset water level range:

This range is used to monitor the reactor water when the level of the water goes off the narrow range scale on the high side. The upset range is also used during refueling to provide redundant level indication in the control room. The design and vessel taps are the same as outlined above. The vessel pressure and temperature condition for accurate indication are at the normal operating point, except for the gauge pressure transmitter used during refueling, which is calibrated to the same conditions as for the shutdown range instrumentation. The upset water level is continuously indicated by a recorder in the control room. The upset range recorder and narrow range indicators are located in close proximity to each other. The upset range upper limit is higher than the narrow range upper limit. Therefore, when the CHAPTER 07 7.7-4 REV. 20, SEPTEMBER 2020

LGS UFSAR indication goes off scale in the upscale direction on the narrow range indicator, water level indication can be read immediately from the upset range recorder.

c. Narrow water level range:

This range uses RPV taps at an elevation near the top of the dryer skirt and taps at an elevation near the bottom of the dryer skirt. The zero of the instrument is the bottom of the dryer skirt and the instruments are calibrated to be accurate at the normal operating point. The water level measurement design is of the condensate reference chamber-type and uses differential pressure devices as the primary elements. The feedwater control system uses this range for its water level control and indication inputs. For more information about the range, trip points, number of channels, and control room indication, see the discussion on the feedwater control system in Section 7.7.1.4.

d. Wide water level range:

This range uses RPV taps at an elevation near the top of the dryer skirt and taps at an elevation near the top of the active fuel. The zero of the instrument is the bottom of the dryer skirt and the instruments are calibrated to be accurate at the normal power operating point. The water level measurement design is of the condensate reference chamber-type, and uses differential pressure devices as the primary elements. Control room indication in the wide range is used for normal and accident conditions. See Section 7.5 for a description of the number of channels, type of indication, and ranges.

e. Fuel zone water level range:

This range uses RPV taps at an elevation near the top of the dryer skirt and taps at the jet pump diffuser skirt. The zero of the instrument is the bottom of the dryer skirt and the instruments are calibrated to be accurate at 0 psig and saturated condition.

The fuel zone level transmitters output signals are electronically compensated for variation in reactor water and steam density with respect to pressure. The water level design is of the condensate reference chamber-type, and uses differential pressure devices as the primary elements. One recorder and one indicator are provided in the control room to indicate reactor water level in the fuel zone range.

The top of the active fuel will be indicated in red.

The condensate reference chamber for the narrow range, wide range, and fuel zone water level range is common, as shown in drawing M-42. A continuous backfill system is connected to each condensing chamber reference leg. The backfill system provides a continuous flow of water from the Control Rod Drive (CRD) System to the reference leg. This flow of water will continuously purge the reference leg and preclude the build up of noncondensable gas in the reference leg.

7.7.1.1.3.1.4 RVI Reactor Core Hydraulics A differential pressure transmitter indicates the core plate pressure drop by measuring the pressure difference between the core inlet plenum and the space just above the core support assembly (core plate and fuel supports). The instrument sensing line used to determine differential pressure is a pipe within a pipe arrangement attached to the reactor vessel tap initially designed to provide differential pressure taps and inject the liquid from the SLCS. Now SLCS is injected through the B Core Spray sparger and the differential pressure sensing line only taps the pressure above and below the core support assembly. The differential pressure of the core plate is recorded in the control room.

CHAPTER 07 7.7-5 REV. 20, SEPTEMBER 2020

LGS UFSAR 7.7.1.1.3.1.5 RVI Pressure Pressure sensors/transducers, indicators, and transmitters detect reactor vessel internal pressure from the same instrument lines used to measure reactor vessel water level. The only nonsafety-related pressure measuring instruments provided are discussed in Section 7.7.1.4.

7.7.1.1.3.1.6 RVI Head Seal Leak Detection Pressure between the inner and outer reactor vessel head seal ring is sensed by a local pressure indicator and a pressure switch that alarms in the control room. If the inner seal fails, the pressure at the pressure sensors increases to the vessel pressure. The plant continues to operate with the outer seal as a backup, and the inner seal can be repaired at the next outage when the head is removed. If both the inner and outer head seals fail, the leak is detected by an increase in drywell temperature and pressure.

7.7.1.1.3.1.7 RVI/SRV Leak Detection Thermocouples are located in the discharge exhaust pipe of the SRV. The temperature signal goes to a multipoint recorder with an alarm and is activated by any temperature in excess of a set temperature, signaling that one of the SRV seals or pilot valves has started to leak (drawing M-41).

The temperature signals can also be used as a means to determine SRV position, hence this instrumentation is a diverse backup to the SRVPI system (Section 7.6.1.5).

7.7.1.1.3.1.8 RVI Other Instruments

a. Steam temperature is measured and transmitted to the control room.
b. Feedwater temperature is measured and transmitted to the control room.
c. Feedwater conductivity is monitored and recorded locally. An instrument trouble alarm is provided in the control room (drawing M-23).

7.7.1.1.3.2 RVI Testability There are no testability requirements for the RVI described in this section because the instruments are nonsafety-related and are not required for any postulated DBA or for safe shutdown. The instruments located outside the drywell can be tested during plant operation, while those that are inaccessible during plant operation can be tested during plant shutdown. Plant test procedures will establish methodology and frequency.

7.7.1.1.4 RVI Environmental Considerations There are no special environmental considerations for the instruments described in this section.

7.7.1.1.5 RVI Operational Considerations 7.7.1.1.5.1 RVI General Information The RVI discussed in this section is designed to augment the existing information from the safety-related systems so that the operator can start up, operate at power, shut down, and service CHAPTER 07 7.7-6 REV. 20, SEPTEMBER 2020

LGS UFSAR the reactor vessel in an efficient manner. None of this instrumentation is required to initiate any safety-related system.

7.7.1.1.5.2 RVI Operator Information The following is the information that the operator has at his disposal from the instrumentation discussed in this section:

a. The temperature of the reactor coolant in the recirculation loops is recorded in the control room.
b. The RWCU inlet temperature is indicated in the control room.
c. Water level in the shutdown range is indicated in the control room.
d. Water level in the upset range is recorded in the control room.
e. Water level in the fuel zone range is indicated and recorded in the control room.
f. The core plate differential pressure and total core flow are recorded in the control room.
g. The reactor pressure is indicated and recorded in the control room.
h. The reactor head seal leak detection turns on an annunciator when the inner reactor head seal fails.
i. The discharge temperatures of all the SRVs are shown on a multipoint recorder in the control room. Any temperature point that exceeds the trip setting turns on an annunciator indicating that a SRV seal or pilot valve has started to leak.
j. Feedwater conductivity is recorded in the control room. The recorder turns on an annunciator in the control room on a high signal.

7.7.1.1.5.3 RVI Setpoints The annunciator alarm setpoints for the reactor head seal leak detection, and SRV seal and pilot valve leak detection are set so that the sensitivity to the variable being measured provides adequate information.

Figure 7.7-1 includes a chart showing the relative indicated water levels at which various alarms are initiated.

Level trips to initiate various alarms and trip the main turbine and the feed pumps are discussed in Section 7.7.1.4.

7.7.1.1.6 Water Level Instrument Line Failure CHAPTER 07 7.7-7 REV. 20, SEPTEMBER 2020

LGS UFSAR An analysis has been performed to verify that there is sufficient redundancy in the water level instrumentation to prevent a sensing line failure (i.e., break, blockage or leak) concurrent with a random single electrical failure from defeating an automatic RPS or ESF actuation. This analysis was conducted based on the methodology provided in Table 7.7-6.

The instrument reference lines common to feedwater control and to protective system sensors are identified in Table 7.7-7. An evaluation was performed to determine the consequences of failures in such reference lines concurrent with additional single failures in protective channels not dependent on the failed sensing line.

In the highly unlikely scenario, the most severe reference line was assumed to fail such that all attached level instruments erroneously indicated high levels. Then, additional worse case single failures were postulated in the circuits connected to the remaining reference line. The criteria for selection of the potential worst case combinations of reference line failure plus additional single failure was to determine those combinations, if any, which preclude automatic operation of a RPS and/or ESF system(s), and which may require manual action by the operator to bring the reactor to a safe condition. Worst case single division power supply loss was considered for ECCS and RCIC, but this is independent from other single failures which could affect RPS or MSIV closure, etc. (i.e., a power bus failure in RPS would fail "safe" causing a trip of that channel).

A review of the various failure combinations resulted in the identification of two cases which would represent the worst postulated failure paths. These are described as follows:

Case 1 was found to be failure of the Division 2 instrument reference line (i.e., connected to condensing chamber B21-D004B) combined with a failure "high" of the B21-N091D or H level transmitter. The worst case also assumes that the FCS level soft majority selector fails to perform its designed function of removing the failed transmitter signal from the level control signal. The feedwater controller responds to the high level error signal by reducing the feedwater flow.

Following the loss of feedwater, water level will decrease to level 4, initiating a low water level alarm. Reactor scram will be initiated when water level decreases to the low water level 3 scram setpoint. The low water level 3 signal will result in the initiation of a second low water level alarm.

As water level decreases to low water level 2, a third low water level alarm will initiate, the RCIC system will automatically start, and both recirculation pumps will trip. The HPCI system is unavailable due to the assumed failure. This scenario is identical to the sizing basis of the RCIC.

The RCIC will provide sufficient water inventory makeup to prevent the water level from dropping below low water level 1, thus avoiding initiation of the ADS and the low pressure ECCS systems.

The core remains covered at all times and no fuel failure would occur.

All other scenarios identified were less severe than the two cases just described. Those that involve failures which do not affect the level 3 scram are accompanied by either HPCI or HPCI +

RCIC automatic initiation by low reactor water level 2 signals. Other scenarios which involve failures in the level 3 scram circuits are accompanied by scram due to ARI at level 2 and automatic initiation of both HPCI and RCIC.

The results of this assessment can be summarized as follows:

a. The LGS reactor system can withstand any reactor vessel level reference line break coupled with an additional worst single failure in a protective channel not dependent on the failed sensing line without compromising safety. This is assured by the following evaluations:

CHAPTER 07 7.7-8 REV. 20, SEPTEMBER 2020

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1. No part of the active fuel is uncovered at any time. This assures no fuel damage and no degradation of the CPR or reactivity release.
2. Both the vessel and the containment remain structurally sound throughout the postulated event. This provides secondary assurance that no reactivity can be released to the public.
3. The scenario postulated is a highly unlikely event (instrument line breakage with coincident random failure) and compounds it with worst case conditions through the event. Though no credit is taken in the scenarios identified, it is highly probable that the operator would recover feedwater immediately by switching the controller to the alternate instrument line because of the alarms that call his attention to level indication mismatch and the numerous low water level alarms. In addition, if the failed instrument reference line is either the Division 3 or 4 instrument line (i.e., connected to condensing chamber B21-D004C or D) no feedwater controller error would occur and there would be no reduction in feedwater flow.

LGS has the capability of providing automatic protection under all postulated scenarios. This is due to its four instrumentation divisions, which provide sufficient redundancy for automatic initiation of protection systems and eliminates the need for manual corrective action by the operator to prevent compromising plant safety.

b. There are no failure combinations that result in failure of both HPCI and RCIC. One of the two systems is always available.
c. Failure combinations which result in failure of the reactor to scram at low reactor water level 3 are always accompanied by reactor scram (due to ARI) and either HPCI or RCIC automatic initiation due to low reactor water level 2 signals.

It is concluded from this assessment of a break in a vessel level sensing line common to control and protective systems plus an additional worst single failure in a protective channel not dependent on the failed sensing line that the resulting accident is less severe and bounded by the DBAs already analyzed in Chapter 15.

7.7.1.2 Reactor Manual Control System - Instrumentation and Controls 7.7.1.2.1 RMCS Identification 7.7.1.2.1.1 RMCS General The objective of the RMCS is to provide the operator with the means to make changes in nuclear reactivity so that reactor power level and power distribution can be controlled. The system allows the operator to manipulate control rods.

The RMCS instrumentation and controls consist of the electrical circuitry, switches, indicators, and alarm devices provided for operational manipulation of the control rods and surveillance of associated equipment.

CHAPTER 07 7.7-9 REV. 20, SEPTEMBER 2020

LGS UFSAR This system includes the interlocks that inhibit rod movement (rod block) under certain conditions.

The RMCS does not include any of the circuitry or devices used to automatically or manually scram the reactor (Section 7.2).

In addition, the mechanical devices of the CRDs and the CRD hydraulic system are not included in the RMCS. The latter mechanical components are described in Section 4.1.3.

7.7.1.2.1.2 RMCS Classification This system is a power generation system and is classified as not related to safety.

7.7.1.2.1.3 RMCS Reference Design Table 7.1-2 lists reference design information. The RMCS is an operational system and has no safety function. Therefore, there are no safety design differences between this system and those of the reference design facilities. This system is functionally identical to the referenced system.

7.7.1.2.2 RMCS Power Sources

a. Normal: The reactor manual control system receives its power from the 120 V ac instrumentation buses. Each of these buses receives its normal power supply from the appropriate 440 V ac standby power system.
b. Alternate: On loss of normal auxiliary power, the station diesel generators provide backup power to the 440 V standby ac power systems.

7.7.1.2.3 RMCS Equipment Design 7.7.1.2.3.1 RMCS General The following discussions examine the control rod movement instrumentation and control aspects of the subject system, and the control rod position information system. The control descriptions include the following:

a. Control rod drive - control system
b. Control rod drive - hydraulic system
c. Rod block interlocks The position descriptions include the following:
a. Rod position probes
b. Display electronics Drawings M-46 and M-47 show the CRD hydraulic system. Drawings C11-1030-F-008, C11-1030-F-009, C11-1030-F-010, C11-1030-F-011, C11-1030-F-012, C11-1030-F-013, and C11-1030-F-014 show the FCD for the CRD hydraulic system.

CHAPTER 07 7.7-10 REV. 20, SEPTEMBER 2020

LGS UFSAR Although they also show the arrangement of scram devices, these devices are not part of the RMCS. Control rods are moved by admitting water under pressure from the CRD water pump into the appropriate end of the CRD cylinder. The pressurized water forces the piston, which is attached by a connecting rod to the control rod, to move. Three modes of control rod operation are used: insert, withdraw, and settle. Four solenoid-operated valves are associated with each control rod to accomplish the actions required for the operational modes. The valves control the path that the CRD water takes to the cylinder.

7.7.1.2.3.2 Rod Movement Controls 7.7.1.2.3.2.1 CRD Control System 7.7.1.2.3.2.1.1 Introduction When the operator selects a control rod for motion and operates the rod insertion control switch, identical messages are formulated in the A and B portions of the rod drive control system as shown in Figure 7.7-3. A comparison test is made of these two messages, and if identical messages are confirmed, a serial message in the form of results is produced; then a serial message in the form of electrical pulses is transmitted to all HCUs. The message contains two portions: the identity or "address" of the selected HCU, and the operation data on the action to be executed. Only the selected HCU responds to this transmission and proceeds to execute the rod action commands.

Hence, the drive water and exhaust valves for the selected rod open and allow the CRD water to follow a path that results in the desired control rod movement.

On receipt of the transmitted signal as shown in Figure 7.7-3, the responding HCU transmits three portions of an acknowledge message back to the auxiliary equipment room panel for comparison with the original message:

a. Its own hard-wire identity "address"
b. Its own operations currently being executed
c. Status indications of valve positions, accumulator conditions, and test switch positions.

In either rod motion direction, the A and B messages are formulated and compared each millisecond; if they agree, the message is transmitted to the HCU selected by the operator.

Continued rod motion depends on receipt of a train of sequential messages because the HCU insert, withdraw, and settle valve control circuits are ac-coupled. The system must operate in a dynamic manner to effect rod motion. Postulated failures within the RMCS generally result in a static condition within the system, which prevents further rod motion.

Any disagreement between the A and B formulated messages or the C acknowledge message prevents further rod motion. However, electrical noise disruptions only momentarily affect the system in proportion to the duration of the offending source. Correct operation of the system resumes when the noise source ceases. In Figure 7.7-4, three action loops of the solid-state RMCS are depicted:

CHAPTER 07 7.7-11 REV. 20, SEPTEMBER 2020

LGS UFSAR

a. The high speed loop (a nominal 0.0002 second duration) services the control rod selected by the operator to transmit action commands and receive status indications.
b. The medium speed loop (a nominal 0.045 second duration) monitors the other control rods in the reactor, one at a time, to update their status display.
c. The low speed loop (on the order of 20 second to 100 second duration) exercises one HCU at a time to ensure correct execution of actions commanded. This provides for a continuous, periodic self-test of the entire RMCS.

The rod selection circuitry is arranged so that a rod selection is sustained until either another rod is selected, or separate action is taken to revert the selection circuitry to a no-rod-selected condition.

Initiating movement of the selected rod prevents the selection of any other rod until the movement cycle of the selected rod has been completed. Reversion to the no-rod-selected condition is not possible (except for loss of control circuitry power) until any moving rod has completed its movement cycle.

The direction in which the selected rod moves is determined by the position of four switches located on the reactor control panel. These four switches, INSERT, WITHDRAW, CONTINUOUS INSERT and CONTINUOUS WITHDRAW, are push buttons that return to an off position by spring action.

7.7.1.2.3.2.1.2 CRD Insert Cycle A description of the detailed operation of the RMCS during an insert cycle follows. The cycle is described in terms of the insert, withdraw, and settle commands emanating from the RMCS. The response of a selected rod to the various commands transmitted has been explained previously.

Drawings C11-1030-F-008, C11-1030-F-009, C11-1030-F-010, C11-1030-F-011, C11-1030-F-012, C11-1030-F-013, and C11-1030-F-014 can be used to follow the sequence of an insert cycle.

With a control rod selected for movement, depressing the INSERT switch and then releasing it energizes the insert command for a specific time. Just as the insert command is removed, the settle command is automatically energized and remains energized for a specific time. The insert command time setting and the rate of drive water flow provided by the CRD hydraulic system determine the distance traveled by a rod. The time setting results in a one notch (6 inch) insertion of the selected rod for each momentary application of a rod-in signal from the rod movement switch. Continuous insertion of a selected control rod is possible by holding the INSERT switch down, as long as rod movement is within pattern restraints.

A second switch can be used to effect insertion of a selected control rod. This switch is the CONTINUOUS INSERT switch. By holding this switch in, the unit maintains the insert command in a continuous, energized state to cause continuous insertion of the selected control rod. When released, the timers are no longer bypassed and normal insert and settle cycles are initiated to stop the drive.

7.7.1.2.3.2.1.3 CRD Withdraw Cycle A description of the detailed operation of the RMCS during a withdraw cycle follows. The cycle is described in terms of the insert, withdraw, and settle commands. The response of a selected rod to CHAPTER 07 7.7-12 REV. 20, SEPTEMBER 2020

LGS UFSAR the various commands transmitted has been explained previously. Drawings C11-1030-F-008, C11-1030-F-009, C11-1030-F-010, C11-1030-F-011, C11-1030-F-012, C11-1030-F-013, and C11-1030-F-014 can be used to follow the sequence of a withdrawal cycle.

With a control rod selected for movement, depressing the WITHDRAW switch energizes the insert valves for a short time. Energizing the insert valves at the beginning of the withdrawal cycle is necessary to allow the collet fingers to disengage the index tube. When the insert valves are de-energized, the withdraw and settle valves are energized for a controlled period of time. The withdraw valve is de-energized before movement is completed. The drive then settles until the collet fingers engage. The settle valve is then de-energized, completing the withdraw cycle . This withdrawal cycle is the same whether the withdrawal switch is held down continuously or momentarily depressed. The timers that control the withdrawal cycle are set so that the rod travels one notch (6-inches) per cycle. Provisions are included to prevent further control rod motion in the event of timer failure.

At high power levels when RWM is not operable, a selected control rod can be continuously withdrawn if the WITHDRAWAL switch is held in the depressed position at the same time that the CONTINUOUS WITHDRAWAL switch is held in the depressed position. With both switches held in these positions, the withdraw and settle commands are continuously energized. When released, the normal settle period completes the withdrawal cycle.

7.7.1.2.3.2.2 CRD Hydraulic System One motor-operated pressure control valve, two air-operated flow control valves, and four solenoid-operated stabilizer valves are included in the CRD hydraulic system to maintain smooth and regulated system operation. These devices are shown in drawing M-46. The motor-operated pressure control valve is positioned by manipulating a switch in the control room. The switch for this valve is located close to the pressure indicator, which responds to the pressure changes caused by the movement of the valve. Only one flow control valve is in service at a time. The air-operated flow control valves are automatically positioned in response to signals from an upstream flow-measuring device. Only two stabilizer valves are in service at a time. The stabilizer valves are automatically controlled by the energization of the insert and withdraw commands. The control scheme is shown in drawings C71-1010-F-002, C71-1010-F-003, C71-1010-F-004, and C71-1010-F-005. Two drive water pumps are controlled by switches in the control room. Each pump automatically trips on indication of low suction pressure.

7.7.1.2.3.2.3 Rod Block Interlocks 7.7.1.2.3.2.3.1 RBI Trip System A portion of the RMCS, upon receipt of input signals from other systems and subsystems, inhibits movement or selection of control rods.

7.7.1.2.3.2.3.2 RBI Grouping of Channels The same grouping of neutron monitoring system equipment (SRM, IRM, APRM, and RBM) that is used in the RPS is also used in the rod block circuitry (Section 7.6.1.4).

Half of the total monitors (SRM, IRM, APRM, and RBM) provides inputs to one of the RMCS rod block logic circuits and the remaining half provides inputs to the other RMCS rod block logic circuit.

Two APRM channels provide recirculation flow upscale rod blocks to one logic circuit; the other two CHAPTER 07 7.7-13 REV. 20, SEPTEMBER 2020

LGS UFSAR APRM channels provide recirculation flow upscale rod block signals to the other logic circuit. SDV high water level signals are provided as inputs into both of the rod block logic circuits. Both rod block logic circuits sense when the high water level scram trip for the SDV is bypassed.

The rod withdrawal block from the RWM trip prevents both notch insertion and continuous insertion. The rod insert block from the RWM functions prevents both notch insertion and continuous insertion.

The APRM rod block settings are varied as a function of recirculation flow and core thermal power, and the RBM rod block settings are varied as a function of core thermal power. Analyses show that the selected settings are sufficient to avoid both RPS action and local fuel damage as a result of a single control rod withdrawal error. Mechanical switches in the SRM and IRM detector drive systems provide the position signals used to indicate that a detector is not fully inserted. Additional detailed information on all the NMS trip channels is available in Section 7.6.1.4. The rod block from SDV high water level uses two nonindicating float switches installed on the scram discharge volume. Two additional float switches provide a control room annunciation of increasing level below the level at which a rod block occurs.

7.7.1.2.3.2.3.3 RBI Functions The following discussion describes the various rod block functions and explains the intent of each function. The instruments used to sense the conditions for which a rod block is provided are discussed later. The rod block functions provided specifically for refueling situations are described in Section 7.7.1.15.

a. With the mode switch in the SHUTDOWN position, no control rod can be withdrawn. This enforces compliance with the intent of the shutdown mode.
b. The circuitry is arranged to initiate a rod block regardless of the position of the mode switch for the following conditions:
1. Any APRM Simulated Thermal Power upscale rod block alarm - The purpose of this rod block function is to avoid conditions that would require RPS action if allowed to proceed. The APRM upscale rod block alarm setting is selected to initiate a rod block before the APRM high neutron flux or Simulated Thermal Power scram setting is reached.
2. Any APRM inoperative alarm - This ensures that no control rod is withdrawn unless the average power range neutron monitoring channels are either in service or correctly bypassed.
3. Any APRM LPRM low count alarm - This ensures that no control rod is withdrawn unless the average power range neutron monitoring channels have the required number of LPRM inputs to be considered operable.
4. SDV high water level - This ensures that no control rod is withdrawn unless enough capacity is available in the SDV to accommodate a scram. The setting is selected to initiate a rod block earlier than the scram that is initiated on SDV high water level.

CHAPTER 07 7.7-14 REV. 20, SEPTEMBER 2020

LGS UFSAR

5. SDV high water level scram trip bypassed - This ensures that no control rod is withdrawn while the SDV high water level scram function is out-of-service.
6. The RWM can initiate a rod insert and a rod withdrawal block. The purpose of these functions is to reinforce procedural controls that limit the reactivity worth of control rods under low power conditions. The rod block trip settings are based on the allowable control rod worth limits established for the design basis rod-drop accident. Adherence to prescribed control rod patterns is the normal method by which this reactivity restriction is observed.

Additional information on the RWM function is available in Section 7.7.1.21.

7. RPIS malfunction - This ensures that no control rod is withdrawn unless the RPIS is in service.
8. Either RBM upscale alarm - This function is provided to stop the erroneous withdrawal of a control rod so that local fuel damage does not result, although local fuel damage poses no significant threat in terms of radioactive material released from the nuclear system.
9. Either RBM inoperative alarm - This ensures that no control rod is withdrawn unless the RBM channels are in service or correctly bypassed.
c. With the reactor mode switch in the RUN position, any of the following conditions initiates a rod block:
1. Any APRM downscale alarm - This ensures that no control rod is withdrawn during power range operation unless the average power range neutron monitoring channels are operating correctly or are correctly bypassed. All unbypassed APRMs must be on scale during reactor operations in the RUN mode.
2. Either RBM downscale alarm - This ensures that no control rod is withdrawn during power range operation unless the RBM channels are operating correctly or are correctly bypassed. Unbypassed RBMs must be on scale during reactor operations in the RUN mode.
3. Any APRM recirculation flow upscale alarm - This ensures that the no control rod is withdrawn unless the APRM recirculation flow signals are operable and the flow rate in not unusually high.
d. With the mode switch in the STARTUP or REFUEL position, any of the following conditions initiates a rod block:
1. Any SRM detector not fully inserted into the core when the SRM count level is below the retract permit level and any IRM range switch is on either of the two lowest ranges - This ensures that no control rod is withdrawn unless all SRM detectors are correctly inserted, because they must be relied on to provide the operator with neutron flux level information.

CHAPTER 07 7.7-15 REV. 20, SEPTEMBER 2020

LGS UFSAR

2. An SRM upscale level alarm - This ensures that no control rod is withdrawn unless the SRM detectors are correctly retracted during a reactor startup.

The block setting is selected at the upper end of the range over which the SRM is designed to detect and measure neutron flux.

3. Any SRM downscale alarm - This ensures that no control rod is withdrawn unless the SRM count rate is above the minimum prescribed for low neutron flux level monitoring.
4. Any SRM inoperative alarm - This ensures that no control rod is withdrawn during low neutron flux level operations unless neutron monitoring capability is available because all SRM channels are in service or correctly bypassed.
5. Any IRM detector not fully inserted into the core -This ensures that no control rod is withdrawn during low neutron flux level operations unless proper neutron monitoring capability is available because all IRM detectors are correctly located.
6. Any IRM upscale alarm - This ensures that no control rod is withdrawn unless the intermediate range neutron monitoring equipment is correctly upranged during a reactor startup. This rod block also provides a means to stop rod withdrawal in time to avoid conditions requiring RPS action (scram) in the event that a rod withdrawal error is made during low neutron flux level operations.
7. Any IRM downscale alarm except when the range switch is on the lowest range - This ensures that no control rod is withdrawn during low neutron flux level operations unless the neutron flux is being correctly monitored. This rod block prevents the continuation of a reactor startup if the operator upranges the IRM too far for the existing flux level. Thus, the rod block ensures that the IRM is on scale if control rods are to be withdrawn.
8. Any IRM inoperative alarm - This ensures that no control rod is withdrawn during low neutron flux level operations unless neutron monitoring capability is available because all IRM channels are in service or are correctly bypassed.

7.7.1.2.3.2.3.4 RBI Bypasses To permit continued power operation during repair or calibration of equipment for selected functions that provide rod block interlocks, a limited number of manual bypasses are permitted as follows:

a. One SRM channel
b. Two IRM channels (one on either Bus A or Bus B)
c. One APRM channel
d. One RBM channel CHAPTER 07 7.7-16 REV. 20, SEPTEMBER 2020

LGS UFSAR The permissible IRM bypasses are arranged as two groups of equal numbers of channels. One manual bypass is allowed in each group. The groups are chosen so that adequate monitoring of the core is maintained with one channel bypassed in each group. The arrangement allows the bypassing of one IRM in each rod block logic circuit.

One of the four APRM channels can be bypassed at any time. The assignment of LPRMs to APRM channels is chosen so that adequate monitoring of the core is maintained with an APRM channel bypassed.

These bypasses are effected by positioning switches in the control room. A light in the control room indicates the bypassed condition.

An automatic bypass of the SRM detector position rod block is effected as the neutron flux increases beyond a preset low level on the SRM instrumentation. The bypass allows the detectors to be partially or completely withdrawn as a reactor startup is continued.

An automatic bypass of the RBM rod block occurs when the power level is below the preselected level or when a peripheral control rod is selected. Either condition indicates that local fuel damage is not threatened and that RBM action is not required.

The RWM system rod block functions are automatically bypassed when reactor power is above a preselected value in the power range. The RWM can be manually bypassed for maintenance at any time.

7.7.1.2.3.2.3.5 RBI Interface Tables 7.7-4 and 7.7-5 detail the rod block interlocks used in the RMCS, as well as the rod blocking functions that originate in the neutron monitoring system.

7.7.1.2.3.2.3.6 RBI Redundancy To achieve an operationally desirable performance objective where most failures of individual components would either be easily detectable or would not disable the rod movement inhibiting functions, the rod block logic circuitry is arranged as two redundant logic circuits. Each logic circuit receives input trip signals from a number of trip channels and each logic circuit can provide a separate rod block signal to inhibit rod withdrawal.

The output of each logic circuit is coupled to a comparator by electro-optical devices in the rod drive control cabinet. The formulated A and B signals are compared and rod blocks applied when either an A or B trip is present. Rod withdrawal is permitted only if the two signals agree at all times and no rod block signals are present. Because the transmitted signals are dynamic and vary with time, any RMCS failure that interrupts the dynamic signals transmitted to the HCUs prevents further control rod motion. Hence, failures consisting of short circuits, open circuits, loss of circuit continuity, or loss of power inhibit manual rod movement.

The rod block circuitry is effective in preventing rod withdrawal, if required, during both normal (notch) withdrawal and continuous withdrawal. If a rod block signal is received during a rod withdrawal, the control rod is automatically stopped at the next notch position, even during a continuous rod withdrawal.

CHAPTER 07 7.7-17 REV. 20, SEPTEMBER 2020

LGS UFSAR The components used to initiate rod blocks in combination with refueling operations provide rod block trip signals to these same rod block circuits. These refueling rod blocks are described in Section 7.7.1.15.

7.7.1.2.3.2.3.7 RBI On-Line Testability On-line testability of the systems and indication of bypassed or inoperable status of the system are provided.

7.7.1.2.3.2.3.8 RBI Environmental Considerations The equipment is mounted in the control structure. Environmental conditions in the control structure are listed in Section 3.11.

7.7.1.2.3.2.3.9 RBI Operational Considerations The rod block trips prevent an operator from withdrawing rods if the associated equipment is not capable of monitoring core response or, if unchecked, the withdrawals might require a RPS action (scram). There are no special operational considerations.

7.7.1.2.3.2.4 RBI Testability In addition to the periodic self-test mode of system operation, the RMCS circuitry can be routinely checked for correct operation by manipulating control rods using the various methods of control.

Detailed testing and calibration can be performed by using standard test and calibration procedures for the various components of the RMCS circuitry.

7.7.1.2.3.3 Rod Position Information System This system includes the rod position probes and the electronic hardware that processes the probe signals and provides the data described above.

7.7.1.2.3.3.1 RPIS Position Probes The position probe is a long cylindrical assembly that fits inside the CRD. It includes 53 magnetically operated reed switches located along the length of the probe and operated by a permanent magnet fixed to the moving part of the hydraulic drive mechanism. As the drive, and with it the control rod blade, moves along its length, the magnet causes reed switches to close as it passes over the switch locations. The particular closed switch then indicates where the CRD, and hence the rod itself, is positioned.

The switches are located as follows: one at each of 25 notch (even) positions; one at each of 24 midnotch (odd) positions; two at the fully inserted position (approximately the same location as the "00" notch); one at the fully withdrawn position (approximately the same location as the "48" notch);

and one at the "overtravel," or decoupled, position.

All of the midnotch, or odd, switches are wired in parallel and treated as one switch (for purposes of external connections), and the two fully inserted switches are wired in parallel and treated as one switch. These and the remaining switches are wired in a 5x6 array (the switches short the CHAPTER 07 7.7-18 REV. 20, SEPTEMBER 2020

LGS UFSAR intersections) and are routed out in an 11 wire cable to the processing electronics (the probe also includes a thermocouple that is wired out separately from the 5x6 array) (Figure 7.7-5).

7.7.1.2.3.3.2 RPIS Electronics The electronics consist of a set of "probe multiplexer cards" (one per four-rod group), a set of "file control cards" (one per 11-probe multiplexer cards), and one set of master control and processing cards serving the whole system. All probe multiplexer cards are the same except that each has a pair of plug-in "daughter cards" containing the identity code of one four-rod group (the probes for the corresponding four rods are connected to the probed multiplexer card).

7.7.1.2.3.3.3 RPIS Operation The system operates on a continuous scanning basis with a complete cycle every 45 milliseconds.

The operation is as follows: The control logic generates the identity code of one rod in the set, and transmits it using time multiplexing to all of the file control cards. These in turn transmit the identity with timing signals to all of the probe multiplexer cards. The one multiplexer card with the matching rod identity responds and transmits its identity (locally generated) plus the raw probe data for that rod back through the file control card to the master control and processing logic. The processing logic does several checks on the returning data. First, a check is made to verify that an answer was received. Next, the identity of the answering data is checked against that which was sent.

Finally, the format of the data is checked for legitimacy. Only a single even position, or full-in plus position "00," or full-out plus position "48," or odd, or overtravel, or blank [no switch closed] are legitimate. Any other combination of switches is flagged as a fault.)

If the data passes all of these tests, it is decoded and transmitted in multiplexed form to the displays in the main control panel and is then loaded into a memory to be read by the computer as required.

As soon as one rod's data are processed, the next rod's identity is generated, processed, and repeated for all of the rods. When data for all rods have been gathered, the cycle repeats.

7.7.1.2.4 RMCS Environmental Considerations The RMCS (control and position indication circuitry) is not required for any plant safety function nor is it required to operate in any associated design basis events or transient occurrences. The reactor manual control circuitry is required to operate only in the normal plant environments during normal power generation operations.

The CRD HCUs are located outside the drywell in the reactor enclosure. The logic, control units, and readout instrumentation are located in the control structure.

The CRDs' position detectors are located beneath the reactor vessel in the drywell. The normal design environments encountered in these areas are given in Section 3.11.

7.7.1.2.5 RMCS Operational Considerations CHAPTER 07 7.7-19 REV. 20, SEPTEMBER 2020

LGS UFSAR 7.7.1.2.5.1 RMCS General Information The RMCS is totally operable from the control room. Manual operation of individual control rods is possible with switches to effect control rod insertion, withdrawal, or settle. Rod position indicators provide the necessary information to ascertain the operating state and position of all control rods.

Conditions that prohibit control rod insertion are alarmed with the rod block annunciator.

7.7.1.2.5.2 RMCS Reactor Operator Information Table 7.7-1 gives information on the process instruments for the CRD hydraulic system. A large rod information display on the rod status display panel is patterned after a top view of the reactor core. The display allows the operator to acquire information rapidly by scanning.

Colored windows provide an overall indication of rod pattern and allow the operator to quickly identify an abnormal indication. The following information for each control rod is presented in the display:

a. Rod fully inserted (green)
b. Rod fully withdrawn (red)
c. Selected rod identification (coordinate position, white)
d. Accumulator trouble (amber)
e. Rod scram (blue)
f. Rod drift (red)

A separate, smaller display (the four-rod display) is located on the control console. This display shows the positions of the control rod selected for movement and the other rods in the rod group.

For display purposes, the control rods are considered in groups of four adjacent rods (a four-rod group) centered around a common core volume monitored by four LPRM strings. Rod groups at the periphery of the core can have less than four rods. The four-rod display shows the positions, in digital form, of the rods in the group to which the selected rod belongs. A lighted background on the digital display indicates which of the four rods is selected for movement.

The four-rod display allows the operator to better focus his attention on the portion of the core experiencing rod motion. A full core rod position display would tend to be confusing and difficult to read. In addition, on demand by the operator, the Plant Monitoring System can provide a printout of all rod positions.

During startup or shutdown, rods of a given sequence which are either fully withdrawn or fully inserted are indicated on the full core display with the full-in or full-out lights. In addition to the whole core display, a drifting rod is indicated by an alarm and red light in the control room. The Plant Monitoring System also monitors the rod drift condition.

An indication is also provided for rod travel beyond the limits of normal rod movement. If the rod drive piston moves to the OVERTRAVEL position, an alarm is sounded in the control room. The overtravel alarm provides a means to verify that the drive-to-rod coupling is intact because, with the CHAPTER 07 7.7-20 REV. 20, SEPTEMBER 2020

LGS UFSAR coupling in its normal condition, the drive cannot be physically withdrawn to the overtravel position.

Coupling integrity can be checked by attempting to withdraw the drive to the overtravel position.

For the displays above, the selected rod identification, accumulator trouble, and rod scram indicators are provided to the displays by the rod drive control system. The remaining information to the displays and the position information for the Plant Monitoring System are provided by the RPIS.

The following control room lights are provided to allow the operator to know the conditions of the CRD hydraulic system and the control circuitry:

a. Stabilizer valve selector switch position
b. Insert command energized
c. Withdraw command energized
d. Settle command energized
e. Withdrawal not permitted
f. Continuous withdrawal
g. Pressure control valve position
h. Flow control valve position
i. Drive water pump low suction pressure (alarm and pump trip)
j. Pump suction filter high differential (alarm only)
k. Charging water (to accumulator) low pressure (alarm only)
l. CRD temperature (alarm only)
m. SDV not drained (alarm only)
n. Scram valve pilot air header high/low pressure (alarm only)
o. System flow
p. SDV vent and drain valve position
q. Accumulator trouble alarm
r. Drive water filter differential pressure (alarm only) 7.7.1.2.5.3 RMCS Setpoints The subject system has no safety setpoints.

CHAPTER 07 7.7-21 REV. 20, SEPTEMBER 2020

LGS UFSAR 7.7.1.3 Recirculation Flow Control System - Instrumentation and Controls 7.7.1.3.1 RFCS Identification 7.7.1.3.1.1 RFCS General The objective of the RFCS is to control reactor power level, over a limited range, by controlling the flow rate of the reactor recirculating water (Figure 7.7-7).

The control valves varying the speed of the recirculation pumps by changing the voltage and frequency of the ac supply to each pump motor. The ac supply to each pump motor is provided by an Adjustable Speed Drive (ASD). The ASD consists of a transformer section coupled to a solid state power cell electronics section. Microprocessor controllers modulate the output of the power cells to convert the input 13.2 kV into the variable frequency MV output.

7.7.1.3.1.2 RFCS Classification This system is a power generation system and not safety-related.

7.7.1.3.1.3 RFCS Reference Design Table 7.1-2 lists reference design information. The RFCS is an operational system and has no safety function; therefore, there are no safety differences between this system and those of the above-referenced facilities. This system is functionally identical to the referenced system.

7.7.1.3.2 RFCS Power Sources 7.7.1.3.2.1 Normal/Alternate

a. Normal: ASD l20 Vac control power is derived from one of two dedicated UPSs (auctioneered) powered from their respective 480 Vac sources. Each UPS is capable of powering the control system for both ASD trains A and B, but normally only one UPS is dedicated to each drive.
b. Alternate: On loss of the normally aligned UPS or its 480 Vac source, the backup UPS automatically provides power to the ASD drive via the diode auctioneered circuit.

7.7.1.3.2.2 Deleted.

7.7.1.3.3 RFCS Equipment Design 7.7.1.3.3.1 RFCS General Reactor recirculation flow is changed by adjusting the speed of the two reactor recirculating pumps.

This is accomplished by adjusting the frequency and voltage of the electrical power supplied to the recirculation pump motor (Figure 7.7-7). Control of pump speed, and thus core flow, is such that at various control rod patterns, different power level changes can be accommodated. For a full power rod pattern, power change control down to approximately 65% of full power is possible using flow variation. Power control is possible over a range of approximately 35% of the maximum operating power level for that rod pattern.

CHAPTER 07 7.7-22 REV. 20, SEPTEMBER 2020

LGS UFSAR An increase in recirculation flow temporarily reduces the void content of the moderator by increasing the flow of coolant through the core. The additional neutron moderation increases the reactivity of the core, which causes the reactor power level to increase. The increased steam generation rate increases the steam volume in the core with a consequent negative reactivity effect, and a new steady-state power level is established. When recirculation flow is reduced, the power level is reduced in the reverse manner.

7.7.1.3.3.2 RFCS Pump Drive Motor 7.7.1.3.3.2.1 RFCS Pump Drive Motor Control Each recirculation pump motor has its own ASD for a power supply. The ASD uses digital controllers and solid state power electronics to vary the frequency and voltage supplied to the pump motor to give the desired pump speed.

7.7.1.3.3.2.2 RFCS Pump Drive Motor Trip Two dual trip coil circuit breakers are placed in series in the power feed to each recirculation pump motor. The RPT feature uses one trip coil of each breaker to trip the recirculation pump motors when either a turbine stop valve closure or turbine control valve fast closure occurs. The RPT subsystem consists of sensors for EHC oil pressure, valve position, and turbine pressure, relays, switches and indications arranged in two separate divisions of logic, either one of which will generate a trip signal.

Turbine stop valve closure is detected by monitoring stop valve position. Turbine control valve fast closure is detected by monitoring control valve oil pressure. Closure of the A and B stop valves or fast closure of the A and B control valves will initiate the Division I RPT logic. Closure of the C and D stop valves or fast closure of the C and D control valves will initiate the Division II RPT logic.

RPT is bypassed when reactor power is 29.5% as detected by turbine first stage pressure.

The recirculation pump motor breakers are also tripped by signals generated in the RRCS (Section 7.6.1.8) and by overcurrent relays monitoring the pump motor feeds (Section 8.1.6.1.12.a). The RRCS trip signal is connected to one trip coil; the RPT and overcurrent trip signals are connected to the other trip coil.

7.7.1.3.3.3 RFCS Variable-Frequency Drive 7.7.1.3.3.3.1 RFCS Adjustable Speed Drive The two ASD trains and their associated controls are identical. The ASD can continuously supply power to the pump motor at any speed to approximately 99.2% of pump rated speed.

The ASD is capable of starting the pump and accelerating it from a standstill to a desired operating speed when the pump motor thrust bearing is fully loaded by reactor pressure acting on the pump shaft. The main components of the ASD are MV input and output termination cabinets, input transformer cabinet, power electronic cell cabinets, control cabinet, relay cabinet and cooling cabinet.

a. Input and Output Termination Cabinets - where the 13.2kV supply and 4kV pump motor feed MV cables are terminated.

CHAPTER 07 7.7-23 REV. 20, SEPTEMBER 2020

LGS UFSAR

b. Transformer Cabinet - the transformer coverts the input 13.2 kV on the primary side down 750 Vac on the secondary side which is then connected directly to the power electronics cells. The ASD can operate under electrical supply variations of +/- 10% of rated voltage.
c. Power Electronics Cell Cabinets - there are three cabinets, each with four power cells connected in series per phase. They convert the input 750 Vac to the variable frequency

/ voltage proportional output which is fed directly to the pump motor. The power cells are composed mainly of a diode bridge, filtering capacitors, and an insulated gate bipolar transistor (IGBT). The lGBT is a semiconductor field effect transistor that fires a bipolar transistor to provide reliable high speed switching with high-power capabilities.

d. Control Cabinet - this cabinet houses the main controllers which control all the essential elements of the ASD and the PLCs which provide most of the interface between the main controllers and the plant. It also contains the pre-charge circuitry used to start the ASD.

The primary functions of the redundant main controllers include:

  • regulate the firing of the power cells to produce the Pulse Width Modulated (PWM) output signal fed to the pump motor
  • communicate to I/O and PLC for status information
  • execute the plant specific System Operational Program (SOP)
  • monitor drive and process parameters for alarm output Only one of the two controllers is actively controlling all the power cells under all conditions. Typically, the A controller is active with the B controller in stand-by to take control should the A controller detect failure.

The primary functions of the redundant PLCs include:

  • interface plant parameters and drive status between the plant and the main controller via I/O
  • perform alarm checking of certain drive parameters
  • the stand-by PLC monitors communications of the active PLC for failure detection
  • actuates alarm and output status information The pre-charge circuitry performs a controlled slow charge of the power cells capacitors to extend capacitor life.
e. Relay Cabinet - this cabinet houses protective relays for pump motor protection and over-frequency protection. These relays are arranged in a two-out-of-three logic and tied into the supply 13.2 kV breaker trip device. They provide backup equipment protection in the event of malfunction of failure of the ASD controllers.
f. Coolant Cabinet - this cabinet houses the coolant pumps, piping, valves and controls for supplying closed loop demineralized cooling water throughout the input transformer, power cells and cabinets. The coolant is routed to a heat exchanger dedicated to each ASD train which is cooled by service water. A third swing heat exchanger is kept in reserve to be CHAPTER 07 7.7-24 REV. 20, SEPTEMBER 2020

LGS UFSAR aligned to either ASD train as a backup, should the dedicated heat exchanger need to be taken OOS.

7.7.1.3.3.3.2 Deleted.

7.7.1.3.3.4 RFCS Speed Control Components 7.7.1.3.3.4.1 RFCS Speed Control Components for ASD The ASD speed control system controls the firing of the power cells for both ASD trains. Each ASO train is individually controlled manually from either the MCR or locally at the ASD control cabinet. The speed control system Human Machine lnterface (HMI) is different for each location.

On MCR console *0-C602, it consists of two columns of three pushbuttons, one column dedicated for raising speed and the other for lowering speed. Speed can be raised or lowered at fixed step changes of 1, 5 or 10 rpm (raising) or 1, 5 or 30 rpm (lowering). All six pushbuttons are one-shot in that only one digital speed change command is read by the ASD for each press and release. Extending speed changes beyond the available increments requires repeat press/release of the appropriate pushbutton. The rate of change in speed of the pump is fixed at 2.5% / second for all commanded speed changes. The MCR controls interface with two redundant Remote I/O (RIO) cabinets per train. The RlOs convert each discrete analog speed change command to a digital signal for communication to the main speed controller. The speed change command must be received by both RlOs in order for the command to be acted upon by the ASD; otherwise a single speed change command to only one of the two RlOs will be ignored. This prevents uncontrolled speed changes from spurious invalid signals.

At the local control panel, it consists of a touch screen display HMI which graphically mimics the MCR console speed controls with the same functionality. The local HMI communicates directly to the main controller as a digital input. Transfer of control to the local HMI is via the keylock switch (controlled by SSV)) mounted on the control panel.

7.7.1.3.3.4.2 Deleted.

7.7.1.3.3.4.3 RFCS Manual Control Station There is no manual control station. Instead, six discrete pushbuttons, each corresponding to a fixed speed change increment and direction, are used on the MCR console or on the local touch screen HMI.

7.7.1.3.3.4.4 RFCS Speed Controller The ASD speed controller is a microprocessor based multi-card rack module device mounted inside the ASD control cabinet. It accepts the speed change command signals from the MCR console via the redundant RIOs and PLCs over redundant communication channels. It also receives speed change commands from the local touch screen HMI when it is enabled.

7.7.1.3.3.4.5 RFCS Signal Failure Alarm The speed change command must be received simultaneously by both redundant RIOs for each ASD train to be acted upon by the controller. Invalid spurious signals are ignored. An algorithm CHAPTER 07 7.7-25 REV. 20, SEPTEMBER 2020

LGS UFSAR within the ASD controller selects which signal is valid and ignores the malfunctioning redundant channel. The algorithm also determines when the ignored channel signal is again valid. Loss of either one or both communication channels for all inputs between the ASO and each RIO results in a MCR alarm.

7.7.1.3.3.4.6 RFCS Startup Signal Generator There is no discrete startup signal generator module within the ASD control system. Once the ASD is ready to run, startup control of the pump motor resides in the internal System Operating Program (SOP) code and eagle code of the speed controller.

7.7.1.3.3.4.7 RFCS Speed Limiter The speed limiters for each ASD train are adjustable high speed limiter setpoints which reside in the speed controller System Operating Procedure (SOP) code.

The speed controller setpoint signal is automatically limited by the speed limiters to approximately 28% of rated speed if the recirculation pump main discharge valve is not fully open, if the feedwater flow is less than 20% of rated flow, or if the reactor level is less than level

3. The setpoint is limited to a speed corresponding to approximately 75% of rated core flow if the reactor level is low and any feedwater pump flow is low or if a condensate pump trips when feedwater flow is greater than 85% of Nuclear Boiler rated flow. If the discharge valve is closed and the pump is at high speed, the pump may overheat. If feedwater flow is less than 20% of rated flow, there is not enough subcooling of the downcomer water to provide the net positive suction head needed by the jet pumps and recirculating pumps to operate at flow rates greater than approximately 30%, this functionality is performed by the ASD internal controller.

7.7.1.3.3.4.8 RFCS Recirculation Loop Starting Sequence Each recirculation loop is independently put into operation by operating the controls of each loop as follows:

a. The operator can not start the recirc pump unitl the Ready to Precharge light on MCR console *0-C602 is lit. This light indicates when all external plant and all internal ASD permissives required for pump start are met.
b. Anytime after the Ready to Pre-charge light is lit (no time limit) the operator can start the pump by taking the MV input breaker control switch to "Start." This initiates the Pre-charge/Start cycle.
c. During the Pre-charge/Start cycle the 480 Vac Pre-charge breaker closes which charges up the capacitors on the power cells.
d. Once the ASD power cell capacitors are fully charged and all permissives are still met, the ASD automatically closes the MV input breaker.
e. Once the MV is sensed at the ASD input, the ASD automatically ramps up the recirc pump motor from 0 to 20% speed at 5% / second, and then from 20% to 28% at 2.5% / second with no operator action.

CHAPTER 07 7.7-26 REV. 20, SEPTEMBER 2020

LGS UFSAR

f. Once at or above 28% speed, the operator can incrementally raise or lower speed while staying above 28%.

7.7.1.3.3.5 RFCS Testability The ASD and associated speed controllers, are functioning during normal power operation. Any abnormal operation of these components can be detected during operation. The components that do not continually function during normal operation can be tested and inspected for calibration and operability during scheduled plant shutdowns. All the recirculation flow control system components can be tested and inspected during scheduled shutdowns.

7.7.1.3.4 RFCS Environmental Considerations The RFCS is not required for safety purposes, nor is it required to operate after a DBA. The system is required to operate in the normal plant environment for power generation purposes only.

The only part of the recirculation flow control equipment in the drywell is the pump motor. It is subject to the design conditions environment specified in Section 3.11.

The logic control units and instrumentation are located in the control structure and are subject to that environment (Section 3.11).

7.7.1.3.5 RFCS Operational Considerations 7.7.1.3.5.1 RFCS General Information The ASD system includes self-test, failure detection, indications and alarms to keep the operator informed of the status of the system and equipment, and to permit the operator to quickly determine the location of malfunctioning equipment. Temperature of the equipment is monitored and alarmed if safe levels are exceeded. A recorder for each loop displays the closed loop cooling temperature into the cabinets and out of the transformer and power cells.

Each recorder also shows ASD output parameters of voltage, current power, frequency, and frequency demand. Indicators are provided to show recirculation loop flow and valve positions.

MCR alarms alert the operator to approximately 256 drive and process alarm conditions which are individually displayed for quick diagnosis on the ASD touch screen display HMI on MCR panel 10-C602 and local control panel.

7.7.1.3.5.2 RFCS Reactor Operator Information Indication and alarms are provided to keep the operator informed of the status of the systems and equipment, and to allow the operator to quickly determine the location of malfunctioning equipment.

The visual display consists of recirculation loop flow, valve position, speed demand, ASD cooling loop temperatures and ASD output parameters of volts, current, frequency and power.

Alarms are provided to alert the operator to various ASD and process alarm conditions such as:

malfunctioning equipment, loss of ASD communications, ASD cabinet and components high temperatures, cooling loop conductivity, temperature and level, high or low level runback initiation or bypass, speed hold, etc.

7.7.1.3.5.3 RFCS Setpoints CHAPTER 07 7.7-27 REV. 20, SEPTEMBER 2020

LGS UFSAR The subject system has no safety setpoints. However, signals from the RRCS can cause a trip of the recirculation pump motor breakers in the event of an ATWS (Section 7.6.1.8).

7.7.1.4 Feedwater Control System - Instrumentation and Controls 7.7.1.4.1 FCS Identification 7.7.1.4.1.1 FCS General The FCS controls the flow of feedwater into the RPV to maintain the water in the vessel within predetermined levels during all plant operating modes. The predetermined range of water level is based on the requirements of the steam separators (this includes limiting carry-over, which affects turbine performance, and carry-under, which affects recirculation pump operation). The feedwater control system employs water level, steam flow, and feedwater flow as a three-element control.

Single-element control is also available based on water level only. Normally, the signal from the feedwater flow is equal to the steam flow signal; thus, if a change in the steam flow occurs, the feedwater flow follows. The steam flow signal anticipates the change in water level that results from a change in load. The level signal corrects any mismatch between the steam and feedwater flow that causes the level of the water in the reactor vessel to rise or fall accordingly.

7.7.1.4.1.2 FCS Classification This system is a power generation system and is classified as not related to safety.

7.7.1.4.1.3 FCS Reference Design Table 7.1-2 lists reference design information. The FCS is an operational system and has no safety function. Therefore, there are no safety differences between this system and those of the above-referenced facilities. The subject system is functionally identical to the referenced system.

7.7.1.4.2 FCS Power Sources The FCS power is supplied by three independent sources so that no single power failure can incapacitate more than one of the three level sensing elements used for narrow range indication.

Power for two of the three narrow range level sensing and indicating channels is supplied from the 120 V ac instrumentation buses and the other channel is powered from a plant 125 V dc battery and an inverter. The 120 V ac instrumentation bus feeds are each powered through a separate uninterruptible power supply. The FCS panel inserts are supplied with independent 120VAC power feeds. In all cases, each AC feed powers a separate power supply and is UPS backed.

The configuration allows continuous operation upon loss of any power feed or power supply unit.

High Level 8 channels A and B level transmitters and corresponding equipment for High Level 8 trip logic are single fed by two independent 120VAC feeds. High Level 8 channel C and D transmitters and corresponding equipment for High Level 8 trip logic are powered by the same 125VDC power source. High Level 8 channel D is also powered by an AC source.

7.7.1.4.3 FCS Equipment Design 7.7.1.4.3.1 FCS General CHAPTER 07 7.7-28 REV. 20, SEPTEMBER 2020

LGS UFSAR During normal plant operation, the feedwater control system automatically regulates feedwater flow into the reactor vessel. The system can be manually operated.

The feedwater flow control instrumentation measures the water level in the reactor vessel, the feedwater flow rate into the reactor vessel, and the steam flow rate from the reactor vessel. During automatic operation, these three measurements are used to control feedwater flow.

The optimum reactor vessel water level is determined by the requirements of the steam separators.

The separators limit water carry-over in the steam going to the turbines and limit steam carry-under in water returning to the core. The water level in the reactor vessel is maintained within +/-2 inches of the setpoint value during normal steady-state operation and within the high and low level trip setpoints during normal plant maneuvering transients.

This control capability is achieved during plant load changes by balancing the mass flow rate of feedwater to the reactor vessel with the steam flow from the reactor vessel. The feedwater flow is regulated by controlling the speed of the turbine-driven feedwater pumps to deliver the required flow to the reactor vessel.

The RRCS can initiate a feedwater runback, reducing flow to 0% within 15 seconds. This runback is independent of the feedwater control operating mode and overrides the loss of signal interlock that prohibits change of feed pump output under loss of control signal conditions. Control of the feedwater system can be regained by the operator 30 seconds after the runback begins. This runback is discussed in Section 7.6.1.8.3.3. ATWS alarm lights are provided on the front of the feedwater control panel. The feedwater system trip contacts associated with the RRCS ATWS runback are required to be high quality but not necessarily safety-grade.

7.7.1.4.3.2 FCS Reactor Vessel Water Level Measurement Reactor vessel narrow range water level is measured by four identical, independent sensing systems. For each channel, a differential pressure transmitter senses the difference between the pressure caused by a constant reference column of water and the pressure caused by the variable height of water in the reactor vessel. The differential pressure transmitters are installed on lines that serve other systems (Section 7.7.1.1). All four narrow range differential pressure transmitters are used as input to the digital FCS. The FCS system performs error checking and soft majority selection (SMS) on the four transmitters to ensure the final level control signal is accurate. This produces a highly reliable and fault tolerant system, reducing the likelihood of a transmitter failure causing a transient or SCRAM. Three narrow range differential pressure transmitters are used for MCR panel indication. All four narrow range differential pressure signals are displayed on the FCS operator workstation and are arranged to provide failure-tolerant trips of the main turbine and feed pump prime movers. A fifth level sensing system (upset range) provides level indication beyond the span of the narrow range devices.

The SMS water level and upset range water level are continually recorded in the control room.

7.7.1.4.3.3 FCS Steam Flow Measurement Steam flow is sensed at each main steam line flow restrictor by a differential pressure transmitter.

The signals from the four transmitters are sent to the FCS where it is manipulated to produce the true mass flow rate for each line. The FCS than transmits the flow rate of each line to indicators in the control room. The four steam line flow signals are also connected to a SMS within the FCS; the SMS selects the mid-value of its inputs. Total steam flow is calculated as this mid-value CHAPTER 07 7.7-29 REV. 20, SEPTEMBER 2020

LGS UFSAR multiplied by the number of valid steam flow signals. The total steam flow is recorded in the control room. All four differential pressure transmitter signals and the SMS flow signal are displayed on the FCS operator work station.

7.7.1.4.3.4 FCS Feedwater Flow Measurement Feedwater flow is sensed at a flow element in each feedwater line by differential pressure transmitters. One transmitter is designated the primary feed flow measurement and the other is designated as the secondary feed flow measurement. In addition, a calculation of the feed flow is performed for each pump based on pump speeds, and on suction and discharge pressures, with miniflow measurement subtracted. The primary and secondary measurements are compared to each other and to the calculated measurement, to validate the values.

The signals from the transmitters are sent to the FCS where they are manipulated to produce a true mass flow rate for each line. The FCS then transmits the flow rate of each line to indicators in the control room. The total feedwater flow rate is sent to a recorder in the control room. The calculated feedwater flow rate is sent to the plant computer. All differential pressure transmitters, total feedwater flow, and calculated feedwater flow are displayed on the FCS operator workstation.

Feedwater flow is also sensed by a second flow measurement system based on ultrasonic flow measurement. The system includes spool pieces integrated into each feedwater line, where each spool piece contains ultrasonic transducers, a pressure trap for measuring fluid pressure, and a thermowell to measure fluid temperature. The ultrasonic transducers measure fluid velocity, while the pressure transmitters measure the feedwater pressure. Measurements from this system are combined to produce volumetric flow, mass flow, and temperature of the feedwater fluid so that the values can be utilized as inputs to core thermal power computations. Outputs from the system are received by the plant process computer.

7.7.1.4.3.5 FCS Feedwater/Level Control The FCS has one manual mode and two modes of automatic operation. The manual mode operates the RFPs in parallel. The two automatic modes are single-element and three-element control. The FCS can automatically transfer from single-element to three-element control and vice versa. Single-element control utilizes only the reactor water level to determine the need for increasing or decreasing feedwater flow. Single-element control is normally used when reactor power is less than 20%. Three-element control receives input from three different parameters:

reactor water level, total steam flow and total feedwater flow. When reactor power is above 20%,

an automatic transfer to three-element control is performed. Automatic transfer can be blocked by the operator selecting forced single-element control before the setpoint is reached.

The FCS has a level controller in the control room. The level controller serves three functions:

setting the level setpoint, transfer between single-element and three-element control and indication.

Level setpoint is increased or decreased by pushbuttons. Transfer between single-element and three-element is performed by lighted pushbuttons. The level and feedwater flow controllers are balanced to allow bumpless transfer from single-element to three-element control.

During three-element control the reactor water level is compared to reactor water level setpoint and level controller. The level controller output signal is then summed together with the total steam flow to create a feedwater flow demand signal. The total feedwater flow is compared to the feedwater flow demand signal and is passed to the feedwater flow controller. The feedwater flow controller CHAPTER 07 7.7-30 REV. 20, SEPTEMBER 2020

LGS UFSAR output is then sent to the RFPs. Upon loss of condensate pump the feedwater flow is automatically limited to nominally 88% of rated feedwater flow.

To improve level control after a SCRAM, a predefined feedwater flow setpoint profile is used. The profile predicts the steam production after a scram and controls the feedwater flow in order to minimize flow and reactor level transients. The profile is also designed to meet the relative runback in the reactor recirculation flow, minimizing thermal stress on the reactor internals caused by non-preheated feedwater.

The scram profile can only be used if three-element control is possible. If the FCS is not in three-element control, then the FCS will automatically switch to three-element control. If the FCS cannot switch to three-element control, then single-element control continues to be active.

Once the scram profile reset logic has been satisfied, the post scram flow profile signals automatically reset and automatic bumpless transfer to single-element control occurs. The scram profile can only be active for a predefined period of time before it is forced to reset.

7.7.1.4.3.6 FCS Interlocks The level control system also provides interlocks and control functions to other systems. When one of the reactor feed pumps is lost and coincident or subsequent low water level exists, or a condensate pump trips when feedwater flow is greater than 80% of Nuclear Boiler rated flow, recirculation flow is reduced to within the power capabilities of the remaining reactor feed pumps.

This reduction helps avoid a low level scram by reducing the steaming rate. Reactor recirculation flow is also reduced on sustained low feedwater flow to ensure that adequate NPSH is provided for the recirculation system.

An interlock from total steam flow is used to initiate insertion of the RWM block. An alarm on low steam flow indicates that the above RWM insertion interlock setpoint is being approached. Alarms are also provided for high and low water level, and for reactor high pressure. Interlocks trip the plant turbine and feedwater pumps if there is reactor high water level. Interlocks trip the reactor recirculation pumps upon loss of the main generator coolant system if feedwater flow exceeds 44%

of Nuclear Boiler rated flow.

7.7.1.4.3.7 FCS Turbine-Driven Feedwater Pump Control Feedwater is delivered to the reactor vessel through turbine-driven feedwater pumps arranged in parallel. The turbines are driven by steam from the reactor vessel during normal operation, or from the auxiliary boilers during startup. The feedwater control signal from the level controller is fed to the turbine speed control systems, which adjust the speed of their associated turbine so that feedwater flow is proportional to the feedwater demand signal. Each turbine can be controlled by its manual/automatic transfer station. The M/A stations are located in the control room and their function is as follows: Manual fast or slow control of the RFP, transfer between Automatic and Manual mode Bias control, and Indication. In manual mode, pushbuttons are used to increase or decrease the M/A station output. Transfer between manual and automatic is performed by lighted pushbuttons. When two or more RFPs are in automatic mode, it is possible to bias the common control output between them. The bias adjustment is performed by the lower/raise pushbuttons on the M/A stations. The bias is maintained as long as all RFPs are in automatic mode. If two or more RFPs are transferred to manual mode, the M/A station output is balanced to actual control output value. The bias value is automatically adjusted to CHAPTER 07 7.7-31 REV. 20, SEPTEMBER 2020

LGS UFSAR zero for the RFP still in automatic mode. Each feedwater pump has a minimum flow valve. The minflow valve can be controlled in manual or automatic mode. Manual mode control is performed from M/A stations in the control room. In automatic mode, the minflow valve positions are controlled in an open loop to avoid the situation where the minflow control and the FCS "fight" each other. If the feedwater control signal to the turbine signal is lost, a digital output in the feedwater control circuit causes the turbine speed control system to lock the turbine speed "as is" and initiates an alarm in the control room. The level controller and the manual/automatic transfer stations associated with each turbine speed controller are the "bumpless transfer" type.

7.7.1.4.3.8 FCS Testability All feedwater flow control system components can be tested and inspected before plant operation and during scheduled shutdowns. Reactor vessel water level indications from the three water level sensing systems can be compared during normal operation to detect instrument malfunctions.

Steam mass flow rate and feedwater mass flow rate can be compared during constant load operation to detect inconsistencies in their signals. The FCS has a built in process model and disturbance tool. The process model is used during outages for testing and verification of parameter and control algorithm modifications. The process model simulates the final control elements. The disturbance tool allows operators to introduce disturbances at predefined locations in the control algorithm. The predefined locations are as follows: reactor water level setpoint, flow demand signal, total steam flow, total feedwater flow, and demand output of each final control element.

7.7.1.4.4 FCS Environmental Considerations The feedwater control system is not required for safety purposes, nor is it required to operate after a DBA. This system is required to operate in the normal plant environment for power generation only. The reactor feed pumps in the turbine enclosure experience the normal design environments listed in Section 3.11.

7.7.1.4.5 FCS Operational Considerations 7.7.1.4.5.1 FCS General Information The level controller is located in the control room where, at the operator's discretion, the system can be operated either manually or automatically.

Manual control of the individual turbine-driven feedwater pumps is available to the operator in the control room. This includes control of the low flow feedwater bypass valve that can be used for startup.

If there is a loss of feedwater and reactor vessel water level falls below the low level setpoint, the RPS causes plant shutdown, thus preventing any further lowering of the reactor vessel water level.

7.7.1.4.5.2 FCS Reactor Operator Information Indicators and alarms, provided to keep the operator informed of the status of the system, are as noted in previous sections. The Operator Station provides an interface through which the Operator CHAPTER 07 7.7-32 REV. 20, SEPTEMBER 2020

LGS UFSAR can observe and interact with plant processes. The Operator can enter commands using the mouse or keyboard. Process alarms and events are time-tagged and presented in a list.

7.7.1.4.5.3 FCS Setpoints The subject system has no safety setpoints.

7.7.1.5 Pressure Regulator and Turbine-Generator System - Instrumentation and Controls 7.7.1.5.1 PRTGS Identification 7.7.1.5.1.1 PRTGS General One of the features of direct cycle BWRs is the direct passage of the nuclear boiler generated steam through the turbine and regenerative system. In this system the turbine is slaved to the reactor in that all steam generated by the reactor is normally accepted by the turbine. The operation of the reactor requires a pressure regulator to maintain a constant (within the range of the regulator controller proportional band setting) turbine inlet pressure.

The turbine pressure regulator normally controls the turbine control valves to maintain constant (within the range of the regulator controller proportional band setting) turbine inlet pressure. In addition, the pressure regulator also operates the steam bypass valves so that a portion of nuclear boiler rated flow can be bypassed when operating at steam flow loads above those that can be accepted by the turbine, as well as during the startup and shutdown phases.

The overall turbine-generator and pressure control system accomplishes the following:

a. Controls turbine speed and turbine acceleration
b. Operates the steam bypass system to keep reactor pressure within limits and avoid large power transients
c. Controls main turbine inlet pressure within the proportional band setting of the pressure regulator 7.7.1.5.1.2 PRTGS Classification The PRTGS is classified as a primary power generation system. That is, it is not a safety system, but its operation is essential to the power production cycle.

7.7.1.5.1.3 PRTGS Reference Design Table 7.1-2 lists reference design information. The subject instrumentation and control system is an operational system and has no safety function. Therefore, there are no safety design differences between this system and those of the reference design facilities. This system is functionally identical to the referenced systems.

7.7.1.5.2 PRTGS Power Sources CHAPTER 07 7.7-33 REV. 20, SEPTEMBER 2020

LGS UFSAR A. The DEHC electrical power is supplied by redundant 120Vac UPS Distribution Panels. Failure of either power supply source will not affected the operation of the PRTGS control functions.

7.7.1.5.3 PRTGS Equipment Design 7.7.1.5.3.1 PRTGS General BWR pressure is controlled by regulating the main steam pressure through modulation of the turbine control or steam bypass valves. Command signals to these valves are generated by redundant control elements using the sensed turbine inlet pressure signals as the feedback, as shown in Figure 7.7-9. For normal operation, the turbine control valves regulate steam pressure; however, whenever the total steam flow demand from the pressure regulator exceeds the capacity of the turbine control valves, the pressure control system sends the excess steam flow directly to the main condenser through the steam bypass valves. The plant's ability to follow grid system load demands is determined by adjusting reactor power level, manually varying reactor recirculation flow, or manually moving control rods. In response to the resulting steam production changes, the pressure control system adjusts the turbine control valves to accept the steam output change, thereby regulating steam pressure.

7.7.1.5.3.2 PRTGS Steam Pressure Control During normal plant operation, steam pressure is controlled by the main turbine control valves, positioned in response to the pressure regulation demand signal (Figure 7.7-9). The steam bypass valves are normally closed.

Two essentially redundant regulators are provided so that the one is the primary controller and the backup controller is in standby tracking the same steam flow demand signal to ensure, if required, an automatic bumpless transfer to the backup regulator. Each regulator accesses the pressure signal that is validated based upon the turbine inlet (throttle) pressure signals.

The turbine control valve (steam flow) demand signal is limited, after passage through the low value gate (Figure 7.7-9), to the signal required to fully open the main control valves. Thus, if the pressure control system requires that additional steam flow be released from the reactor when the control valves reach wide open, the control signal error to the bypass valves increases and causes bypass actuation.

7.7.1.5.3.3 PRTGS Steam Bypass System The steam bypass system is designed to control steam pressure when reactor steam generation exceeds turbine requirements during startup (pressure, speed ramping, and synchronizing),

sudden load reduction, and cooldown.

The bypass capacity of the system is 24.3% of the rated steam flow; sudden load reductions within the capacity of the steam bypass can be accommodated without reactor scram.

Normally, the bypass valves are closed and the pressure regulator controls the turbine control valves, directing all steam flow to the turbine. If the speed governor or the load limiter restricts steam flow to the turbine, the regulator controls system pressure by opening the bypass valves. If CHAPTER 07 7.7-34 REV. 20, SEPTEMBER 2020

LGS UFSAR the capacity of the bypass valves is exceeded while the turbine cannot accept an increase in steam flow, the system pressure rises and RPS action causes the reactor to shutdown.

The bypass valves are an automatically operated, regulating-type that are proportionally controlled by the turbine pressure regulator and control system.

The turbine control system provides a signal to the bypass valves that corresponds to both the "error" between the turbine control valve opening required by the controlling pressure regulator and the turbine control valve position demanded by the output of the low value gate circuit (Figure 7.7-9). An adjustable bias signal is provided to keep the bypass valves closed during momentary differences that occur during normal operational transients.

Bypass valves and controls are designed so that bypass steam flow is shut off upon loss of control system electrical power, hydraulic pressure, or low condenser vacuum.

7.7.1.5.3.4 PRTGS Turbine Speed-Load Control Systems 7.7.1.5.3.4.1 Normal Operation During base load plant operation, the turbine load reference is held above the desired load, so that the pressure regulation demand governs the turbine control valves.

7.7.1.5.3.4.2 PRTGS Behavior of Turbine Outside of Normal Operation

a. Turbine startup Prior to turbine startup, sufficient reactor steam flow is generated to permit the steam bypass valves to maintain reactor pressure control while the turbine is brought up to speed and synchronized under its speed-load control.
b. Partial load rejection During partial load rejection transients, which appear to the reactor as a reduction in turbine load demand resulting from an increase in generator (or grid) frequency, the turbine pressure control scheme allows the reduced turbine speed-load demand to override the pressure regulation demand and thereby directly regulate the turbine control valves. The pressure controller modulates the bypass valves to maintain reactor pressure.
c. Turbine shutdown or turbine-generator trip During turbine shutdown or turbine-generator trip conditions, the main turbine stop valves and control valves are closed. Reactor steam flow is then passed through the steam bypass valves under steam pressure control and through the reactor SRVs, as needed.
d. Steam bypass operation Fast opening of the steam bypass valves during turbine trips or generator load rejections requires coordinated action with the turbine control system. When the turbine control valves are under pressure control, no bypass steam flow is demanded; conversely, when the turbine speed-load demand falls below the CHAPTER 07 7.7-35 REV. 20, SEPTEMBER 2020

LGS UFSAR pressure regulation demand, a net bypass flow demand is computed. During turbine or generator trip events resulting in fast closure of the turbine stop or control valves, the turbine control valve demand is immediately reduced to zero, causing the bypass steam flow demand to equal the initial pressure regulation demand.

e. Loss of turbine control system power Turbine controls and valves are designed so that the turbine stop and control valves close upon loss of control system power or hydraulic pressure.

7.7.1.5.3.5 PRTGS Turbine-Generator to RPS Interface The RPS initiates reactor scram when any monitored plant condition requires it. Two such conditions are: turbine stop valve closure and turbine control valve fast closure when reactor power is above 29.5% of rated power. The turbine stop valve closure signal is generated before the turbine stop valves have closed more than 5%. This signal originates from position switches that sense stop valve motion away from fully open. A limit switch is provided on each of the four turbine stop valves. The switches are closed when the stop valves are fully open, and open within 10 milliseconds after the setpoint is reached. The switches are electrically isolated from each other and from other turbine plant equipment.

The control valve fast closure signal is generated by four hydraulic oil pressure sensors that are distributed equally among the control valves and that sense hydraulic oil pressure decay as an indication of fast control valve closure. The switches are closed when the valves are open, and open within 30 milliseconds after the control valves start to close in a fast closure mode.

To avoid reactor scram due to stop valve closure or control valve fast closure when power is below 29.5% of rated power, two independent sensing lines are provided from pressure taps located in the turbine steam supply lines upstream of the high pressure turbine first stage and are connected to pressure switches to supply power level signals to the RPS. The pressure taps are located to provide a pressure signal proportional to turbine steam flow. The pressure taps are shared with other instrumentation sensors. All sensors have individual shutoff valves.

7.7.1.5.3.6 PRTGS Turbine-Generator to Main Steam Isolation System Interface Four independent main condenser vacuum sensors provide an isolation signal to the NSSS MSIVs. Condenser vacuum transmitters and trip units are discussed in Section 7.3.1.1.2.

7.7.1.5.3.7 PRTGS Testability Controls are provided to test the turbine valve RPS interface signal switches in the following ways:

a. Actuate each stop valve individually to the 10% closed point with no interaction with other valves
b. Actuate the following pairs of stop valves to the 10% closed point, one pair at a time: 1 and 2; 3 and 4; 1 and 3; 2 and 4
c. Actuate one control valve fast closure hydraulic oil pressure switch at a time by individually closing the control test mode
d. Individually test each main condenser low vacuum instrument channel CHAPTER 07 7.7-36 REV. 20, SEPTEMBER 2020

LGS UFSAR 7.7.1.5.4 PRTGS Environmental Considerations The turbine-generator control system is required to operate in the normal plant environment for power generation purposes only.

Instruments and controls on the turbine experience the turbine building normal design environment as listed in Section 3.11. The logic, remote control units, and instrument terminals located in the control structure experience the environment listed in Section 3.11.

7.7.1.5.5 PRTGS Operational Considerations 7.7.1.5.5.1 PRTGS General Information Process variables controlled by the pressure regulator and speed/ load control system are displayed in the control room on the turbine-generator section of the control boards.

Manual and automatic control modes for the various turbine-generator operational modes (such as startup, normal operation, and shutdown) are available to the operator from the control room.

Automatic HMI display indication is provided to inform the operator as to the operating mode of the turbine-generator unit. The pressure control is active during all modes of operation and cannot be selected or deselected. The default mode allows the operator to select speed reference and rate, start up the turbine, close the bypass valves, and set the turbine in Turbine-Follow-Reactor mode.

Two pressure control channels, operating redundantly, receive inputs from validated pressure signals from the two pressure transducers in the main steam line upstream of the main steam stop valves and from the pressure reference function. Main steam pressure indications and pressure setpoint adjustments/indications are located on the HMI workstation of the turbine control panel. Minor and Major Trouble Pressure Control annunciation is provided in the control room.

7.7.1.5.5.2 PRTGS Reactor Operator Information The NSSS pressure regulator has the following controls and information displayed in the control room:

a. Main steam pressure transducer output
b. Main steam pressure transducer output
c. Main steam pressure regulator setpoint A
d. Main steam pressure regulator setpoint B
e. Individual bypass valve position
f. Individual bypass valve demand control signal
g. Bypass valve test controls CHAPTER 07 7.7-37 REV. 20, SEPTEMBER 2020

LGS UFSAR

h. Pressure regulator selection control
i. Individual control valve position
j. Individual control valve demand control signal 7.7.1.5.5.3 PRTGS Setpoints There are no safety setpoints associated with this system.

7.7.1.6 Neutron Monitoring System - Instrumentation and Controls The NMS consists of the following nonsafety-related systems:

a. Source range monitor system
b. Rod block monitor system
c. Traversing incore probe system The safety-related portion is described in Section 7.6.1.4.

7.7.1.6.1 Source Range Monitor System 7.7.1.6.1.1 SRM Equipment Design 7.7.1.6.1.1.1 SRM Circuit Description The SRM provides neutron flux information during reactor startup and low flux level operations.

There are four SRM channels. Each includes one detector that can be physically positioned in the core from the control room (Figure 7.6-2).

The detectors are inserted into the core for reactor startup. They can be withdrawn if the indicated count rate is between preset limits or if the IRM is on the third range or above (Figure 7.6-3).

During initial startup, neutron flux is monitored by source range neutron monitoring channels, providing a scram signal when the preset flux level of any channel has been reached. This logic is removed from the scram circuitry after completion of initial startup.

a. Power Supply - The power for the monitors is supplied from the two 120 V ac RPS buses. Two monitors are powered from each bus.
b. Physical Arrangement - Each detector assembly consists of a miniature fission chamber and a low-noise, quartz fiber insulated transmission cable. The sensitivity of the detector is 1.2x10-3 cps/nv nominal, 5.0x10-4 cps/nv minimum, and 2.5x10-3 cps/nv maximum. The detector cable is connected underneath the reactor vessel to the multiple shielded coaxial cable. This shielded cable carries the pulses to a pulse current preamplifier located outside the drywell.

CHAPTER 07 7.7-38 REV. 20, SEPTEMBER 2020

LGS UFSAR The detector and cable are located inside the reactor vessel in a dry tube sealed against reactor vessel pressure. A remote controlled detector drive system moves the detector along the dry tube. The active length of fuel is positioned vertically up to 30 inches below the reactor fuel region (Figure 7.6-2). When a detector arrives at a travel end point, detector motion is automatically stopped.

SRM drive control arrangement and logic are presented in Figure 7.6-3. The electronics for the SRM, their trips, and their bypasses are located in four cabinets. Source range signal conditioning equipment is designed so that it can also be used for open vessel experiments.

c. Signal Conditioning - A current pulse preamplifier provides amplification and impedance matching for the signal conditioning electronics (Figure 7.7-13).

The signal conditioning equipment converts the current pulses to analog dc currents that correspond to the logarithm of the count rate. The equipment also derives the period. The output is displayed on front panel meters and is provided to remote meters and recorders. The logarithm of the count rate meter displays the rate of occurrence of the input current pulses. The period meter displays the time in seconds for the count rate to change by a factor of 2.7. In addition, the equipment contains integral test and calibration circuits, trip circuits, power supplies, and selector circuits.

d. Trip Functions - The trip outputs of the SRM operate in the fail-safe mode. Loss of power to the SRM causes the associated outputs to become tripped.

The SRM provides signals indicating SRM upscale, downscale, inoperative, and incorrect detector position to the RMCS to block rod withdrawal under certain conditions. Any SRM channel can initiate a rod block. These rod blocking functions are discussed in Section 7.7.1.6.2. Appropriate lights and annunciators are also actuated to indicate the existence of these conditions. Refer to Table 7.7-4 for a description of the trip functions.

7.7.1.6.1.1.2 SRM Bypasses and Interlocks One of the four SRM channels can be bypassed at any one time by a switch on the operator's control panel.

7.7.1.6.1.1.3 SRM Redundancy and Diversity SRM channels are not redundant because SRM detectors are spatially dependent and do not serve as a backup to other detectors.

7.7.1.6.1.1.4 SRM Testability Each SRM channel can be fully tested and calibrated using written procedures. Inspection and testing are performed as required on the SRM detector drive mechanism; the mechanism can be checked for full insertion and retraction capability. The various combinations of SRM trips can be introduced to ensure the operability of the rod blocking functions.

7.7.1.6.1.2 SRM Environmental Considerations CHAPTER 07 7.7-39 REV. 20, SEPTEMBER 2020

LGS UFSAR The wiring, cables, and connectors located within the drywell are designed for continuous-duty in the conditions described in Section 3.11. The above SRM system components are designed to operate during and after certain design basis events such as earthquakes and anticipated operational occurrences.

7.7.1.6.1.3 SRM Operational Considerations The SRM system provides information to the operator. It is operated by inserting the SRM detectors into the core whenever these channels are needed, and withdrawing them, when permitted, to reduce their burnup.

7.7.1.6.2 Rod Block Monitor System 7.7.1.6.2.1 RBM Equipment Design 7.7.1.6.2.1.1 RBM Circuit Description The RBM has two channels (Figure 7.6-4). Each channel uses input signals from a number of LPRM channels. A trip signal from either RBM channel initiates a rod block. One RBM channel can be bypassed without loss of system function. The minimum number of LPRM inputs required for each RBM channel to prevent an instrument inoperative alarm is four when using four LPRM assemblies, three when using three LPRM assemblies, and two when using two LPRM assemblies.

a. Power Supply - The RBM power is received from two independent 120 V ac buses, each of which is supplied by an UPS. Each RBM is supplied by two redundant DC power supplies. Each DC power supply is supplied by one of the two 120 Vac buses.
b. Signal Conditioning - The RBM signal is generated by averaging a set of LPRM signals. The LPRM signals used depend on the control rod selected. Upon selection of a rod for withdrawal or insertion, the conditioned signals from the LPRMs around that rod will be automatically selected by the two RBM channels (Figure 7.7-14 shows examples of the four possible LPRM/selected rod assignment combinations). For typical non-edge rod, each RBM channel averages LPRM inputs from two of the four B-level and D-level detectors, and all four of the C-level detectors (see Figure 7.7-14). A-level LPRM detectors are not included in the RBM averages, but are displayed to the operator. When a rod near, but not at, the edge of the core is selected, where there are fewer than four but at least two LPRM strings around the rod, the number of detectors used by the RBM channels is either six of four depending on how many LPRM strings are available. If a detector has been bypassed in the LPRM system, that detector is automatically deleted from the RBM processing and the averaging logic is adjusted to average only the remaining detectors.

After selection of a control rod, each RBM channel calculates the average of the related LPRM detectors and calculates a gain factor that will adjust the average to 100. Thereafter, until another rod is selected, the gain factor is applied to the LPRM CHAPTER 07 7.7-40 REV. 20, SEPTEMBER 2020

LGS UFSAR average to obtain the RBM signal value. The RBM signal value is compared to RBM trip setpoints.

When a peripheral rod is selected or the RBMs associated reference APRM power signal is below the automatic bypass level, the RBM function is automatically bypassed, the rod block outputs are set to permissive, and the RBM average is set to zero.

In the operating range, the RBM signal is accurate to approximately 1% of full-scale.

7.7.1.6.2.1.2 RBM Trip Function The RBM is designed to prohibit erroneous withdrawal of a control rod during operation at core high power levels. This prevents local fuel damage under permitted bypass and/or LPRM detector chamber failure conditions and prevents local fuel damage during a single rod withdrawal error.

Local fuel damage poses no significant threat relative to radioactive release from the plant.

The RBM supplies a trip signal to the RMCS to inhibit control rod withdrawal. The trip is initiated when RBM output exceeds the rod block setpoint. The RBM has three upscale trip levels and one downscale trip level. Figure 7.7-17 illustrates the trip setpoints. Below 30 percent rated power, fuel damage cannot occur for any single control rod withdrawal; hence, the RBM system is automatically bypassed. The low trip setpoint (LTSP) is enforced between 30 percent and 65 percent rated power, the intermediate trip setpoint (ITSP) is enforced between 65 percent and 85 percent rated power, and the high trip setpoint (HTSP) is enforced between 85 percent and 100 percent rated power. The percent rated power input used to automatically select the applicable RBM trip is provided by the APRM. The RBM system is automatically bypassed if the control rod has one or more adjacent fuel bundles comprising the outer boundary of the core. The operator can bypass one of the two RBM channels at any time. Either RBM channel can inhibit control rod withdrawal.

7.7.1.6.2.1.3 RBM Bypasses The operator can bypass one of the two RBM channels at any time. Both RBM channels can be bypassed provided the following conditions are met:

for power >30% and <90%, MCPR >1.70 for power >90%, MCPR >1.40 7.7.1.6.2.1.4 RBM Redundancy Although the RBM does not perform a safety-related function, in the interest of plant economics and availability, it is designed to meet certain salient design principles of a safety system. These include the following:

a. Redundant, separate, and isolated RBM channels.
b. Redundant, separate, isolated rod selection information.
c. Independent, isolated RBM level readouts and status displays from the RBM channels.

CHAPTER 07 7.7-41 REV. 20, SEPTEMBER 2020

LGS UFSAR

d. A mechanical barrier between channels A and B of the manual bypass switch.
e. Multiple manual RBM channel bypass is prohibited by switch design.
f. Fail-safe design; loss of power initiates a rod block.
g. A trip of either RBM channel initiates a rod block.
h. Redundant APRM simulated thermal power reference signals provided to each RBM for trip level information
i. Each RBM channel has a backup reference APRM 7.7.1.6.2.1.5 RBM Testability The RBM channels are tested and calibrated with written procedures. The RBMs are functionally tested by introducing test signals into the RBM channels.

7.7.1.6.2.2 RBM Operational Considerations Trip setpoint automatically changes to the next higher rod block setpoint line.

The RBM interfaces with the following safety-related systems:

a. LPRM: Separate, isolated LPRM flux level is provided to each RBM channel.
b. APRM: Independent, separate, isolated APRM simulator thermal power reference signals are supplied to each RBM channel for trip reference.

7.7.1.6.3 Traversing Incore Probe System 7.7.1.6.3.1 TIP System Identification 7.7.1.6.3.1.1 TIP General Flux readings along the axial length of the core are obtained by fully inserting the traversing probe into one of the calibration guide tubes, then taking data as the probe is withdrawn. The analog data are available for display on a recorder or for use by the process computer. One traversing probe and its associated drive mechanism is provided for each of five groups of nine or ten incore guide tubes.

7.7.1.6.3.1.2 TIP Classification This system is a power generation system and is not related to safety.

7.7.1.6.3.1.3 TIP Reference Design CHAPTER 07 7.7-42 REV. 20, SEPTEMBER 2020

LGS UFSAR Table 7.1-2 lists reference design information. The subject instrumentation and controls system is an operational system and has no safety function. Therefore, there are no safety design differences between this system and those of the referenced design facilities. This system is functionally similar to the referenced system.

7.7.1.6.3.2 TIP Power Source The power for the subject system is supplied from an instrument ac power source.

7.7.1.6.3.3 TIP Equipment Design 7.7.1.6.3.3.1 TIP General There are five TIP machines as shown in Figure 7.6-1. The TIP machines have the following components:

a. One TIP
b. One drive mechanism
c. One indexing mechanism
d. Up to 10 incore guide tubes The system allows calibration of LPRM signals by correlating TIP signals to LPRM signals as the TIP is positioned in various radial and axial locations in the core. The guide tubes inside the reactor are divided into groups. Each group has its own associated TIP machine.

7.7.1.6.3.3.2 TIP Equipment Arrangement A TIP drive mechanism uses a gamma sensitive probe assembly attached to a flexible drive cable (Figure 7.7-10). The cable is driven from outside the drywell by a gearbox assembly. The flexible cable is contained by guide tubes that penetrate the reactor core. The guide tubes are a part of the LPRM detector assembly. The indexing mechanism allows the use of a single detector in any one of 10 different tube paths. The 10th tube is used for TIP cross calibration with the other TIP machines. The control system provides for both manual and semi-automatic operation. The electronics in the TIP panel in the control room amplifies and displays the TIP signal. Core position versus gamma flux may be plotted on an X-Y recorder in the control room, and is also provided to the Plant Monitoring System. A block diagram of the drive system is shown in Figure 7.6-3.

7.7.1.6.3.3.3 TIP Testability The TIP equipment is tested and calibrated with written procedures.

7.7.1.6.3.4 TIP Environmental Considerations The equipment and cabling located in the drywell are designed for continuous-duty up to 150F and 100% relative humidity.

7.7.1.6.3.5 TIP Operational Considerations CHAPTER 07 7.7-43 REV. 20, SEPTEMBER 2020

LGS UFSAR The TIP can be operated during reactor operation to calibrate the LPRM channels. This system has no safety setpoints.

7.7.1.7 Deleted 7.7.1.8 Reactor Water Cleanup System - Instrumentation and Controls 7.7.1.8.1 RWCU System Identification 7.7.1.8.1.1 RWCU General The purpose of the RWCU system is to provide for removal of soluble and insoluble waterborne impurities to reduce the secondary source of and radiation.

7.7.1.8.1.2 RWCU Classification This is a power generation system and is classified as not related to safety.

7.7.1.8.1.3 RWCU Reference Design Table 7.1-2 lists reference design information. This control system is an operational system and has no safety function.

7.7.1.8.2 RWCU Power Sources The RWCU system is powered from the plant instrumentation bus, with the exception of the Noble Metals Monitoring System Data Acquisition System, which is powered from a lighting panel power supply. No backup power source is necessary since the RWCU system is not a safety-related system. Adequate fuse protection is provided so that a short circuit within the system has only a local effect that can be easily corrected without interrupting reactor operation.

7.7.1.8.3 RWCU Equipment Design 7.7.1.8.3.1 RWCU General The RWCU system is described in Section 5.4.8.

7.7.1.8.3.2 RWCU Circuit Description The cleanup system is protected against overpressurization by relief valves. The ion exchange resin is protected from high temperature by temperature switches upstream of the filter/

demineralizer unit. One switch activates an alarm while a second switch closes the outboard isolation valve, which subsequently trips the cleanup pumps. The isolation valves also close automatically on a reactor low water level signal, when the SLCS is actuated, and on a "leak signal" from the leak detection system. See Section 7.3 for a detailed description of the isolation function.

A high differential pressure across the filter/demineralizer or its discharge strainer automatically isolates the units and sounds an alarm. The holding pump starts whenever there is a low flow through a filter/demineralizer.

CHAPTER 07 7.7-44 REV. 20, SEPTEMBER 2020

LGS UFSAR A sampling station is provided to obtain reactor water samples from the inlet and outlet of both filter/demineralizers.

Flow, pressure, temperature, and conductivity are recorded or indicated on a panel in the control room. Instrumentation and controls for backwashing and precoating the filter/demineralizers are on a local panel outside the drywell. Alarms are sounded in the control room to alert the operator to abnormal conditions.

7.7.1.8.3.3 RWCU Testability Because the RWCU system is usually in service during plant operation, satisfactory performance is demonstrated without the need for any special inspection or testing beyond that specified in the manufacturer's instructions.

7.7.1.8.4 RWCU Environmental Considerations The RWCU system is not required for safety purposes nor is it required to operate after the DBA.

The RWCU system is required to operate in the normal plant environment for power generation purposes only.

RWCU instrumentation and controls located in the RWCU equipment area are subject to the environment described in Section 3.11.

7.7.1.8.5 RWCU Operational Considerations 7.7.1.8.5.1 General Information The RWCU system instrumentation and controls are not required for safe operation of the plant.

They provide a means of monitoring parameters of the system and protecting the system.

7.7.1.8.5.2 RWCU Reactor Operator Information See drawing M-44.

7.7.1.8.5.3 RWCU Setpoints No safety-related setpoints are associated with the RWCU system.

7.7.1.9 Process Radiation Monitoring Systems - Instrumentation and Controls 7.7.1.9.1 South Stack Effluent Radiation Monitoring System 7.7.1.9.1.1 SSE-RMS Identification The objective of the SSE-RMS is to monitor the quantity of radioactive gases, particulates, and iodines emitted from the south stack. Inasmuch as the south stack is a main point of potential escape of radioactive material to the environment, these data are used for the generation of Regulatory Guide 1.21 reports.

7.7.1.9.1.2 SSE-RMS Classification CHAPTER 07 7.7-45 REV. 20, SEPTEMBER 2020

LGS UFSAR This system is a power generation system and is classified as not related to safety. Inasmuch as the south stack will be isolated if there is an accident, the capability to monitor accident and postaccident conditions is not required.

7.7.1.9.1.3 SSE-RMS Reference Design Table 7.1-2 lists reference design information. The SSE-RMS is an operational monitoring system and has no safety function.

7.7.1.9.1.4 SSE-RMS Power Sources Power is provided by the 120 V ac instrument bus. Safety-related power is not required for this application.

7.7.1.9.1.5 SSE-RMS Equipment Design 7.7.1.9.1.5.1 SSE-RMS General The south stack consists of two discrete ducts for the two units. Two isokinetic, representative samples of air are drawn from each duct in accordance with Regulatory Guide 1.21 and ANSI N13.1. Consequently, four independent radiation monitoring subsystems are provided. Each of these subsystems is provided with three channels for monitoring radioactive particulates, iodines, and noble gases in the stack effluents.

Each subsystem consists of three detectors, a microprocessor, and a local control and readout module. All subsystems are provided with high, high-high, and downscale alarm outputs.

7.7.1.9.1.5.2 SSE-RMS Circuit Description Each monitoring subsystem has two upscale trips for high and high-high radiation and a downscale trip that indicates instrument trouble. These trips are actuated by the above conditions in any of the subsystem channels. These trips annunciate, but provide no control action. The trip circuits are set so that loss of power causes an alarm. The range of each channel is six decades and is selected to cover both normal and possible maximum radiation levels in the south stack effluents.

The monitors are located on the roof of the reactor enclosure, where the background radiation environment is minimal. These monitors are interfaced to the RMMS (Section 11.5.6) to allow indication, trending, and remote control at RMDS display consoles in the control room. Remote controls are provided for purging the noble gas chamber. Controls are also provided for remotely resetting the alarm setpoints and for checking detector operability.

7.7.1.9.1.5.3 SSE-RMS Testability Check sources are provided for each detector to verify operability of each channel. Access to the controls and setpoints is restricted by the operator's key and coded input signals.

7.7.1.9.1.6 SSE-RMS Environmental Considerations CHAPTER 07 7.7-46 REV. 20, SEPTEMBER 2020

LGS UFSAR The monitoring racks, microprocessors, and isokinetic sampling systems are designed to meet the environmental conditions on the reactor enclosure roof. These systems are protected by a room provided with ventilation, heating, and cooling. Radiation background in this location is minimal.

See Section 3.11 for the environmental conditions of the equipment locations.

7.7.1.9.1.7 SSE-RMS Operational Considerations 7.7.1.9.1.7.1 SSE-RMS General Information Annunciator outputs in the control room are shared for all subsystems. Redundancy and independence ensure continuous monitoring during periods of equipment checking, servicing, and shutdown. Fixed filters provide for continuous accumulation of radioactive samples that can be analyzed in the counting room. Digitized outputs of all three channels, as well as flow measurements, are provided as input data to the RMMS for the generation of Regulatory Guide 1.21 reports.

7.7.1.9.1.7.2 SSE-RMS Reactor Operator Information Processed inputs to the control room provide the reactor operator with early warnings of anomalous changes in radioactive effluents. This information can be used as a guide for remedial action.

7.7.1.9.2 Radwaste Equipment Rooms Ventilation Exhaust Radiation Monitoring System 7.7.1.9.2.1 RERV-RMS Identification The RERV-RMS is an off-line process radiation monitoring system that measures the concentration of radioactive material in the ventilation duct from the charcoal offgas treatment compartment. The purpose of this monitoring system is to detect leakage of radioactive noble gases from the charcoal beds. Such leakage could escape to the north stack.

7.7.1.9.2.2 RERV-RMS Classification The RERV-RMS is a power generation system and is not related to safety.

7.7.1.9.2.3 RERV-RMS Reference Design Table 7.1-2 lists reference design information. The RERV-RMS is designed to alarm in response to excessive concentrations of noble gases in the effluents released to the north stack, indicating leakage from the charcoal offgas treatment system.

7.7.1.9.2.4 RERV-RMS Power Sources Power is supplied to the monitor by the 120 V ac uninterruptible bus B. The pumps and solenoid valves are supplied from the 120 V ac local power. The recorder and purge controls are supplied from the 120 V ac instrument bus.

7.7.1.9.2.5 RERV-RMS Equipment Design 7.7.1.9.2.5.1 RERV-RMS General CHAPTER 07 7.7-47 REV. 20, SEPTEMBER 2020

LGS UFSAR The RERV-RMS is a standard off-line process radiation monitoring system. It draws an air sample from a ventilation duct, passes this sample through a chamber containing a radiation detector, and then returns the sample to the duct.

7.7.1.9.2.5.2 RERV-RMS Circuit Description The RERV-RMS is provided with a gamma-sensitive detector and converter that transmits its signal to a rate meter located in the auxiliary equipment room. This rate meter indicates the concentration of radioactivity in the air sample and transmits both a signal to a recorder in the control room and upscale/downscale alarm signals to annunciators in the control room. No control trip capabilities are provided. A flow indicator trips an annunciator if there is a malfunction of the sampling pump.

7.7.1.9.2.6 RERV-RMS Testability Test signals are fed into the monitor via the control module in the auxiliary equipment room. A check source is provided for secondary calibration. Purge controls are also provided.

7.7.1.9.2.7 RERV-RMS Environmental Considerations The monitoring rack is designed to meet the environmental conditions in the radwaste enclosure.

The control module and recorder are in the environment of the auxiliary equipment room and control room. See Section 3.11 for the environmental conditions of the equipment locations.

7.7.1.9.2.8 RERV-RMS Operational Considerations 7.7.1.9.2.8.1 RERV-RMS General Annunciator outputs of the RERV-RMS alarm signals are provided in the control room. Fixed filters for particulates and iodines provide for continuous accumulation of radioactive materials that can be analyzed in the counting room. However, very few, if any, particulates or iodines are expected in this application.

7.7.1.9.2.8.2 RERV-RMS Reactor Operator Information Inputs to the control room provide the operator with an indication of leakage from the charcoal beds. If there is such an indication, remedial action can be taken.

7.7.1.9.3 Charcoal Treatment System Process Exhaust Radiation Monitoring System 7.7.1.9.3.1 CTSP-RMS Identification The CTSP-RMS is identical with the system discussed in Section 7.7.1.9.2, except that the monitored sample is drawn from the charcoal offgas treatment system effluent pipe. The purpose of this monitoring system is to provide ongoing records and alarm capability for indicating unacceptable concentrations of noble gases in the effluents being released to the north stack.

7.7.1.9.3.2 CTSP-RMS Classification The CTSP-RMS is a power generation system and is not related to safety.

CHAPTER 07 7.7-48 REV. 20, SEPTEMBER 2020

LGS UFSAR 7.7.1.9.3.3 CTSP-RMS Reference Design Table 7.1-2 lists reference design information. The CTSP-RMS is designed to alarm in response to excessive concentrations of noble gases in effluents from the charcoal offgas treatment system.

7.7.1.9.3.4 CTSP-RMS Power Sources Power is supplied to the monitors by the 120 V ac uninterruptible bus A. The pumps and solenoid valves are supplied from the 120 V ac local power. The recorder and purge controls are supplied from the 120 V ac instrument bus.

7.7.1.9.3.5 CTSP-RMS Equipment Design 7.7.1.9.3.5.1 CTSP-RMS General The CTSP-RMS is a standard off-line process radiation monitoring system. It draws a gas sample from a process pipe, passes the sample through a chamber containing a radiation detector, and then returns the sample to the pipe.

7.7.1.9.3.5.2 CTSP-RMS Circuit Description The CTSP-RMS is provided with a gamma-sensitive detector and converter that transmits its signal to a rate meter located in the auxiliary equipment room. This rate meter indicates the concentration of radioactive material in the gas sample and transmits both a signal to a recorder in the control room and upscale/downscale alarm signals to annunciators in the control room. No control trip capabilities are provided. A flow indicator trips an annunciator if the sampling pump malfunctions.

7.7.1.9.3.6 CTSP-RMS Testability Test signals are fed into the monitor via the control module in the auxiliary equipment room. A check source is provided for secondary calibration. Purge controls are also provided.

7.7.1.9.3.7 Environmental Considerations The monitoring rack is designed to meet the environmental conditions in the radwaste enclosure.

The control module and recorder are in the environment of the auxiliary equipment room and control room. See Section 3.11 for the environmental conditions of the equipment locations.

7.7.1.9.3.8 CTSP-RMS Operational Considerations 7.7.1.9.3.8.1 CTSP-RMS General Annunciator outputs of the CTSP-RMS alarm signals are provided in the control room. Fixed filters for particulates and iodines provide for continuous accumulation of radioactive materials that can be analyzed in the counting room. However, very few, if any, particulates or iodines are expected in this application.

7.7.1.9.3.8.2 CTSP-RMS Reactor Operator Information CHAPTER 07 7.7-49 REV. 20, SEPTEMBER 2020

LGS UFSAR Inputs to the control room provide the operator with an indication of the effectiveness of the charcoal offgas treatment system in arresting releases of radioactivity to the north stack. If there are unacceptable inputs, remedial action can be taken.

7.7.1.9.4 Recombiner Rooms and Hydrogen Analyzer Compartments Exhaust Radiation Monitoring System 7.7.1.9.4.1 RRHAC-RMS Identification This off-line process radiation monitoring system detects the presence of radioactive material escaping to the north stack from the recombiner compartments, hydrogen analyzer compartments, and equipment drain sump vent. If there is a high radiation alarm, the manifolded sample line leading from the main ducts can be isolated, and each branch of this line can be checked separately to determine the exact source of leakage.

7.7.1.9.4.2 RRHAC-RMS Classification This monitoring system is a power generation system and is not related to safety.

7.7.1.9.4.3 RRHAC-RMS Reference Design Table 7.1-2 lists reference design information. This system is designed to alarm in response to significant concentrations of gross radioactivity in the duct effluents from the recombiner compartments, hydrogen analyzer compartments, or equipment drain sump vent. Capability is provided to locate the source of this radioactivity.

7.7.1.9.4.4 RRHAC-RMS Power Source Power is supplied to the monitor by the 120 V ac uninterruptible bus B. The pumps and solenoid valves are supplied from the 120 V ac local power. The recorder and purge controls are supplied from the 120 V ac instrument bus.

7.7.1.9.4.5 RRHAC-RMS Equipment Design 7.7.1.9.4.5.1 RRHAC-RMS General This standard off-line radiation monitoring system draws samples from two exhaust ventilation ducts, combines them, passes the samples through a chamber containing a radiation detector, and then returns the sample to the exhaust duct. The operator can select the branch sample line he chooses to check using a set of solenoid valves. In this way, seven discrete sources of radioactivity can be monitored with a single instrument system.

7.7.1.9.4.5.2 RRHAC-RMS Circuit Description This monitoring system is provided with a gamma-sensitive detector and converter that transmits its signal to a rate meter located in the auxiliary equipment room. This rate meter indicates radioactivity in the air sample and transmits both a signal to a recorder in the control room and upscale/downscale alarm signals to an annunciator in the control room. No control trip capability is provided. A flow indicator trips an annunciator if there is a malfunction of the sampling pump. A CHAPTER 07 7.7-50 REV. 20, SEPTEMBER 2020

LGS UFSAR hand selector switch in the control room provides for determination of which of the following sources to monitor:

a. Main exhaust ducting
b. Recombiner compartment A
c. Recombiner compartment B
d. Hydrogen analyzer 1A, Hydrogen analyzer 2A
e. Hydrogen analyzer 1B, Hydrogen analyzer 2B
f. Hydrogen analyzer 1C, Hydrogen analyzer 2C
g. Equipment drain sump The hand selector switch is normally set for the main exhaust duct.

7.7.1.9.4.6 RRHAC-RMS Testability Test signals are fed into the monitor via the control module in the auxiliary equipment room. A check source is provided for secondary calibration. Purge controls are also provided.

7.7.1.9.4.7 RRHAC-RMS Environmental Considerations The monitoring rack is designed to meet the environmental conditions in the control structure. The control module and recorder are in the environment of the auxiliary equipment room and control room. See Section 3.11 for the environmental conditions of the equipment locations.

7.7.1.9.4.8 RRHAC-RMS Operational Considerations 7.7.1.9.4.8.1 RRHAC-RMS General Annunciator outputs of alarms from this monitoring system are provided in the control room. Fixed filters for particulates and iodines provide for continuous accumulation of radioactive materials that can be analyzed in the counting room. However, very few, if any, particulates or iodines are expected in this application.

7.7.1.9.4.8.2 RRHAC-RMS Reactor Operator Information Inputs to the control room provide the operator with an indication of any evidence of leakage from the sources tabulated above. When such evidence is perceived, remedial action can be taken.

7.7.1.9.5 Steam Exhauster Discharge and Vacuum Pump Exhaust Radiation Monitoring System 7.7.1.9.5.1 SEDVP-RMS Identification The SEDVP-RMS is an off-line radiation monitoring system that detects the presence of short half-life radioactive isotopes (mostly N-16) leaking from the steam seals of the turbine. The purpose of this system is to indicate the existence of such leakage. The monitor also provides the capability to sample the mechanical vacuum pump exhaust when this equipment is operated.

CHAPTER 07 7.7-51 REV. 20, SEPTEMBER 2020

LGS UFSAR 7.7.1.9.5.2 SEDVP-RMS Classification The SEDVP-RMS is a power generation system and is not related to safety.

7.7.1.9.5.3 SEDVP-RMS Reference Design Table 7.1-2 lists reference design information. The SEDVP-RMS is designed to alarm in response to significant evidence of radioactive isotopes escaping from the steam seals of the turbine or the mechanical vacuum pump exhaust.

7.7.1.9.5.4 SEDVP-RMS Power Source Power is supplied to the monitors by the 120 V ac uninterruptible bus A. The pumps and solenoid valves are supplied from the 120 V ac local power. The recorder and purge controls and supplied from the 120 V ac instrument bus.

7.7.1.9.5.5 SEDVP-RMS Equipment Design 7.7.1.9.5.5.1 SEDVP-RMS General The SEDVP-RMS is a standard off-line process radiation monitoring system. It draws a combined gas sample from the exhaust line leading from the steam packing exhaust condenser and mechanical vacuum pump exhaust, passes the sample through a chamber containing a radiation detector, and then returns the sample to the exhaust line. The source of detected radioactivity can be identified by manually isolating one of the sample lines.

7.7.1.9.5.5.2 SEDVP-RMS Circuit Description The SEDVP-RMS is provided with a gamma-sensitive detector and converter that transmits its signal to a rate meter in the auxiliary equipment room. This rate meter indicates radioactivity in the gas sample and transmits a signal to a recorder in the control room. No control trip capabilities are provided.

7.7.1.9.5.5.3 SEDVP-RMS Testability Test signals are fed into the monitor via the control module in the auxiliary equipment room. A check source is provided for secondary calibration. Purge controls are also provided.

7.7.1.9.5.5.4 SEDVP-RMS Environmental Considerations The monitoring rack is designed to meet the environmental conditions in the turbine enclosure.

The control module and recorder are in the environmental conditions of the auxiliary equipment room and control room. See Section 3.11 for the environmental conditions of the equipment locations.

7.7.1.9.5.5.5 SEDVP-RMS Operational Considerations 7.7.1.9.5.5.5.1 SEDVP-RMS General Fixed filters for particulates and iodines continuously accumulate radioactive solids that can then be analyzed in the counting room. However, very few, if any, particulates or iodines are expected CHAPTER 07 7.7-52 REV. 20, SEPTEMBER 2020

LGS UFSAR in this application. A cooler and moisture separator are provided in the sample line to prevent damage to the detector as a result of excessive sample temperature.

7.7.1.9.5.5.5.2 SEDVP-RMS Reactor Operator Information A recorder provides the operator with an indication of defectiveness of the turbine steam seals or excessive radiation in the vacuum pump exhaust.

7.7.1.9.6 Radwaste Enclosure and Chem. Lab. Expansion Ventilation Exhaust Radiation Monitoring System 7.7.1.9.6.1 REV-RMS Identification The REV-RMS is an off-line instrument system that measures the concentration of gross radioactive effluents in the ventilation exhaust duct leading from the radwaste enclosure and chem.

lab. expansion to the north stack. The purpose of the system is to provide ongoing records and alarm capability for gaseous radioactivity in effluents released to the north stack from the radwaste enclosure.

7.7.1.9.6.2 REV-RMS Classification The REV-RMS is a power generation system and is not related to safety.

7.7.1.9.6.3 REV-RMS Reference Design Table 7.1-2 lists reference design information. The REV-RMS is designed to alarm in response to excessive concentrations of gross radioactivity in the ventilation exhaust from the radwaste enclosure.

7.7.1.9.6.4 REV-RMS Power Sources Power is supplied to the monitors by the 120 V ac uninterruptible bus A. The pumps and solenoid valves are supplied from the 120 V ac local power. The recorder and purge controls are supplied from the 120 V ac instrument bus.

7.7.1.9.6.5 REV-RMS Equipment Design 7.7.1.9.6.5.1 REV-RMS General The REV-RMS is a standard off-line process radiation monitoring system. It draws an air sample from a ventilation duct, passes the sample through a chamber containing a radiation detector, and then returns the sample to the duct.

7.7.1.9.6.5.2 REV-RMS Circuit Description The REV-RMS is provided with a gamma-sensitive detector and converter that transmits its signal to a rate meter in the auxiliary equipment room. This rate meter indicates radioactivity in the air sample and transmits both a signal to a recorder in the control room and upscale/downscale alarm signals to an annunciator in the control room. No control trip capabilities are provided. A flow indicator trips an annunciator if the sampling pump malfunctions.

7.7.1.9.6.6 REV-RMS Testability CHAPTER 07 7.7-53 REV. 20, SEPTEMBER 2020

LGS UFSAR Test signals are fed into the monitor via the control module in the auxiliary equipment room. A check source is provided for secondary calibration. Purge controls are also provided.

7.7.1.9.6.7 REV-RMS Environmental Considerations The monitoring rack is designed to meet the environmental conditions in the control structure. The control module and recorder are in the environment of the auxiliary equipment room and control room. See Section 3.11 for the environmental conditions of the equipment locations.

7.7.1.9.6.8 REV-RMS Operational Considerations 7.7.1.9.6.8.1 REV-RMS General Annunciator outputs of the REV-RMS alarm signals are provided in the control room. Fixed filters for particulates and iodines continuously accumulate radioactive solids that can be analyzed in the counting room. However, an insignificant quantity of particulates and iodines is expected.

7.7.1.9.6.8.2 REV-RMS Reactor Operator Information Inputs to the control room provide the operator with evidence of the escape of airborne radioactive material from the radwaste enclosure. This can be used as a basis for diagnostic investigation.

7.7.1.9.7 Air Ejector Offgas Effluent Radiation Monitoring System 7.7.1.9.7.1 AEO-RMS Identification The AEO-RMS measures the concentration of gross radioactivity in effluents from the steam jet air ejector downstream of the recombiner.

7.7.1.9.7.2 AEO-RMS Classification This monitoring system is a power generation system and is not related to safety.

7.7.1.9.7.3 AEO-RMS Reference Design Table 7.1-2 lists reference design information. This system serves the following functions:

a. It provides an ongoing record and indication of concentrations of radioactivity in air ejector offgas.
b. It provides the capability to detect fuel element contamination in the air ejector offgas.
c. It provides alarm annunciation capability.
d. It provides a grab sampling capability for detailed analysis in the counting room.

7.7.1.9.7.4 AEO-RMS Power Sources CHAPTER 07 7.7-54 REV. 20, SEPTEMBER 2020

LGS UFSAR Power is supplied to the two log monitors by the 120 V ac uninterruptible buses A and B. The linear monitor is supplied from a 20 V dc bus which is supplied from the 120 V ac uninterruptible Bus A. The pump and solenoid valves for grab sampling are supplied from 120 V ac local power.

The recorder and purge controls are supplied from the 120 V ac instrument bus.

7.7.1.9.7.5 AEO-RMS Equipment Design 7.7.1.9.7.5.1 AEO-RMS General This monitoring system is designed to provide indication and alarm if there are excessive discharges of radioactivity and to detect fuel element contamination in the air ejector effluents. It also provides grab sampling equipment for diagnostic purposes.

7.7.1.9.7.5.2 AEO-RMS Circuit Description Two ionization chamber-type detectors with sensitivity ranges 1 mR/hr to 106 mR/hr are used to detect gross radioactive concentration in the air ejector effluents. In addition, a linear response channel is provided as an expanded scale device to aid in detecting fuel element contamination.

The three channels provide outputs to recorders in the control room. A flow indicating switch provides annunciation if there is a high or low flow condition. By means of a local hand-selector switch, sampling pump controls, and solenoid valves controls, grab samples can be taken. The three rate meters are located in the auxiliary equipment room. Outputs from the log rate meters are fed to annunciators in the control room. Monitor purge controls are located in the control room.

7.7.1.9.7.6 AEO-RMS Testability A built-in source of adjustable current is provided to simulate sensor input to each radiation monitor for test purposes.

7.7.1.9.7.7 AEO-RMS Environmental Considerations The monitoring rack is designed to meet the environmental conditions in the control structure. The control module and recorders are in the environment of the auxiliary equipment and control rooms.

See Section 3.11 for the environmental conditions of the equipment locations.

7.7.1.9.7.8 AEO-RMS Operational Considerations 7.7.1.9.7.8.1 AEO-RMS General Annunciator outputs and control trip outputs are provided by this monitoring system. Controls are both local and located in the auxiliary equipment room or control room. Gross radioactivity readings can be further diagnosed using grab samples.

7.7.1.9.7.8.2 AEO-RMS Reactor Operator Information Inputs to the control room apprise the operator of concentrations of radioactivity in the air ejector effluents, including indications of fuel element deterioration. Diagnostic capability is available.

7.7.1.9.8 Primary Containment Leak Detector Radiation Monitor CHAPTER 07 7.7-55 REV. 20, SEPTEMBER 2020

LGS UFSAR 7.7.1.9.8.1 PCLD-RMS Identification The objective of the PCLD-RMS is to indicate and record radiation levels of the primary containment atmosphere. The instrument provides alarm capability in response to a rapid increase in this radiation level. Alarm setpoints are discussed in Section 5.2.5.2.1.5.3.

7.7.1.9.8.2 PCLD-RMS Classification The PCLD-RMS is a power generation system and is not related to safety.

7.7.1.9.8.3 PCLD-RMS Reference Design Table 7.1-2 lists reference design information. The PCLD-RMS does not have a safety function.

7.7.1.9.8.4 PCLD-RMS Power Sources Power is supplied to the monitor and recorder by a 120 V ac uninterruptible bus A. The pumps and solenoid valves are supplied from the 120 V ac local power. The purge control circuit is supplied from the 120 V ac instrument bus.

7.7.1.9.8.5 PCLD-RMS Equipment Design 7.7.1.9.8.5.1 PCLD-RMS General A gas sample is drawn from the primary containment and is passed through the PCLD-RMS to measure the concentration of noble gases. Alarm capability is provided. Outputs are transmitted to a control module in the auxiliary equipment room, and to a recorder in the control room.

7.7.1.9.8.5.2 PCLD-RMS Circuit Description The monitor channel has an upscale trip that indicates a rate of increase of radioactivity in the primary containment. A downscale trip is also provided to indicate instrument trouble. These trips annunciate, but provide no control action. The trip circuits are set so that loss of power causes an alarm. The five decade range is selected to cover both normal and maximum radiation levels (short of an accident resulting in shutdown) in the primary containment. If an accident occurs, the PCLD-RMS will be isolated, and its functions will cease.

The PCLD-RMS monitoring rack is adjacent to the primary containment. Outputs of this instrument are transmitted to the auxiliary equipment room and the control room. Remote controls purge the noble gas chamber, reset the alarm setpoints, and check the calibrations.

7.7.1.9.8.5.3 PCLD-RMS Testability Test signals are fed into the monitor via the control module in the auxiliary equipment room. Check sources are provided for secondary calibration of each channel.

7.7.1.9.8.6 PCLD-RMS Environmental Considerations The monitoring rack is designed to be suitable for the environmental conditions adjacent to the primary containment. Control modules and recorders are subjected to the environment of the CHAPTER 07 7.7-56 REV. 20, SEPTEMBER 2020

LGS UFSAR auxiliary equipment room and control room. See Section 3.11 for the environmental conditions of the equipment locations.

7.7.1.9.9 Hot Maintenance Shop Ventilation Exhaust Radiation Monitoring System 7.7.1.9.9.1 HMS-RMS Identification The objective of the HMS-RMS is to indicate and record the quantity of radioactive material that escapes from the hot maintenance shop ventilation exhaust duct. Filters are provided in this duct, and sample air is drawn from a station downstream of the filters. Consequently, the quantity of radioactive material in the sample is normally negligible.

7.7.1.9.9.2 HMS-RMS Classification This system is a power generation system and is not related to safety.

7.7.1.9.9.3 HMS-RMS Reference Design Table 7.1-2 lists reference design information.

7.7.1.9.9.4 HMS-RMS Power Sources The power source of the HMS-RMS is the 120 V ac lighting power.

7.7.1.9.9.5 HMS-RMS Equipment Design 7.7.1.9.9.5.1 HMS-RMS General The system consists of two channels: one for monitoring particulate effluents, and the other for monitoring iodine effluents. Radioactive material in the gas sample is accumulated on filters where it is sensed by scintillator-type detectors. Output signals are transmitted to a microprocessor and then retransmitted to the RMMS (Section 11.5.6) for the purpose of insertion into Regulatory Guide 1.21 reports.

7.7.1.9.9.5.2 HMS-RMS Circuit Description Each channel has two upscale trips that indicate high and high-high radiation and a downscale trip that indicates instrument trouble. These trips sound local alarms, but cause no control action.

There are three trip indicating lights per channel on the control module. The range of each channel is six decades and is selected to cover both normal and possible maximum levels at the monitoring location. The system is self-contained, and all outputs are indicated on the local control module and RMMS.

7.7.1.9.9.5.3 HMS-RMS Testability Check sources are provided to verify operability of each channel.

7.7.1.9.9.5.4 HMS-RMS Environmental Considerations The system is designed to function within the environmental range of hot shop conditions.

CHAPTER 07 7.7-57 REV. 20, SEPTEMBER 2020

LGS UFSAR 7.7.1.9.10 Liquid Radwaste Discharge Radiation Monitoring System 7.7.1.9.10.1 LRD-RMS Identification The LRD-RMS is an off-line process radiation monitoring system that measures the concentration of radioactive material in the liquid radwaste discharge to the cooling tower blowdown line. This liquid radwaste discharge includes discharges from the laundry drain, chemical wastes, floor drains, and oily waste treatment water sump. If unacceptable radioactive concentrations are detected in these discharges, valves in the discharge line automatically close, isolating the radwaste discharge system for the cooling tower blowdown line.

7.7.1.9.10.2 LRD-RMS Classification The LRD-RMS is a power generation system and is not related to safety. However, control trip capability is provided to prevent the release of radioactive contamination to the environment.

7.7.1.9.10.3 LRD-RMS Reference Design Table 7.1-2 lists reference design information. The LRD-RMS is designed to trip the isolation valves in the radwaste discharge line and to annunciate an alarm in the control room if there is an unacceptable concentration of radioactivity in the radwaste discharge. Data are also transmitted to a recorder in the radwaste control room.

7.7.1.9.10.4 LRD-RMS Power Sources Power is supplied to the LRD-RMS from the 120 V ac uninterruptible bus A. Recorder power is supplied by the instrument bus in the radwaste control room. Power to the sample pump is supplied from the 120 V ac local bus.

7.7.1.9.10.5 LRD-RMS Equipment Design 7.7.1.9.10.5.1 LRD-RMS General The LRD-RMS is a standard off-line process radiation monitoring system for liquids. It draws a sample from the radwaste discharge pipe, passes the sample through a chamber containing a radiation detector, and then returns the sample to the pipe.

7.7.1.9.10.5.2 LRD-RMS Circuit Description The LRD-RMS has a gamma-sensitive detector and converter that transmits its signal to a rate meter located in the auxiliary equipment room. This rate meter indicates the concentration of radioactivity in the liquid sample and transmits a signal to a recorder in the radwaste control room and an upscale/downscale trip signal to the annunciator in the main control room. The upscale trip signal also actuates the trip relays to the radwaste discharge valves. A flow indicator switch trips an annunciator if the sample pump malfunctions.

7.7.1.9.10.6 LRD-RMS Testability CHAPTER 07 7.7-58 REV. 20, SEPTEMBER 2020

LGS UFSAR Test signals are fed into the monitor via the control module in the auxiliary equipment room. A check source is provided for secondary calibration.

7.7.1.9.10.7 LRD-RMS Environmental Considerations Off-line monitoring was selected for this application in preference to in-line monitoring because of the problems involved in removing crud buildup and in decontamination. The monitoring rack is designed to meet the environmental conditions in the radwaste enclosure. The control module is in the environment of the auxiliary equipment room, and the recorder is in the environment of the radwaste control room. See Section 3.11 for the environmental conditions of the equipment locations.

7.7.1.9.10.8 LRD-RMS Operational Considerations 7.7.1.9.10.8.1 LRD-RMS General Control trip signals close the radwaste discharge valves and alarm in the control room. Operation of the system is automatic, requiring minimum attention.

7.7.1.9.10.8.2 LRD-RMS Reactor Operator Information Inputs to the radwaste control room and main control room provide the operators with an indication of excessive release of radioactive material via the radwaste liquids. Such releases are terminated automatically, after which remedial action can be taken.

7.7.1.9.11 Service Water Radiation Monitoring System 7.7.1.9.11.1 SW-RMS Identification The SW-RMS is an off-line process radiation monitoring system that measures the concentration of radioactive material in the plant service water. The SW-RMS draws a sample from the main discharge pipe to the cooling tower. Plant service water is used to cool such normally nonradioactive areas as air compressors, turbine auxiliary systems, and pump bearings. It also cools the containment enclosure closed cooling water system through a heat exchanger. An increase in the radiation level of the service water stream may indicate a leak into the system from a contaminated stream.

7.7.1.9.11.2 SW-RMS Classification The SW-RMS is a power generation system and is not related to safety.

7.7.1.9.11.3 SW-RMS Reference Design Table 7.1-2 lists reference design information. The SW-RMS is designed to provide an alarm annunciation in the control room when a significant concentration of radioactive contamination is detected in the service water. Data are also transmitted to a control room recorder.

7.7.1.9.11.4 SW-RMS Power Sources CHAPTER 07 7.7-59 REV. 20, SEPTEMBER 2020

LGS UFSAR Power is supplied to the SW-RMS from the 120 V ac uninterruptible bus B. Power is supplied to the recorder from the 120 V ac instrument bus. The sample pump is powered by the 120 V ac local bus.

7.7.1.9.11.5 SW-RMS Equipment Design 7.7.1.9.11.5.1 SW-RMS General The SW-RMS is a standard off-line process radiation monitoring system for liquids. It draws a sample from the service water discharge pipe, passes the sample through a shielded chamber containing a radiation detector, and then returns the sample to the pipe.

7.7.1.9.11.5.2 SW-RMS Circuit Description The SW-RMS has a gamma-sensitive detector and converter that transmits a signal to a rate meter in the auxiliary equipment room. This rate meter indicates the concentration of radioactivity in the water sample and transmits upscale/downscale signals to the control room annunciator. It also transmits data to the control room recorder. A flow indicator switch trips an annunciator if the sample pump malfunctions.

7.7.1.9.11.6 SW-RMS Testability Test signals are fed into the monitor via the control module in the auxiliary equipment room. A check source is provided for secondary calibration.

7.7.1.9.11.7 SW-RMS Environmental Considerations The monitoring rack is designed to meet the environmental conditions of the turbine enclosure.

The control module is in the environment of the auxiliary equipment room, and the recorder is in the environment of the control room. See Section 3.11 for the environmental conditions of the equipment locations.

7.7.1.9.11.8 SW-RMS Operational Considerations 7.7.1.9.11.8.1 SW-RMS General Upscale/downscale trip signals actuate the annunciator in the control room. The monitoring system does not have a control trip function. Radiation data are continuously recorded in the control room.

7.7.1.9.11.8.2 SW-RMS Reactor Operator Information The SW-RMS provides the operator with information on the concentration of radioactivity in the service water leaving the plant. If this concentration is unacceptable, remedial action can be taken.

7.7.1.9.12 Reactor Enclosure Cooling Water Radiation Monitoring System 7.7.1.9.12.1 RECW-RMS Identification CHAPTER 07 7.7-60 REV. 20, SEPTEMBER 2020

LGS UFSAR The RECW-RMS is an off-line process radiation monitoring system that measures the concentration of radioactive material in the RECW system. The RECW-RMS draws a sample from the pipe downstream of the cleanup nonregenerative heat exchangers, the reactor recirculation pump seal, and motor oil coolers. This cooling system constitutes a closed circuit and, consequently, no radioactive material in this circuit could be discharged to the environment. The reactor enclosure closed cooling water system cools potentially contaminated areas such as the nonregenerative heat exchanger, recirculation pumps, and various sample coolers. Radioactivity detected in this system indicates leakage in one or more of the heat exchangers.

7.7.1.9.12.2 RECW-RMS Classification The RECW-RMS is a power generation system and is not related to safety.

7.7.1.9.12.3 RECW-RMS Reference Design Table 7.1-2 lists reference design information. The RECW-RMS is designed to provide an alarm annunciation in the control room when a significant concentration of radioactive contamination is detected in the cooling water. Data are also transmitted to a recorder in the control room.

7.7.1.9.12.4 RECW-RMS Power Sources Power is supplied to the RECW-RMS from the 120 V ac uninterruptible bus B. Power is supplied to the recorder from the 120 V ac instrument bus. The sample pump is powered by the 120 V ac local bus.

7.7.1.9.12.5 RECW-RMS Equipment Design 7.7.1.9.12.5.1 RECW-RMS General The RECW-RMS is a standard off-line process radiation monitoring system for liquids. It draws a sample from the cooling system pipe, passes the sample through a shielded chamber containing a radiation detector, and then returns the sample to the pipe.

7.7.1.9.12.5.2 RECW-RMS Circuit Description The RECW-RMS has a gamma-sensitive detector and converter that transmits a signal to a rate meter in the auxiliary equipment room. This rate meter indicates radioactivity in the water sample and transmits signals to the recorder and annunciator in the control room. A flow indicator switch trips an annunciator if the sample pump malfunctions.

7.7.1.9.12.6 RECW-RMS Testability Test signals are fed into the monitor via the control module in the auxiliary equipment room. A check source is provided for secondary calibration.

7.7.1.9.12.7 RECW-RMS Environmental Considerations The monitoring rack is designed to meet the environmental conditions of the reactor enclosure.

The control module is in the environment of the auxiliary equipment room, and the recorder is in CHAPTER 07 7.7-61 REV. 20, SEPTEMBER 2020

LGS UFSAR the environment of the control room. See Section 3.11 for the environmental conditions of the equipment locations.

7.7.1.9.12.8 RECW-RMS Operational Considerations 7.7.1.9.12.8.1 RECW-RMS General Upscale/downscale trip signals actuate the annunciator in the control room. The monitoring system provides no control trip function. Radiation data are continuously recorded in the control room.

7.7.1.9.12.8.2 RECW-RMS Reactor Operator Information The RECW-RMS provides a backup for the service water radiation monitor. The RECW-RMS detects radiation only if radiation is present in the service water and if leaks occur between the service water system and the RECW system.

7.7.1.10 Area Radiation Monitoring System - Instrumentation and Controls 7.7.1.10.1 ARMS Identification 7.7.1.10.1.1 ARMS General The ARMS indicates and records gamma radiation levels in areas where radioactive material is present, stored, handled, or inadvertently introduced.

7.7.1.10.1.2 ARMS Classification This is a power generation system and is classified as not related to safety.

7.7.1.10.1.3 ARMS Reference Design Table 7.1-2 lists reference design information. The ARMS instrumentation and control system is an operational monitoring system and has no safety function.

7.7.1.10.2 ARMS Power Sources The power source for the ARMS is the 120 V ac instrument bus.

7.7.1.10.3 ARMS Equipment Design 7.7.1.10.3.1 ARMS General The ARMS is described in Section 12.3.4. Each channel consists of a combined sensor and converter unit; a combined indicator and trip unit; a shared power supply; and a shared multipoint recorder. Each channel also has a local audio alarm auxiliary unit.

7.7.1.10.3.2 ARMS Circuit Description CHAPTER 07 7.7-62 REV. 20, SEPTEMBER 2020

LGS UFSAR Each monitor has an upscale trip that indicates high radiation, and a downscale unit that can indicate instrument trouble. These trips sound alarms but cause no control action. The trip circuits are set so that loss of power causes an alarm. There are two trip indicating lights in the front face of an indicator and trip unit. The range of the channel is six decades and is selected to cover both normal and possible maximum radiation levels at the monitoring location.

The monitors are located (Table 12.3-7) at various places in the containment, auxiliary/reactor enclosure, radwaste facility, turbine enclosure, control room, and other areas where radiation monitoring is desired, based on the following objectives:

a. To monitor the radioactivity level in areas where personnel may be required to work.
b. To provide a record of the radioactivity as a function of time at key locations throughout the plant.

Ranges and sensitivities are selected for each location based on the anticipated radioactivity level as provided by experimental measurements of levels in similar plants and shielding calculations.

Local alarming and indication is provided at those remote sensor locations where a substantial increase in radiation levels might be immediately important to personnel in the area. This system does not have any bypass or interlock interfaces, however each channel can be bypassed.

7.7.1.10.3.3 ARMS Testability An internal trip test circuit, adjustable over the full range of the trip circuit, is provided. The test signal is fed into the indicator and trip unit input so that a meter reading is provided in addition to a real trip. All trip circuits are of the latching-type and must be manually reset at the front panel.

A facility for calibrating these monitor units is provided. This is a test unit designed for use in the adjustment procedure for the area radiation monitor sensor and converter unit. It provides several gamma radiation levels between approximately 4 mR/hr to 300 MR/hr. The calibration unit source is an approved source.

A cavity in the calibration unit receives the sensor and converter unit. A window through which radiation from the source emanates is located on the back wall of the cylindrical lower half of the cavity. A chart on each unit indicates the radiation levels available from the unit for the various control settings.

7.7.1.10.4 ARMS Environmental Considerations The sensor, converter, and local audio alarm auxiliary unit are designed to operate under the environmental conditions at the monitoring locations. The indicator/trip unit and the power supply are designed to operate under auxiliary equipment room environmental conditions and power source fluctuations to guarantee successful operation of the system. See Section 3.11 for the equipment location environmental conditions.

7.7.1.10.5 ARMS Operational Considerations 7.7.1.10.5.1 ARMS General Information CHAPTER 07 7.7-63 REV. 20, SEPTEMBER 2020

LGS UFSAR An annunciator window is provided for each operational area, such as the turbine building area or the auxiliary/reactor enclosure area, although there are a number of ARMS channels in the area, and any one of them can trip the annunciator.

7.7.1.10.5.2 ARMS Reactor Operator Information An operator is able to identify the channel that causes the annunciation by checking the indicating lights of the indicator/trip units.

7.7.1.11 Gaseous Radwaste System - Instrumentation and Controls 7.7.1.11.1 GRS Identification 7.7.1.11.1.1 GRS General The objective of the GRS is to process and control the release of gaseous radioactive wastes to the site environs so that the total radiation exposure to persons outside the controlled area is as-low-as-practicable and does not exceed applicable regulations.

7.7.1.11.1.2 GRS Classification This system is required for power generation only and is classified as not related to safety.

7.7.1.11.1.3 GRS Reference Design Table 7.1-2 lists reference design information. The subject instrumentation and control system is an operational system and has no safety function. Therefore, there are no safety design differences between this system and those of the reference design facilities. This system is functionally identical to the referenced system.

See drawings M-69 and M-70 for the system P&IDs.

7.7.1.11.1.4 GRS Power Source The 120 V ac instrument bus provides power for the GRS instrumentation.

7.7.1.11.2 GRS Equipment Design 7.7.1.11.2.1 GRS General This system is monitored by flow, temperature, pressure, conductivity, and hydrogen analyzers to ensure correct operation and control. Drawings M-69 and M-70 show the process parameters that are instrumented to alarm in the control room. They also indicate whether the parameters are recorded or just indicated. The reactor operator is in control of the system at all times.

The air ejector holdup pipe discharge radiation monitor continuously monitors radioactive effluents to the offgas charcoal treatment system. This monitor provides an alarm in response to high radiation in the offgas. A radiation monitor downstream of the charcoal offgas treatment system continuously monitors radioactive effluents from the charcoal beds. Thus, the radioactivity of the gas entering and leaving the charcoal offgas treatment system is continuously monitored, so that CHAPTER 07 7.7-64 REV. 20, SEPTEMBER 2020

LGS UFSAR the performance of this system is known to the operator at all times. Provisions are also made for the periodic extraction and analysis of grab samples of the influent and effluent gases for determining their compositions.

7.7.1.11.2.2 GRS Recombiner Instrumentation Recombiner vessel temperatures are monitored by inlet and outlet thermocouples. Inlet process gas is monitored for temperature and annunciated in the main control room if its temperatures are high or low. Outlet gas temperature is measured and high temperature is annunciated in the main control room. In addition, gas outlet hydrogen concentration is recorded and alarmed (high) in the control room (drawing M-69). Recombiner inlet hydrogen can also be measured if required.

7.7.1.11.2.3 GRS Offgas Aftercondenser Condensate High and Low Level The offgas aftercondenser condensate high and low levels are annunciated in the main control room. The level switches also control the aftercondenser drain valve. A control board mounted level controller provides remote/manual control of this valve.

7.7.1.11.2.4 GRS Offgas System Inlet Gas Measurements The gas inlet to the offgas components is monitored for temperature, pressure, and moisture content on local panels and in the control room. These parameters are annunciated to alert the operator that corrective action is required.

7.7.1.11.2.5 GRS Hydrogen and Oxygen Analyzer Measurement System Two oxygen analyzers are used to measure the oxygen content of the offgas process stream at the discharge of the aftercondenser. The system is designed to ensure that there is sufficient oxygen available to combine with the hydrogen and yet ensure that the oxygen level does not get too high so as to impose an increased fire hazard for the offgas charcoal beds. Oxygen concentration %

output from the analyzer is displayed in the control room.

Three parallel independent hydrogen analyzers are used to measure the hydrogen content of the offgas process steam downstream of the offgas aftercondenser. One analyzer can be selected to measure hydrogen level upstream of the recombiner. The hydrogen concentration percentage output from the analyzers is recorded in the control room along with independent alarm annunciation for a high hydrogen concentration. Each hydrogen analyzer continuously withdraws a sample of the process offgas, analyzes the hydrogen content, and returns the sample gas to the main condenser. During normal plant operation, the main condenser vacuum provides the pumping force to withdraw the sample gas from the hydrogen analyzer system. Hydrogen percentage calibration checks are made by closing off the line to the offgas process line and admitting a hydrogen calibration gas or a hydrogen free gas; this test is periodically performed. The specified analyzer element is a thermal conductivity cell type of unit, and is designed to prevent hydrogen ignition. An auxiliary vacuum pump is provided to withdraw sample gases in the absence of sufficient main condenser vacuum.

7.7.1.11.2.6 Moisture Measurement One independent moisture detector is placed upstream of the charcoal beds. The moisture content is indicated on a local panel and in the control room, and a high or above moisture content point alarms and annunciates on a local panel and in the control room to notify the operator of the possibility that corrective action may have to be taken.

7.7.1.11.2.7 GRS Charcoal Vessel Temperature Monitoring CHAPTER 07 7.7-65 REV. 20, SEPTEMBER 2020

LGS UFSAR The temperature of each charcoal vessel is monitored and indicated on a local panel and recorded in the control room. This parameter is also annunciated at both locations.

7.7.1.11.2.8 GRS Differential Pressure Measurements Differential pressure is measured across the filter located downstream of the charcoal beds. High differential pressure is alarmed on a local panel and in the control room, and local differential pressure indication is provided.

7.7.1.11.2.9 GRS Testability Because this is a process on-line monitoring, the cross-correlation of the data provides sufficient confirmation of the system's correct operation.

7.7.1.11.3 GRS Environmental Considerations 7.7.1.11.3.1 GRS General The GRS is not required for safety purposes or required to operate after the DBA. The control system is required to operate in the normal plant environment for power generation purposes only.

Radwaste controls and instrumentation located in the offgas equipment area are subject to the environmental and design conditions listed in Section 3.11.

7.7.1.11.3.2 GRS Local Instrument Panels The local instrument panels are located in the operating area outside of the process stream's shield wall. The environmental conditions are the same as described above (Section 3.11).

7.7.1.11.3.3 GRS Special Considerations The instrument sensing lines are rated for 2500 psig. This is sufficient to resist a hydrogen detonation in the process line, or in the tubing.

7.7.1.11.4 GRS Operational Considerations 7.7.1.11.4.1 GRS General Information No operator action is required on the equipment described unless an alarmed condition occurs.

7.7.1.11.4.2 GRS Reactor Operator Information Operator indicators and alarms are shown in drawing M-69.

7.7.1.11.5 GRS Setpoints A hydrogen level of 2% alarms and annunciates in the control room .

7.7.1.12 Liquid Radwaste System - Instrumentation and Controls 7.7.1.12.1 LRS Identification CHAPTER 07 7.7-66 REV. 20, SEPTEMBER 2020

LGS UFSAR 7.7.1.12.1.1 LRS Function The objective of the LRS is to control the release of liquid and solid radioactive waste material to the environs.

7.7.1.12.1.2 LRS Classification This system is required for power generation only and is classified as not related to safety.

7.7.1.12.1.3 LRS Reference Design Table 7.1-2 lists reference design information.

7.7.1.12.2 LRS Power Sources

a. Normal: The 120 V ac instrument power is used for the liquid radwaste process control system.
b. Alternates: Because this is not an essential or safety system, no alternative power source is required.

7.7.1.12.3 LRS Equipment Design The LRS is designed to process liquid waste water to remove particulates, impurities, and other materials and return the processed water for plant use.

Only those portions of the liquid radwaste system providing information that requires operator attention are described to show operator ability to take corrective action when needed.

7.7.1.12.3.1 LRS Equipment Operation Waste water is collected in various sumps throughout the plant and is pumped into the radwaste collection tanks where it is processed. Excess processed liquids that are discharged from the plant are radiation-monitored, flow-controlled, and recorded.

The instrumentation and control system of the radwaste process is typical of a standard chemical and water treatment process. Tank levels are indicated and recorded in the radwaste control room and high tank levels are annunciated in the radwaste and main control room (drawings M-61, M-62, M-63 and M-64).

7.7.1.12.3.1.1 LRS Drywell Sumps Control The drywell sumps discharge to the liquid radwaste system collection tanks (drawing M-61) by gravity. Flow integrators are provided in each sump discharge line to monitor flow out of the drywell sumps. The liquid discharge line from each drywell sump is provided with two drain valves for containment isolation and for sump level control during normal operation. When either isolation valve is closed, the liquid discharge from the drywell sump is stopped. Section 5.2.5.2.1.3.c gives a description of the operation of the sump discharge valves in conjunction with the drywell sump level monitoring system.

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LGS UFSAR 7.7.1.12.3.1.2 LRS Reactor and Turbine Enclosure Sumps These sumps collect waste water from their respective areas and automatically pump out the sumps on level control. Alarm and annunciation in the radwaste and main control rooms occurs on a high-high and low sump level to allow the operator to take corrective action. Run-time meters with indicating lights are also provided to indicate the elapsed run-time of both sump pumps.

The discharge line from the floor drain sample tank to the cooling tower blowdown line isolates upon high radiation detection and alerts the operator. The discharge flow shutoff valves are operated by a key-lock switch that requires plant supervisory control of any releases.

7.7.1.12.3.1.3 LRS Tank Level and Process Control All tanks containing waste liquids throughout the radwaste liquid processing system are provided with liquid level indicators, recorders, and alarm and annunciation in the radwaste and main control rooms to inform the operator that corrective action is to be taken for high or low liquid levels. The process is controlled by an operator from the radwaste control room panel. The control system is designed for manual startup and automatic stop when a process is completed (i.e., tank liquid contents have been emptied to the next process). Because this is a batch system, the operator has full control and responsibility for the system control process.

7.7.1.12.4 LRS Environmental Considerations The radwaste control systems are not required for safety purposes or required to operate after the DBA. The radwaste control systems are required to operate in the normal plant environment for power generation purposes only. The environmental conditions are listed in Section 3.11.

7.7.1.12.5 LRS Operational Considerations 7.7.1.12.5.1 LRS General Information The operator is in full control of the process system batches.

7.7.1.12.5.2 LRS Reactor Operator Information Indicators and recorders are provided for all liquid tanks to inform the operator of the status of the system. Alarms and annunciation are provided to inform the operator that a tank must be emptied or processed, or that a particular piece of equipment malfunctioned, and corrective action must be taken.

7.7.1.12.5.3 LRS Setpoints All tank levels are set to alarm and annunciate before reaching overflow level. This is sufficient time for the operator to take corrective action in the process control.

7.7.1.13 Solid Radwaste System - Instrumentation and Controls 7.7.1.13.1 SRS Identification CHAPTER 07 7.7-68 REV. 20, SEPTEMBER 2020

LGS UFSAR The SRS packages and solidifies radioactive wastes. The system contains the instruments and controls necessary to operate the principal process system. It is a power generation system and is classified as not related to safety. Table 7.1-2 lists reference design information.

7.7.1.13.2 SRS Power Sources The power for the solids radwaste handling system is from an ac power source.

7.7.1.13.3 SRS Equipment Design The process instrumentation and controls provided are as shown in drawings M-66 and M-67.

The instrumentation and control equipment controls the operation of the system that is discussed in Section 11.4.

7.7.1.13.4 SRS Environmental Considerations Controls and instrumentation for the solids radwaste handling system are designed to operate during those environmental conditions listed in Section 3.11.

7.7.1.13.5 SRS Operational Considerations The SRS provides a means of monitoring parameters of the system and permits operator intervention if required.

7.7.1.14 Fuel Pool Cooling and Cleanup System - Instrumentation and Controls 7.7.1.14.1 FPCC Identification 7.7.1.14.1.1 FPCC General The objective of the FPCC is to remove decay heat from the spent fuel storage pool water to ensure adequate cooling of irradiated stored assemblies. The fuel pool cooling system also purifies the spent fuel storage pool water. It maintains water clarity for fuel handling operations and fills and drains the fuel transfer canal. The process instrumentation and controls are shown in drawing M-53.

7.7.1.14.1.2 FPCC Classification This system is required for power generation only and is classified as not related to safety.

7.7.1.14.1.3 FPCC Reference Design Table 7.1-2 lists reference design information.

7.7.1.14.2 FPCC Power Sources The 120 V ac instrument bus provides power for the FPCC system instrumentation.

7.7.1.14.3 FPCC Equipment Design The following process instrumentation is provided:

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a. Fuel pool water level alarms in the control room
b. Skimmer surge tank water level alarms in the main control room
c. Fuel pool cooling pump suction and discharge low pressure alarms in the main control room
d. Fuel pool cooling heat exchangers inlet temperature recorder and alarms in the main control room
e. Fuel pool cooling heat exchangers outlet temperature recorder and alarms 7.7.1.14.4 FPCC Environmental Considerations Controls and instrumentation for the FPCC system are designed to operate during those environmental conditions listed in Section 3.11.

7.7.1.14.5 FPCC Operational Considerations 7.7.1.14.5.1 FPCC General Information No operator action is required unless an alarm condition occurs.

7.7.1.14.5.2 Reactor Operator Information See Section 7.7.1.14.3.

7.7.1.15 Refueling Interlocks - Instrumentation and Controls 7.7.1.15.1 RI Identification The purpose of the RI is to restrict the movement of the control rods and the operation of refueling equipment to reinforce operational procedures that prevent the reactor from becoming critical during refueling operations.

This equipment is not required to operate during a seismic event. The operability of the equipment can be verified after a seismic event without jeopardizing safety.

7.7.1.15.2 RI Power Sources There is only one source of power for both channels of the logic circuits. However, this power source supplies the CRD system as well. A failure of this power supply prevents any rod motion.

7.7.1.15.3 RI Equipment Design 7.7.1.15.3.1 RI Circuit Description The refueling interlock circuitry senses the condition of the refueling equipment and the control rods. Depending on the sensed condition, interlocks are actuated to prevent the movement of the refueling equipment or withdrawal of control rods (rod block). Redundant circuitry is provided to sense the following conditions:

a. All rods inserted (Section 7.7.1.15.3.2)

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b. Refueling platform positioned near or over the core
c. Refueling platform main hoist, fuel-loaded
d. Deleted
e. Reactor mode switch in REFUEL position 7.7.1.15.3.2 RI Logic and Sequencing The indicated conditions are combined in logic circuits to satisfy all restrictions on refueling equipment operations and control rod movement (drawings C11-1030-F-008, C11-1030-F-009, C11-1030-F-010, C11-1030-F-011, C11-1030-F-012, C11-1030-F-013, and C11-1030-F-014).

A two channel circuit indicates that all rods are in. The rod-in condition for each rod is established by the closure of a magnetically operated reed switch in the rod position indicator probe. The rod-in switch must be closed for each rod before the all-rods-in signal is generated. This is not the same switch that provides rod numerical position information to the plant monitoring system (PMS) and four rod position display. Both channels must register the all-rods-in signal before the refueling interlock circuitry can indicate the all-rods-in condition.

During refueling operations, no more than one control rod is permitted to be withdrawn from core cells loaded with any fuel assemblies. This is enforced by a redundant logic circuit that uses the all-rods-in signal and a rod selection signal to prevent the selection of a second rod for movement with any other rod not fully inserted. Control rod withdrawal is prevented by comparison checking between the A and B portions of the RMCS and subsequent message transmission to the affected control rod. The simultaneous selection of two control rods is prevented by the interconnection arrangement of the select push buttons. With the mode switch in the refuel position, the circuitry prevents the withdrawal of more than one control rod and the movement of the refueling platform with the main hoist fuel-loaded over the core with any control rod withdrawn from core cells with any fuel assemblies.

The operation of refueling equipment is prevented by interrupting the motion commands to the equipment. The refueling platform is provided with two mechanical switches attached to the platform, which are tripped open by a long, stationary ramp mounted adjacent to the platform rail.

The switches open before the platform or any of its hoists are physically located over the reactor vessel to indicate that the platform is approaching its position over the core.

Load cell readout is provided for all hoists. Indicators display given hoist loads directly to the hoist operator. Load sensing is by electronic load cells. Associated interlock functions are performed by electronic switches that compare the setpoint values with the signal generated by the electronic load cells.

The main hoist on the refueling platform is provided with switches that open when the hoist is fuel-loaded. The switches open at a load weight that is lighter than that of a single fuel assembly.

This indicates when fuel is loaded on any hoist.

7.7.1.15.3.3 RI Bypasses and Interlocks The rod block interlocks and refueling platform travel interlocks provide two independent levels of interlock action. The interlocks that restrict operation of the platform hoist and grapple provide a third level of interlock action because they would be required only after a failure of a rod block and CHAPTER 07 7.7-71 REV. 20, SEPTEMBER 2020

LGS UFSAR refueling platform travel interlock. The strict procedural control exercised during refueling operations is a fourth level of backup, even though this is actually the primary means of control.

7.7.1.15.3.4 RI Redundancy and Diversity Although the refueling interlocks are not designed or required to meet the IEEE 279 (1971) criteria for nuclear power plant protection systems, a single component failure does not cause an interlock failure. Refueling interlocks are provided for use during planned refueling operations. Criticality is prevented during fuel assembly insertion, provided that the control rod in the associated control cell is fully inserted. The interlock systems accomplish this by:

a. Preventing operation of fuel-loaded refueling equipment over the core whenever any control rod is withdrawn
b. Preventing control rod withdrawal whenever fuel-loaded refueling equipment is over the core
c. Preventing withdrawal of more than one control rod when the mode switch is in the REFUEL position The refueling interlocks have been designed utilizing redundancy of sensors and circuitry to provide a high level of reliability and assurance that the design bases are met. Each of the individual refueling interlocks discussed above need not meet the single failure criteria because the four essentially independent levels (including procedural control) of protection ensure that the design basis is met. For any of the "situations" listed in Table 7.7-3, a single interlock failure does not cause an accident, result in potential physical damage to fuel, or result in abnormal radiation exposure to personnel during fuel handling operations.

7.7.1.15.3.5 RI Actuated Devices The refueling interlocks from the RMCS to the refueling equipment trip logic in the refueling equipment controls that interrupts motion commands to the equipment. This prevents fuel-loaded hoist operation over the core.

The interlocks from the refueling equipment to the RMCS actuate circuitry that provides a control rod block. The rod block prevents the operator from withdrawing any control rods.

7.7.1.15.3.6 RI Separation The refueling interlocks are not designed to, or required to meet, the IEEE 279 (1971) criteria for nuclear power plant protection systems. However, a single interlock failure does not cause an accident. The interlocks are used in conjunction with administrative controls during planned refueling operations. Criticality is prevented during fuel insertion, provided that the control rod in the associated control cell is fully inserted.

7.7.1.15.3.7 RI Testability Complete functional testing of all refueling interlocks prior to use positively indicates that the interlocks operate in the situations for which they are designed. The interlocks can be subjected to valid operational tests by loading the hoist with a dummy fuel assembly, positioning the refueling platform, and withdrawing the control rods. Where redundancy is provided in the logic circuitry, CHAPTER 07 7.7-72 REV. 20, SEPTEMBER 2020

LGS UFSAR tests can be performed on a periodic basis to ensure that each redundant logic element can independently perform its function.

7.7.1.15.4 RI Environmental Considerations Refueling equipment is subjected to the conditions listed in Section 3.11 during normal operation.

Refueling components are designed to withstand design basis events such as earthquakes, accidents, and anticipated operational occurrences without consequential damage, but they are not required to be functional during or after the event without repair.

7.7.1.15.5 RI Operational Considerations 7.7.1.15.5.1 RI General Information The refueling interlocks are provided for use during planned refueling operations.

7.7.1.15.5.2 RI Reactor Operator Information In the refueling mode, the control room operator has indication whenever all control rods are fully inserted. This indication is control rod position data from the computer as well as control rod in and out status on the full core status display. Furthermore, whenever a control rod withdrawal block situation occurs, the operator receives annunciation and computer logs of the rod block. He can compare these outputs with the status of the variable providing the rod block condition. Both channels of the control rod withdrawal interlocks must agree that permissive conditions exist in order to move control rods; otherwise, a control rod withdrawal block is placed into effect. Failure of one channel may initiate a rod withdrawal block and does not prevent application of a valid control rod withdrawal block from the remaining operable channel.

In terms of refueling platform interlocks, the platform operator has digital-type readout indicators for the platform main hoist X-Y position relative to the reactor core.

The position of the grapple is shown on a digital indicator. Digital load cell indications of hoist loads are given for each hoist by locally mounted indicators. Individual joysticks or pendants are provided for local control of the platform and its hoists. The platform operator can immediately determine whether the platform and hoists are responding to his local instructions and can, in conjunction with the control room operator, verify proper operation of each of the three categories of interlocks listed previously.

7.7.1.15.5.3 RI Setpoints There are no safety setpoints associated with this system because this system does not perform a safety function.

7.7.1.16 Leak Detection System - Instrumentation and Controls The nonsafety-related portion of the leak detection system consists of the following systems:

a. Recirculation pump seal leak detection
b. RHR system leak detection CHAPTER 07 7.7-73 REV. 20, SEPTEMBER 2020

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c. Drywell leak detection
d. Safety/relief valve leak detection
e. Reactor pressure vessel head leak detection
f. Core spray system leak detection 7.7.1.16.1 LDS System Identification This section describes the instrumentation and controls associated with the nonsafety-related portion of the leak detection system. Section 7.6 describes the instrumentation and controls in the safety-related portion. This system is discussed in Section 5.2.5.

The purpose of the leak detection system instrumentation and controls is to detect and annunciate before predetermined limits are exceeded.

7.7.1.16.2 LDS Power Sources Power for the nonsafety-related portion of the leak detection system is supplied from a reliable 120 V ac instrument bus.

7.7.1.16.3 LDS Equipment Design 7.7.1.16.3.1 LDS General The systems or parts of systems that contain water or steam coming from the reactor vessel or supply water to the reactor vessel and that are in direct communication with the reactor vessel are provided with leakage detection systems.

The systems within the primary containment share a common area. Therefore, a common leakage detection system is used. Each of the methods used for leakage detection inside the primary containment is designed with the capability to detect leakage at less than the established leakage rate limits.

Major components within the primary containment are, by nature of their design, sources of leakage (e.g., pump seals, valve stem packing, and equipment warming drains), but have their leakage contained and piped to an equipment drain sump and are thereby identified.

Steam or water leaks from other equipment are collected in the floor drain sumps.

Each of the sumps is protected against overflowing to prevent leaks of an identified source from masking those from unidentified sources.

7.7.1.16.4 Recirculation Pump Seal Leak Detection System 7.7.1.16.4.1 RPS-LDS Identification There are two recirculation pump seal leak detection systems, one for each of the pumps in the recirculation loops. The recirculation pump leak detection system consists of two types of CHAPTER 07 7.7-74 REV. 20, SEPTEMBER 2020

LGS UFSAR monitoring circuits (Figure 7.6-2). The first of these monitors the pressure levels within the seal cavities, presenting the plant operator with a visual display of the sensed pressure in each of the two cavities. The second type of monitoring circuit used by the leak detection system monitors the rate of liquid flow from the seal cavities.

7.7.1.16.4.2 RPS-LDS Pump Seal Cavity Pressure Monitoring 7.7.1.16.4.2.1 Circuit Description The pressure levels within seal cavity number 1 and seal cavity number 2 are measured with identical instruments arranged similarly. The pressure within the seal cavities is measured using a pressure transmitter. The pressure transmitter produces an output signal whose magnitude is proportional to the sensed pressure within its dynamic range. This output signal is then transmitted to a pressure indicator for plant operator readout.

7.7.1.16.4.2.2 Logic and Sequencing No automatic action is initiated by the pump seal cavity pressure monitoring circuit.

7.7.1.16.4.2.3 Bypasses and Interlocks No bypasses and interlocks are provided.

7.7.1.16.4.2.4 Redundancy and Diversity No redundancy is provided in this monitoring circuit. The pump seal cavity pressure monitoring is a diverse method of leak detection to the seal cavity flow rate monitoring.

7.7.1.16.4.3 RPS-LDS Liquid Flow Rate Monitoring 7.7.1.16.4.3.1 Circuit Description All condensate flowing past the recirculation pump seal packings and into the seal cavities is collected and sent by one of two drain systems to the primary containment equipment sump for disposal. The first drain system drains the major portion of the condensate collected within the Number 2 seal cavity. The condensate flow rate through the drain system is monitored by a flow switch. The point at which the flow switch contacts close can be adjusted so that the switch is actuated only above or below certain flow rates. Excessively high or low flow rates through this drain system activate an annunciator in the control room.

The second drain system drains the cavity beyond the Number 2 seal cavity collecting the condensate that has seeped (or leaked) past the outer seal. Flow instrumentation monitors the condensate flow through this drain system. A high rate through this system activates an annunciator in the control room.

7.7.1.16.4.3.2 Logic and Sequencing No automatic action is initiated by the seal cavity flow and the second drain system's monitoring circuit.

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LGS UFSAR 7.7.1.16.4.3.3 Bypasses and Interlocks There are no bypasses or interlocks provided.

7.7.1.16.4.3.4 Redundancy and Diversity Redundancy is not provided. Backup indication of seal leakage is provided, however, by monitoring both seal cavities to verify seal failure.

Seal cavity monitoring diversity is provided by monitoring seal cavity pressure.

7.7.1.16.5 Residual Heat Removal Leak Detection System 7.7.1.16.5.1 RHR-LDS Identification The leak detection system constantly monitors the steam lines of the RHR system for leaks. Leaks are detected by steam flow rate monitoring in the common RHR/HPCI steam line. If the monitored parameter indicates that a leak exists, the RHR-LDS activates an annunciator in the control room and initiates an HPCI isolation signal.

The RHR-LDS consists of three types of monitoring circuits. The first of these monitors equipment area ambient and differential temperature, actuating an annunciator when the temperature rises above a present maximum. The second type of circuit monitors the flow rate (differential pressure) through the common RHR/HPCI steam line, actuating an annunciator when the differential pressure (flow) rises above a present maximum. The third type of circuit monitors the shutdown cooling water flow rate and is also annunciated.

Detection of leakage from the ECCS during the long-term post-LOCA cooldown recovery involves detecting leaks from the RHR system. RHR room ambient and differential temperatures can be monitored from the control room. RHR flow and reactor water level can be also monitored from the control room to detect a large leak in the RHR system. All of these instruments are Class 1E instruments.

7.7.1.16.5.2 RHR-LDS Area Temperature Monitoring Leak Detection 7.7.1.16.5.2.1 Circuit Description The LDS constantly monitors the RHR compartment of the RHR system. Leaks are detected by equipment area ambient and differential temperature, which actuates an annunciator when the temperature rises above a preset limit.

7.7.1.16.5.2.2 Logic and Sequencing The RHR and ambient and differential temperature monitoring circuit activates an annunciator when the observed temperature exceeds a preset limit.

7.7.1.16.5.2.3 Bypasses and Interlocks No bypasses or interlocks are associated with this system.

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LGS UFSAR 7.7.1.16.5.2.4 Redundancy and Diversity One channel for ambient and one channel for differential temperature monitoring detect leaks in each RHR system equipment area.

Diversity is satisfied by providing ambient temperature, differential temperature, and RHR/HPCI steam line flow monitoring.

7.7.1.16.5.3 RHR-LDS Flow Rate Monitoring 7.7.1.16.5.3.1 Circuit Description Flow rate is monitored on the RHR shutdown cooling return line.

7.7.1.16.5.3.2 Logic and Sequencing Minimum flow bypass actuation is the only action initiated by RHR flow monitoring.

7.7.1.16.5.3.3 Bypasses and Interlocks No bypasses or interlocks are associated with the process line pressure monitoring instrumentation for the RHR/HPCI common steam flow line.

7.7.1.16.5.3.4 Redundancy and Diversity Isolation of the RHR system on a low pressure condition occurs by signaling a closure of both the common RHR/HPCI inboard and outboard isolation valves. Two redundant pressure sensing switch contacts are used in two-out-of-two coincidence in each logic division (i.e., A and B) to provide pressure monitoring in the common steam line feeding the HPCI and RHR systems.

Because the RHR system is isolated by independent actuation of either logic channel, a single failure of a system component in either channel does not prevent the required isolation function.

Diversity is accomplished through isolation by the following monitored variables: steam line high flow, steam line pressure, and ambient and differential temperature.

7.7.1.16.6 Drywell Leak Detection System 7.7.1.16.6.1 D-LDS Identification Drywell leak detection is accomplished by detecting the level change in two sumps. The level is monitored to detect a rate change that could indicate an abnormality within the primary containment. This instrumentation detects a small leak and provides early warning to the control room operator, permitting corrective measures to be implemented before the abnormality becomes severe (drawings M-61).

7.7.1.16.6.2 D-LDS Circuit Description Each sump is continuously monitored by a transmitter. The signal, which is proportional to level, is linearized for the cylindrical shape of the sump tank and is converted to a flow rate. The flow rate into each sump tank is indicated and alarmed. Flow from each sump tank to radwaste is recorded and alarmed. The alarms are described in sections 5.2.5.2.1.3 and 9.3.3.5.

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LGS UFSAR 7.7.1.16.6.3 D-LDS Logic and Sequencing Annunciation actuation is the only action initiated by the drywell leak detection monitoring circuit.

7.7.1.16.6.4 D-LDS Bypasses and Interlocks There are no bypasses or interlocks associated with this subsystem.

7.7.1.16.7 Safety/Relief Valve Leak Detection System 7.7.1.16.7.1 SRV-LDS Identification Normally, the SRVs are in the shut-tight condition and are all at about the same temperature.

Steam passage through a valve increases the monitored temperature at the exhaust, causing an abnormal temperature reading on the Data Acquisition System (DAS). Relay contacts on the DAS, configured to actuate at a predetermined setpoint, close to complete an annunciator circuit.

Leakage from a valve is usually characterized by a temperature increase on a single relief valve exhaust line.

7.7.1.16.7.2 SRV-LDS Discharge Line Temperature Monitoring 7.7.1.16.7.2.1 Circuit Description A temperature element is placed in the discharge pipe of each SRV for remote indication of leakage. The output of each temperature element is connected to the BOP-DAS.

7.7.1.16.7.2.2 Logic and Sequencing Annunciator actuation is the only action initiated by the SRV temperature monitoring circuit.

7.7.1.16.7.2.3 Bypasses and Interlocks There are no bypasses or interlocks associated with this system.

7.7.1.16.7.2.4 Redundancy and Diversity No redundancy or diversity is required for this system.

7.7.1.16.8 Reactor Pressure Vessel Head Leak Detection System 7.7.1.16.8.1 RPVH-LDS Identification Pressure between the inner and outer head seal-ring is sensed by a pressure switch. If the inner seal leaks, the pressure sensor senses the increased pressure and actuates an alarm. The plant continues to operate with the outer seal as a backup. If both the inner and outer head seals leak, the leak is detected by an increase in drywell temperature and pressure.

7.7.1.16.8.2 RPVH-LDS Head Seal Integrity Pressure Monitoring 7.7.1.16.8.2.1 RPVH-LDS Circuit Description CHAPTER 07 7.7-78 REV. 20, SEPTEMBER 2020

LGS UFSAR A pressure sensor monitors the pressure between the inner and outer head seals.

7.7.1.16.8.2.2 RPVH-LDS Logic and Sequencing No action is initiated by the reactor vessel head pressure monitoring circuit.

7.7.1.16.8.2.3 RPVH-LDS Bypasses and Interlocks There are no bypasses or interlocks associated with this system.

7.7.1.16.8.2.4 Redundancy and Diversity Redundant pressure-sensing instrumentation for detecting inner seal failure is not provided and the outer seal assembly provides backup if an inner seal leak occurs.

7.7.1.16.9 Core Spray System Leak Detection System 7.7.1.16.9.1 CS-LDS Identification CS-LDS is designed to provide core spray cooling to the reactor core. CS-LDS has a safety-related function and is required to be operable following a LOCA.

7.7.1.16.9.2 CS-LDS Circuit Description CS-LDS leaks are detected by the pressure differential across the reactor nozzle and core shroud actuating an annunciator when the pressure differential rises above a preset limit (drawings M-52).

7.7.1.17 Emergency Response Facility Data System Instrumentation and Controls 7.7.1.17.1 ERFDS Identification The ERFDS is a part of the Plant Monitoring System described in Section 7.7.1.22. ERFDS performs the software functions described in this section utilizing the PMS hardware platform described in Section 7.7.1.22.3.

The ERFDS performs two major functions. First, it is the system that conforms with regulatory requirements for a SPDS. The ERFDS can generate real-time and historical visual displays, print records and plots which can be used to provide plant information that aid plant personnel during abnormal and emergency conditions in determining and controlling the value and trend of the key reactor and primary containment control parameters. In addition to the parameters required by Regulatory Guide 1.97 (Rev 2), as discussed in Section 7.5, the ERFDS will monitor the following safety-related parameters:

a. APRM
b. Scram signals
c. HPCI, initiation, speed, discharge pressure, EGM output, ramp output, flow controller output, steam line P
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e. RHR heat exchanger level and pressure
f. RCIC suction pressure control output The SPDS portion of the LGS ERFDS meet the requirements of NUREG-0737 Supplement 1 as discussed below.

The other function is TRA, which provides a real-time and historical perspective for the operation of the power plant. The TRA functions provide high resolution recording capabilities for various plant parameters and the means for event monitoring, data archival, plotting, trending and analyzing in order to support the determination and analysis of plant transients.

7.7.1.17.2 Safety Parameter Display System 7.7.1.17.2.1 SPDS Human Factors The SPDS portion of the LGS ERFDS meets the NRC requirements for an SPDS, in that, it provides aid to the operator in determining the safety status of the plant during abnormal or emergency conditions. The graphic displays available to the control room operator are based on the current revision of the EPGs and are formatted to give maximum assistance in following the LGS EOPs.

Human factors engineering has also been taken into account during development of the LGS SPDS to maximize the operator's ability to readily determine plant status and to minimize errors by the operator during its use.

The SPDS portion of the LGS ERFDS displays is a subset of the generic GE Emergency Response Information System displays. As such, its displays use patterns and colors for conveying current parameter values, trends, limit indications, and validation status that are based upon those used in the generic GE Emergency Response Information System displays.

The human factors reviews of the generic GE Emergency Response Information System displays consisted of a static review by human factor engineering professionals and a dynamic review which emphasized operator integration under dynamic conditions. This dynamic review consisted of a human factor engineering check using a checklist approach, the administration of 12 unique simulated transients, operator/system performance evaluations during the transients using the Perry Nuclear Power Plant EOPs, and data collection for the measurement of the usefulness of the Emergency Response Information System SPDS-related displays.

In general, the Emergency Response Information System was perceived by the operators as a significant aid in plant control during emergencies and was judged as presenting an exceptional source of synthesized/centralized information with regard to plant performance.

The SPDS will be operational during all modes of reactor operation defined in table 1.2 of the LGS Technical Specifications. However, operability of the SPDS does not constitute a Limiting Condition for Operation because hardwired Class 1E instrumentation is available in the control room for the operator if the SPDS should become inoperable.

7.7.1.17.2.2 SPDS Parameter Validation CHAPTER 07 7.7-80 REV. 20, SEPTEMBER 2020

LGS UFSAR A means is incorporated in the design to assure that the data displayed are valid. The validation method used in the LGS SPDS is identical to the method common to all GE SPDSs. The following is a summary of the generic design.

All SPDS-related control parameters of the LGS ERFDS are validated parameters. The validation process generates a weighted average of control parameter signals consisting of either an average of all consistent signals or an average of all in-range signals if there are less than two consistent signals. The validation process generates a validation status that defines whether the average is validated (signals are consistent), nonvalidated (signals are in-range but not consistent), or bad data (signals cannot be measured). If the average cannot be determined, the validation process parameter is assigned "bad data" and the parameter value is replaced with asterisks. The signal average calculated is used to represent the instrument readings of the process variable unless additional compensation is performed on the signal average (e.g., reactor power), in which case the compensated value is used to represent the adjusted instrument reading. For those parameters that are not directly measured (e.g., RPV temperature and other bulk temperatures),

calculations are performed to derive these variables from measured parameters.

7.7.1.17.2.3 SPDS Display Description The LGS SPDS control room displays present the fundamental information needed by nuclear power plant personnel to respond to an emergency. Using standard computer keyboard and mouse at the graphic HMI display console, the user can manually select displays for viewing on the monitor.

The displays available at each graphic display console consist of:

No. of Displays

a. RPV Control display 1
b. Containment Control display 1
c. Critical plant variables 1
d. Two-dimensional plots 9
e. Trend plots 12
f. Validation status displays 8
g. H2 / O2 Control 1
h. Containment Isolation Status 10
i. Remote Manual Isolation Valve Menu 5
j. System Status 16
k. Plant Parameter Status 1 CHAPTER 07 7.7-81 REV. 20, SEPTEMBER 2020

LGS UFSAR These displays provide real-time data with emphasis on showing the current plant status and recent trend history. RPV Control and Containment Control displays are keyed to the appropriate LGS EOPs. The critical plant variables display shows the LGS EOP entry conditions for Reactor Pressure Vessel (RPV) Control and Primary Containment Control. Trend plot displays contain real-time digital information, but their overall emphasis is to show the most recent trends. Two-dimensional plots present the limits defined in the LGS EOPs which are curves showing the relationship between two or more parameters. Validation status displays supply an evaluation of plant control parameter signals.

7.7.1.17.2.3.1 RPV Control Display This display provides control room operators with the primary plant information required to execute the LGS EOPs developed from the RPV Control Guideline. This display is not intended to provide information to unlicensed personnel or personnel whose emergency response functions are not defined by the EOPs (e.g., engineers, supervisor, or management), although these personnel may use this display for detailed plant status information.

7.7.1.17.2.3.1.1 Event Targets There are four event targets on the RPV Control Display. They give the status of the following events.

a. Group Isolation - Has a demand for isolation occurred and has the required isolation been successfully completed?
b. Safety/Relief Valve - Is any SRV open?
c. Main Steam Isolation Valve - Is an MSIV closure signal present and are the MSIVs open or shut?
d. Scram - Has a scram been initiated and have all control rods been fully inserted?

The event labels and color coding for border and text indicate event status of INACTIVE, SAFE, CAUTION, and ALARM.

7.7.1.17.2.3.1.2 Control Parameter Trend Plots Each control parameter, as defined by the EPG, is presented in a trend plot minidisplay consisting of a time history data plot, bar graph, and digital readout. Control parameters for the RPV Control display are RPV water level, pressure, reactor power, and RPV temperature.

The horizontal scale of the time history data plot for all control parameters is the most recent ten minutes.

The bar graph and digital readout are used to highlight and pinpoint the current value of the control parameter. The color of the bar graph and border around the digital readout reflects the control parameter validation status.

CHAPTER 07 7.7-82 REV. 20, SEPTEMBER 2020

LGS UFSAR A trend line tracks the value of each control parameter, and its color coding is the same as that for the bar graph. Whenever the trend line goes off the vertical scale, it appears either at the top of the plot if above scale or at the bottom of the plot if below scale. The user can manually enter upper and lower values of the desired scale range to establish the vertical plot scales.

7.7.1.17.2.3.1.3 Limit Tags A control parameter may have up to five limit tags associated with it, each corresponding to a process limit identified by the LGS EOPs. Table 7.7-8 lists the limit tags that are associated with each of the trend plots on the RPV control display. The process limits are of two types:

dynamic limits and control parameters and, therefore, may change with time. Static limits are limits that remain constant with time. In addition, each of the two types of process limits fall into two categories: upper limits and lower limits. Upper limits are limits that alert the system user when the limit is approached or exceeded from below. Lower limits are limits that alert the system users when the limit is approached or exceeded from above. The process limits further belong to two classes: alarm limits and permissive limits. An alarm limit informs the operator that an operating limit has been exceeded, whereas a permissive limit lets the operator know when an action is capable of being performed (e.g., 100% bypass value).

The process limit status is indicated by the color of the limit tag border. For permissive limit tags, the border colors indicate INACTIVE, ACTIVE, or BAD DATA/DATA NOT MEASURED.

Permissive limit tags are ACTIVE, when the control parameter equals or exceeds the process limit and INACTIVE otherwise. For alarm limit tags, the border colors reflect SAFE, CAUTION, ALARM, or BAD DATA/DATA NOT MEASURED. Alarm limit tags are in the "alarm" state when the control parameter equals or exceeds the process limit, the "caution" state when the control parameter approaches the process limit, and the "safe" state if not in the "alarm" or "caution" state.

Limit lines are presented with trend lines to track the value of dynamic limits associated with the control parameters. Whenever data for a limit line is bad or not measured, the limit line is not plotted.

7.7.1.17.2.3.2 Containment Control Display This top level display provides control room operators with the primary plant information required to execute the LGS EOP developed from the Containment Control Guideline.

7.7.1.17.2.3.2.1 Event Targets There are four event targets on the Containment Control display. They give the status of the following events:

a. Group isolation (Section 7.7.1.17.2.3.1.1)
b. Safety/Relief Valve (Section 7.7.1.17.2.3.1.1)
c. Scram (Section 7.7.1.17.2.3.1.1)
d. HPCI - Should HPCI flow be reduced or should HPCI be secured?

CHAPTER 07 7.7-83 REV. 20, SEPTEMBER 2020

LGS UFSAR The event labels border and text are color coded to indicate event status of INACTIVE, SAFE, CAUTION, and ALARM.

7.7.1.17.2.3.2.2 Control Parameter Trend Plots Control parameters plotted for the Containment Control display are containment level, drywell pressure, drywell temperatures, wetwell pressure, suppression pool temperature, and suppression pool water level. Except for the drywell temperatures, all containment control parameters are validated parameters. The trend plot description is the same as given in Section 7.7.1.17.2.3.1.2.

7.7.1.17.2.3.2.3 Limit Tags As on the RPV Control display, limit tags are associated with each of the trend plots on the Containment Control display. In addition, limit tags for primary containment hydrogen and oxygen concentrations are provided. The limit tag description is the same as given in Section 7.7.1.17.2.3.1.3. Table 7.7-9 lists the limit tags that are associated with each of the trend plots on the Containment Control display.

7.7.1.17.2.3.3 Critical Plant Variables Display This top level display provides the control room operators, shift supervisor, TSC supervisor, and emergency director with the status of the critical plant variables, which are the variables controlled by the LGS EOPs.

The Critical Plant Variables display is an image of the plant and presents two types of EOP information: control parameters and their limits, and event indications. For each control parameter, the current digital readout is shown with the upper limit above and/or lower limit below.

The label and color coding for each digital readout, limit tag, and event indication are identical to the corresponding digital readout, limit tag, or event indication in the RPV or Containment Control displays.

7.7.1.17.2.3.4 Trend Plot Displays Trend plot displays are available for all control parameters. Each trend plot consists of a time history data plot, a bar graph giving the current reading, and a digital readout. Limit tags and limit lines are also supplied. The bar graph, digital readout, limit tags, trend lines, and limit lines are as specified for the control parameter trend plots and limit tags in the RPV or Containment Control displays.

The horizontal plot scale for all inputs is the most recent 30 minutes, except for Drywell/Suppression Pool Hydrogen Concentration and Drywell/Suppression Pool Oxygen Concentration which are the most recent 60 minutes. The plot displays include:

a. RPV Water Level
b. Suppression Pool Level
c. RPV Pressure
d. Reactor Power CHAPTER 07 7.7-84 REV. 20, SEPTEMBER 2020

LGS UFSAR

e. RPV Temperature
f. Containment Water Level
g. Drywell Pressure
h. Wetwell Pressure
i. Suppression Pool Temperature
j. Drywell Temperature
k. Drywell/Suppression Pool Hydrogen Concentration
l. Drywell/Suppression Pool Oxygen Concentration 7.7.1.17.2.3.5 Two-Dimensional Plot Displays These lower level displays provide operators in the control room and plant engineers in the control room and TSC with plots of the two-dimensional limits defined in the EOPs. These limits are also presented as limit tags on the RPV and Containment Control displays.

There are dynamic and static two-dimensional plot displays. Each two-dimensional display consists of an x-y plot with a forbidden region and a historical track, and digital readings of the current values of the dependent and independent parameters. Color coding for digital readings of control parameters is as specified for the RPV and Containment Control displays. Except for the Heat Capacity Temperature Limit, the forbidden region appears on each plot as cross hatch, and the historical track is the curve that continuously tracks the values of the two parameters. The cursor at one end of the historical track represents the current plant status and is color coded the same as the limit status. The two-dimensional displays include:

a. Safety Relief Valve Tail Pipe Level Limit (dynamic)
b. Deleted
c. Heat Capacity Temperature Limit (dynamic)
d. Primary Containment Pressure Limit A (dynamic)
e. Primary Containment Pressure Limit B (dynamic)
f. Pressure-Suppression Pressure (dynamic)
g. Deleted
h. Reference Leg Saturation Limit (dynamic)
i. Drywell Spray Initiation Pressure Limit (dynamic)
j. Reactor Pressure Vessel Pressurization Limit (dynamic)
k. Minimum Debris Retention Injection Rate (dynamic)

CHAPTER 07 7.7-85 REV. 20, SEPTEMBER 2020

LGS UFSAR 7.7.1.17.2.3.6 Validation Status Displays These lower level displays provide operators in the control room and plant engineers in the control room and TSC with an evaluation of the signals for all the control parameters plus core flow. The displays list each instrument that supplies a control parameter signal; indicate if the instrument raw data value is within its calibrated range; compensate individual signals if appropriate; present the compensated value; indicate if these values are consistent with each other; and average these values for a final value, with weighting factors if appropriate.

The upper right field of the display presents the final averaged value, plus indication if that value is compensated and/or validated. The color coding of the readouts are consistent with the control parameter trend plots in Section 7.7.1.17.2.3.1.2. The validation displays include:

a. RPV Level
b. RPV Pressure
c. Reactor Power
d. RPV Temperature
e. Drywell Pressure
f. Drywell Temperature
g. Pool Level
h. Pool Temperature
i. Reactor Core Flow
j. Containment Level
k. Wetwell Pressure 7.7.1.17.2.4 SPDS Isolation Devices Except for the inputs from PRNM, isolation for Class 1 E SPDS signals connected to the LGS ERFDS is provided by analog and digital isolation amplifiers, provided by Scientech l&C. The isolators have been tested generically for electrical isolation, environmental qualification, and seismic qualification.

The analog isolation amplifier isolates the input, safety related, circuitry from the output, non-safety related, circuitry with three components. An isolation amplifier chip transmits each analog signal across a capacitive barrier in a pulse width modulated format. The power and common for the circuitry on the input side of the isolator is separated from the output by isolated DC-DC converters. The circuit board of the isolator module has an isolation gap between input CHAPTER 07 7.7-86 REV. 20, SEPTEMBER 2020

LGS UFSAR and output circuits. The adequacy of these insulation barriers has been tested by the application of credible live fault voltages to the output circuitry in a fault isolation test.

The digital isolation amplifier isolates the input, safety related, circuitry from the output, non-safety related, circuitry with three components. Each digital signal is optically coupled across the isolation barrier by an IC chip. The power and common for the circuitry on the input side of the isolator is separated from the output by isolated DC-DC converters. The circuit board of the isolator module has an isolation gap between input and output circuits. The adequacy of these insulation barriers has been tested by the application of credible live fault voltages to the output circuitry in a fault isolation test.

7.7.1.17.2.5 Verification and Validation Program Historical Information: GE V&V information is historical information pertaining to the original SPDS creation for Limerick.

The Verification and Validation Program used in the development of the original GE LGS SPDS was identical to the program generically used for all GE-supplied SPDSs.

The methods employed in the Verification and Validation procedures ensure that the LGS SPDS will have the functions and characteristics required and that the functions perform correctly. The review and testing processes are designed to identify problems or weaknesses in the design requirements, the design, and the implementation of the design, and to correct those problems and weaknesses.

The specific V&V plan identifies quality audit points along the LGS SPDS development path.

These quality audit points range from performing specification reviews to code walk-throughs to several levels of software and system testing. Heavy emphasis is placed on achieving independent V&V; that is, employing reviewers and testers who have not been directly involved in the design.

Major V&V milestones consist basically of preparation and review of design specifications, coding and review of coding, development of test plans and procedures, conductance of tests and review of test results. The V&V procedure is not only sequential but also iterative. Results that identify areas requiring correction are used to modify the design or further define requirements in order to resolve concerns.

The review of specifications is accomplished by the reviewers documenting comments on a controlled issued document during the development of these specifications. Each comment must be resolved by the responsible engineer to the satisfaction of the reviewers prior to the issue of any document. The testing phases produce test reports which show any discrepancies between expected and actual test results, these discrepancies must be resolved by the responsible engineer of the design group and the test repeated.

By performing the V&V procedures, a systematic and structured method is implemented to ensure that the correct functions are provided and that the functions provided are correct.

The RTime SPDS mirrors the original SPDS. RTime SPDS was verified using the SPDS Software Requirements Specifications and Software Design Documentation. RTime SPDS was validated CHAPTER 07 7.7-87 REV. 20, SEPTEMBER 2020

LGS UFSAR during Factory Acceptance Testing, Site Acceptance Testing, and Modification Acceptance Testing.

7.7.1.17.3 Transient Recording and Analysis The TRA function is used to record, monitor and display plant parameters during the startup testing and at rated power but does not provide control interaction. The system provides transient recordings of the parameters during startup and monitors them at full power.

7.7.1.18 Containment Instrument Gas System - Instrumentation and Controls 7.7.1.18.1 CIGS Identification This system is designed to provide instrument gas to the pneumatic devices located inside the drywell and suppression chamber. The CIGS has no safety-related function and is not required to be operable following a LOCA. Two parts of this system are safety-related. One is the isolation valves provided on all instrument gas lines that penetrate the containment. These valves are part of the primary containment and reactor vessel isolation control system that is described in Section 7.3.1.1.2. The other part provides nitrogen gas to the ADS valves. This portion is described in Section 7.6.

7.7.1.18.2 CIGS Power Sources The instrument panel power supply is provided by a 120 V ac Class 1E source to 120 V ac / 24 V dc power supply. The power to the instrument panel is shed after receiving a LOCA signal. The power may be manually reapplied as loading on the Class 1E buses permits.

7.7.1.18.3 CIGS Equipment Design Equipment design is described in Section 9.3.1.

7.7.1.18.3.1 CIGS Initiating Circuits A signal from a pressure switch located on the instrument gas receiver automatically starts and stops the instrument gas compressor.

7.7.1.18.3.1.1 CIGS Logic and Sequencing Sequencing is not applicable for this system.

During normal unit operation, one of the two instrument gas compressors is selected as the lead compressor and is automatically started or stopped in response to the instrument gas system demand. The other instrument gas compressor serves as a standby. The standby compressor starts automatically if the lead compressor fails or if the lead compressor's continuous operation cannot meet the instrument gas system demand. The two trains are cross-connected by a common header. A backup to the PCIG system is provided by an intertie to the instrument air system.

7.7.1.18.3.1.2 CIGS Bypasses and Interlocks CHAPTER 07 7.7-88 REV. 20, SEPTEMBER 2020

LGS UFSAR This switch must be in the closed position for the bypass permissive to be satisfied. A separate switch is provided for the valves in each division. Placing the bypass switch in the bypass position initiates an alarm in the control room.

7.7.1.18.3.1.3 CIGS Redundancy and Diversity Instrumentation and controls are provided on a one-to-one basis with the mechanical equipment.

7.7.1.18.3.2 CIGS Actuated Devices The instrument gas compressors are actuated by hand switches in the instrument gas compressor control panel and by pressure switches located in the instrument gas receiver.

7.7.1.18.3.3 CIGS Separation The outboard isolation valves, except for the valve in the A line header, are treated as Division II isolation valves. The inboard isolation valve and A line header outboard isolation valve 1 are treated as Division I isolation valves. This arrangement ensures that both instrument gas supply lines are not isolated on the trip of only one isolation logic. The power to the instrument panels is separated in accordance with their respective header isolation valves.

7.7.1.18.3.4 CIGS Testability The isolation valves can be tested as described in Section 7.3.1. However, operation of the instrument gas system is disrupted during the test.

7.7.1.18.4 CIGS Environmental Considerations The controls for this system are located in the reactor enclosure and the control room. For environmental considerations, see Section 3.11.

7.7.1.18.5 CIGS Operational Considerations See Section 9.3.1.

7.7.1.19 Fire Protection and Suppression System - Instrumentation and Controls 7.7.1.19.1 FPSS Identification 7.7.1.19.1.1 FPSS General The FPSS as related to the PGCC is described below. A detailed description and analysis of the PGCC can be found in GE topical report NEDO-10466-A, "Power Generation Control Complex."

The defense in depth concept of fire protection and suppression is implemented in the PGCC by operator use of hand-held fire extinguishers as a backup during fires involving the PGCC.

The total plant FPSS description is found in Section 9.5.1.

7.7.1.19.1.2 FPSS Classification CHAPTER 07 7.7-89 REV. 20, SEPTEMBER 2020

LGS UFSAR The PGCC fire protection and suppression system is classified as not related to safety.

7.7.1.19.1.3 FPSS Reference Design The total plant fire protection design is discussed in Section 9.5.1.

7.7.1.19.2 FPSS Power Sources The fire detection and alarm systems are provided with a 120 VAC emergency diesel/generator backed power supply, a 125 VDC BOP battery supply, and a 24 VDC battery backed power supply. Some heat detectors are powered by the 125 VDC BOP batteries. The other heat detectors are powered by 24 VDC from BOP converters, backed by 24 VDC batteries. The smoke detectors are powered by the 120 VAC emergency diesel/generator backed supply. The power source used for the automatic heat detection also provides power to the electrically operated valves of the automatic suppression system.

7.7.1.19.3 FPSS Equipment Design 7.7.1.19.3.1 FPSS General The PGCC FPSS consists of products of combustion and rate-compensated/temperature detectors with a Halon 1301 fire suppressant system. In-plant fire water service and distribution systems are described in Section 9.5.1.

7.7.1.19.3.2 FPSS Equipment Arrangement PGCC design provides a defense-in-depth approach to fire protection. Each floor section contains four smoke detectors and eight thermal detectors. Fire-stops of refractory material covered by silicone rubber are installed in the cable ducts. A Halon 1301 extinguishing agent is introduced into the floor section cable ducts via a header manifold and nozzle distribution system. This provides at least a 6% concentration of Halon within 10 seconds of activation, and with additional discharge to reach a 20% concentration within 45-55 seconds, there is sufficient Halon to maintain a 20%

concentration for at least 20 minutes. In addition, the floor plate design allows for quick removal so that the control room operators may use hand-held fire extinguishers when required.

There are four smoke detectors in each termination cabinet. The various detectors are wired to contact points in the termination cabinets.

The gas bottles, gas control, suppression initiation and alarm annunciation equipment as well as the total plant alarm system are covered in Section 9.5.1.

7.7.1.19.3.3 FPSS Testability Each hazard area was tested with Halon 1211 or carbon dioxide to check the pneumatic remote control, the tightness of the system and also verified that all pressure-operated controls functioned as specified and that all discharge nozzles were not obstructed.

A discharge concentration test of the Halon suppression system was performed and determined that the design concentration of Halon was obtained within the design discharge time in CHAPTER 07 7.7-90 REV. 20, SEPTEMBER 2020

LGS UFSAR accordance with previously approved calculations. Each system is provided with a thermal detector for test purposes.

7.7.1.19.4 FPSS Environmental Considerations The PGCC fire protection system will operate satisfactorily through a temperature range from -65°F to 104°F and through a humidity range from 50% to 100% relative humidity.

7.7.1.19.5 FPSS Operational Considerations 7.7.1.19.5.1 FPSS General The PGCC detectors provide annunciation at the trouble panel located near the control room entrance. The trouble panel provides an alarm in the control room, and indicates a trouble condition in the PGCC area.

Activation of the Halon suppression system is automatic or manual by the panel operator. Shift supervision has key-lock access to the Halon Release Control Panel to deactivate the Halon suppression system during repair work. Safe operation procedures for use and testing of the total plant fire protection system are covered in Section 9.5.1.

7.7.1.19.5.2 FPSS Reactor Operator Information Operator indications and alarms are shown in drawings M-22.

7.7.1.20 Nonsafety-Related Equipment Area Cooling Ventilation Systems - Instrumentation and Controls 7.7.1.20.1 Reactor Enclosure Ventilation System See Section 9.4.2 for a description of this system.

7.7.1.20.2 Turbine Enclosure Ventilation System See Section 9.4.4 for a description of this system.

7.7.1.20.3 Radwaste Enclosure and Chem. Lab. Expansion Ventilation System See Section 9.4.3 for a description of this system.

7.7.1.20.4 Hot Maintenance Shop Ventilation System See Section 9.4.8 for a description of this system.

7.7.1.20.5 Miscellaneous Enclosures Ventilation Systems See Section 9.4.9 for descriptions of these systems.

7.7.1.21 Rod Worth Minimizer - Instrumentation and Controls CHAPTER 07 7.7-91 REV. 20, SEPTEMBER 2020

LGS UFSAR 7.7.1.21.1 RWM Identification 7.7.1.21.1.1 RWM General The NUMAC RWM is a dedicated microprocessor-based system with an RWM operator display, and a continuous operating self-test feature. It monitors and enforces adherence to established low power level rod insert and withdraw sequences. This function prevents the operator from establishing control rod patterns that are not consistent with the prescribed sequence by initiating appropriate rod insert block, rod withdraw block, and RWM inoperable. The RWM enforces control rod sequences designed to limit (and thereby minimize) individual control rod worths to acceptable levels as determined by the rod-drop accident design basis. The RWM function does not interfere with normal reactor operation and, in the event of failure, does not itself cause rod patterns to be established which would violate the rod-drop accident design basis. The RWM can operate independently of the PMS computer.

The RSCS has been deleted from the LGS plant. Its function is performed by the RWM system.

The RWM acts to prevent withdrawal of an out-of-sequence control rod, to prevent continuous control rod withdrawal errors during reactor startup, and to minimize the core reactivity transient during a rod-drop accident. The consequences of a rod withdrawal error in the startup range are analyzed in Appendix 15B, demonstrating that the licensing basis criterion for fuel failure is still satisfied even when the RWM fails to block rod withdrawal. Thus, the RWM is not safety-related.

The safety action required for the continuous control rod-drop incident (a reactor scram) is provided by the safety-related IRM system of the NMS. If the core flux scram trip setpoint is reached during a flux transient, the IRM will both block further rod withdrawal and initiate a scram. Furthermore, a second safety-related NMS scram trip, supplied the APRM, can terminate the core power transient.

The RWM does not interface with safety-related systems. Refueling interlocks are not considered safety-related.

7.7.1.21.1.2 RWM Classification The RWM is a power generation system and is classified as a system not related to safety.

7.7.1.21.1.3 RWM Reference Design This system is an operational system and has no safety function. Table 7.1-2 lists the reference design information. There are no safety design differences between this system and the reference design since neither system performs a safety function. This system is functionally identical to the referenced system.

7.7.1.21.2 RWM Power Sources The power for the RWM is supplied by a reliable source.

7.7.1.21.3 RWM Equipment Design 7.7.1.21.3.1 RWM Circuit Description The RWM has the following system characteristics:

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LGS UFSAR

b. Selection of operate/bypass/test inop modes
c. Substitute rod position entry capability from the RWM operator's display under key-lock (procedural) control
d. Capability to remove/restore rod from/to group, under key-lock (procedural) control
e. Rod coordinate indication of insert and withdraw error
f. Status indication of insert and withdraw blocks
g. Annunciation of RWM failure to RPIS
h. Identity and position of rods with an insert or withdraw error
i. Single group identification with subgroup capability
j. Text error messages at the RWM operator's display and RWM chassis
k. Rod by rod and group by group sequence control
l. Rapid power reduction 7.7.1.21.3.2 RWM Testability The RWM processor design includes a self-test capability that identifies a failure to a functional module, annunciates the failure, and provides testing capability for verification of repair. The primary function of the self-test feature is to maximize instrument channel availability.

7.7.1.21.4 RWM Environmental Considerations The instrument is designed for normal operation in the range of 5C to 50C and 10% to 90%

relative humidity (noncondensing). It meets EMI requirements for the control room environment.

7.7.1.21.5 RWM Operational Considerations 7.7.1.21.5.1 RWM Operator Information Display information includes multiple displays for different modes of operation and level of detail.

Alternate displays are selected from the display panel through the use of SOFT SWITCHES whose functions are clearly marked on the panel. The use of alternate displays provides additional detail or an alternate set of displays for different modes of operation.

7.7.1.21.5.2 RWM System Interface The RWM is designed to interface with other plant systems as follows:

a. RWM interfaces RPIS and RMCS data to the PMS via a fiber-optic multiplexed data link.

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b. Rod identity and position data, and the identification of the selected rod, are obtained from the RPIS.
c. Interface to the RWM, for downloading, uploading, and printing sequences and for obtaining certain RWM status messages is accomplished via PMS terminal and printer.
d. Connections for ROD SELECTED AND DRIVING, RPIS INOP, LOW POWER SETPOINT, and LOW POWER ALARM POINT are input to the RWM. The RWM transmits these points to the PMS computer.
e. Required output from the RWM include INSERT PERMISSIVE, WITHDRAW PERMISSIVE, RWM OPERABLE, and RWM INOPERABLE signals. These signals interface to Channel 2 of the rod drive control system. The RWM is a dedicated processor with a continuously operating self-test feature. In the event of failure of the RWM, rod blocks and RWM INOP are applied. When the reactor power level is above the LOW POWER SETPOINT, the RWM rod blocks are automatically inhibited by the RWM.
f. Required inputs from the feedwater system are LOW POWER SETPOINT and LOW POWER ALARM POINT.

7.7.1.21.5.3 RWM Function The RWM acts to prevent withdrawal of an out-of-sequence control rod, to prevent continuous control rod withdrawal errors during reactor startup, and to minimize the core reactivity transient during a rod-drop accident. The consequences of a rod withdrawal error in the startup range are analyzed in Appendix 15B, demonstrating that the licensing basis criterion for fuel failure is still satisfied even when the RWM fails to block rod withdrawal. Thus, the RWM is not safety-related.

The safety action required for the continuous control rod-drop incident (a reactor scram) is provided by the safety-related IRM system of the NMS. If the core flux scram trip setpoint is reached during a flux transient, the IRM will both block further rod withdrawal and initiate a scram. Furthermore, a second safety-related NMS scram trip, supplied by the APRM, can terminate the core power transient. The RWM does not interface with safety-related systems. Refueling interlocks are not considered safety-related.

7.7.1.22 Plant Monitoring System - Instrumentation and Controls 7.7.1.22.1 PMS Identification 7.7.1.22.1.1 PMS General The PMS performs six major functions. First, it performs the ERFDS functions as described in section 7.7.1.17. The ERFDS is the system that conforms with regulatory requirements for a SPDS. The PMS can generate real-time and historical visual displays, print records and plots which can be used to provide plant information that aid plant personnel during abnormal and emergency conditions in determining and controlling the value and trend of the key reactor and primary containment control parameters. The second function that the PMS performs is the necessary evaluation of the NSSS portion of nuclear power plant operation. The PMS gathers data, processes data, and provides current outputs for analysis and evaluation purposes. The NSSS programs perform the calculations and data logging required to provide current reactor core performance information. Evaluations can be performed, including thermal power distribution, CHAPTER 07 7.7-94 REV. 20, SEPTEMBER 2020

LGS UFSAR power ratios, energy summaries, enthalpies, data summaries, and calibration and diagnostics for the analysis of the nuclear steam supply. The third function performs BOP performance calculations such as turbine cycle, condenser, electrical, and heat exchanger performance. The fourth function is point log and alarm, which provides alarm display and logging, group point display, log generation and reporting, log and display editor, point data display/change, summary processing and reporting, trend pen recorder, and alarm sequence of events functions. The fifth function is transient recording and analysis, which provides a real-time and historical perspective for the operation of the power plant. The TRA functions provide high resolution recording capabilities for various plant parameters and the means for event monitoring, data archival, plotting, trending, analyzing, and automatic and on-demand logging in order to support the determination and analysis of plant transients. The sixth function is to collect and display meteorological data from the meteorological instrumentation described in Section 2.3 and transmit this meteorological data to the EPDS computer.

7.7.1.22.1.2 PMS Classification The PMS is classified as a power generation system and is not required to perform any safety-related functions.

7.7.1.22.1.3 PMS Reference Design The PMS is an operational system and has no safety function. Table 7.1-2 lists the reference design information. There are no safety design differences between this system and the reference design since neither system performs a safety function. This system is functionally identical to the referenced system.

7.7.1.22.2 PMS Power Sources The power for the computer system is supplied by a reliable source. The DAS is supplied with divisional power as required.

7.7.1.22.3 PMS Equipment Design 7.7.1.22.3.1 PMS Circuit Description The PMS hardware consists of three major elements:

a. The DAS, which consists of Chassis, I/O cards, node processor, and the chassis processor
b. Central processor units and their peripherals
c. Human Machine Interface (HMI) and peripheral devices 7.7.1.22.3.1.1 Data Acquisition System The DAS used for PMS samples signals at high rates to allow high resolution of sequence of events or to perform transient analysis. The deletion of a point from processing shall not remove CHAPTER 07 7.7-95 REV. 20, SEPTEMBER 2020

LGS UFSAR it from its normal scheduled scan by the I/O equipment and the scanned value shall be available for system maintenance and troubleshooting functions. This value shall be available as adjusted counts, raw signal and EU value when displayed.

Features of the DAS include the following:

a. RTP 3000 Series Chassis and I/O cards.
b. Processor cards send plant signals to the PMS (non-Class 1 E) through copper and fiber optic cables. For Class 1 E Panels, an isolation device has been installed between the field signals and the RTP chassis so as to meet the divisional separation and isolation.

7.7.1.22.3.1.2 Central Processing Unit The CPU 64 bit, high speed and high throughput virtual memory computer. The processor's instruction set includes integral floating point, packed decimal arithmetic, and character string instructions. The virtual memory operating system provides for a multiuser, multi-language programming environment. There are two CPUs which are configured in a primary/backup arrangement. This allows CPU failover to be executed in the event of a CPU failure.

7.7.1.22.3.1.3 Human Machine Interface (HMI) and Peripheral Devices In the man/machine interface, a color graphic display system is used. Internal memory capacity of this display system is capable of storing over 200 displays. The color graphics display terminals, along with smaller display terminals, allow access to the PMS functions. The system also includes other peripheral devices such as printer, display, hard copy devices, and high speed line printers.

7.7.1.22.3.2 PMS Testability The PMS has certain self-checking provisions. It performs diagnostic checks to determine the operability of certain portions of the system hardware and performs internal programming checks to verify that input signals and selected program computations are either within specific limits or within reasonable bounds.

7.7.1.22.4 PMS Environmental Considerations All the computer equipment, except for peripherals, is designed for continuous-duty up to 24C and 60.0% relative humidity. The peripherals are designed to operate at up to 30C and 90%

(noncondensing) relative humidity.

7.7.1.22.5 PMS Operational Considerations 7.7.1.22.5.1 PMS NSSS Performance Calculation Programs The NSSS subsystem is designed to enable plant operating personnel to calculate core performance data rapidly to allow the nuclear steam supply to be operated as close to its limits as possible, to aid in safe operation of the plant, and to monitor fuel for uniform exposure and reloading purposes.

CHAPTER 07 7.7-96 REV. 20, SEPTEMBER 2020

LGS UFSAR The functions of the nuclear steam supply performance evaluation subsystem include:

a. Periodic programs for logging a range of reactor performance data.
b. On-demand programs for reactor performance analysis.
c. Programs for a variety of other functions.

7.7.1.22.5.2 PMS Reactor Operators Information (Monitor, Alarm, and Logging Program)

a. Alarms The PMS includes the capability to provide alarm notification and handle their acknowledgement by the user. Alarms are initiated automatically in response to a variable that has reached a specified condition. These alarms are provided for information, user acknowledgment, and historical perspective. The following features are included:
1. Choice of views; i.e., total plant or selected category
2. Use of color for conveying status
3. Alarm prioritization
4. Alarm chronology
5. Ability to segregate acknowledged and unacknowledged alarms
b. Logs The logging function provides the capability to print out or display various logs on demand or control the automatic initiation of the logs which are either printed on a periodic basis or triggered by various plant conditions. The logging functions include special logs, periodic logs, post-trip/scram logs, turbine/generator logs, and SOE logs.

The system has been designed to avoid the loss of data if a log printer fails.

c. Sequence of Events The SOE log is used as follows:
1. To aid in establishing the cause of a reactor scram and determine whether the reactor can be returned to normal operation
2. To provide records of RPS sensor trips
3. To verify proper operation of the ECCS CHAPTER 07 7.7-97 REV. 20, SEPTEMBER 2020

LGS UFSAR All SOE point changes will automatically be logged when a change of state occurs for any SOE input. The SOE function is capable of monitoring 500 digital points, and order of occurrences can be resolved to 100 ms.

7.7.1.22.5.3 PMS BOP Performance Calculation Programs These programs perform calculations and log plant performance data not directly related to the nuclear system. The data stored by the BOP program can be printed out on logs. The BOP Periodic Log gives hourly and daily values for temperatures, power outputs, and flows associated with the main generator and turbines, and with the feedwater, recirculation, and RWCU systems.

The BOP Monthly Log contains monthly averages and accumulations for plant gross and net power outputs, load distributions, turbine heat rates, and fuel burnup. Standard BOP performance calculations include the flow calculations, electrical calculations, turbine cycle performance calculations, condenser calculations, feedwater heater and moisture separator performance calculations, and unit performance calculations.

7.7.1.22.5.4 Collection and Display of Meteorological Data The PMS automatically collects current meteorological data from the meteorological instrumentation described in Section 2.3 through a dedicated data link. Data collection and storage is per the requirements of Regulatory Guide 1.23, rev. 1. This data is available for display during normal operating conditions through PMS displays. These displays may also be available under post accident conditions.

The collected meteorological data is transmitted to the EPDS computer for display and use by the dose assessment model to calculate atmospheric dispersion values and dose from an accidental release of gaseous effluent.

The collected meteorological data is transmitted to the meteorological consultant for performance of activities described in Section 2.3.

7.7.2 ANALYSIS The purpose of this subsection is to:

a. Demonstrate by direct or referenced analysis that the systems described are not required for any plant safety function
b. Demonstrate by direct or referenced analysis that the safety-related systems described elsewhere are capable of coping with all failure modes of the subject control systems In response to Item (a), after considering and evaluating the design basis (Section 7.1.1),

descriptions, and evaluations presented herein relative to the system, it can be concluded that these systems do not perform any safety function.

For consideration of Item (b), it is necessary to refer to the safety evaluations in Chapter 15. These sections demonstrate that the subject systems do not provide any design basis safety function and that any required safety functions are provided by other qualified systems.

CHAPTER 07 7.7-98 REV. 20, SEPTEMBER 2020

LGS UFSAR 7.7.2.1 Reactor Vessel Instrumentation 7.7.2.1.1 RVI General Functional Requirements Conformance The RPV instrumentation is designed to provide redundant or augmented information from the safety-related systems. The operator uses this information to start up, operate at power, shut down, and service the reactor system in an efficient manner. None of this instrumentation is required to initiate or control any safety-related system.

7.7.2.1.2 RVI Specific Regulatory Requirements Conformance Following are the specific regulatory requirements as well as those in 10CFR50, Appendix A, General Design Criteria requirements imposed on this sector vessel instrumentation.

7.7.2.1.2.1 RVI Conformance to Regulatory Guides 7.7.2.1.2.1.1 RVI - Regulatory Guide 1.11 (1971) - Instrument Lines Penetrating Primary Reactor Containment (Safety Guide 11)

Conformance to Regulatory Guide 1.11 is addressed in Section 6.2.4.

7.7.2.1.2.1.2 RVI - Regulatory Guide 1.75 (1978) - Physical Independence of Electric Systems Conformance to Regulatory Guide 1.75 is discussed in Section 7.1.2.5.19.

7.7.2.1.2.2 Conformance to 10CFR50, Appendix A, General Design Criteria 7.7.2.1.2.2.1 RVI - GDC 13 - Instrumentation and Controls The RPV information provides the operator with information on the reactor vessel operating variables during normal plant operation and anticipated operational occurrences so that the need to use the safety systems, although ready and able to respond, is minimized. Portions of this instrumentation are used in the feedwater system to maintain the reactor vessel operating variables within prescribed operating ranges.

7.7.2.1.2.2.2 RVI - GDC 24 - Separation of Protection and Control Systems This instrumentation is not part of, or related to, any safety system. The circuitry of the safety systems is completely independent of this instrumentation so that failures of this instrumentation do not prevent the safety systems from initiating any action when required.

7.7.2.1.2.3 RVI Conformance to Industry Codes and Standards 7.7.2.1.2.3.1 RVI - IEEE 279 (1971), Paragraph 4.7 - Control and Protection System Interaction This instrumentation is separate from and independent of the safety systems circuitry. There is no direct circuit-to-circuit or functional interactions between this instrumentation and the safety systems. No failures in this instrumentation can prevent the safety systems from meeting their minimum performance requirements.

CHAPTER 07 7.7-99 REV. 20, SEPTEMBER 2020

LGS UFSAR 7.7.2.2 Reactor Manual Control System - Instrumentation and Controls 7.7.2.2.1 RMCS General Functional Requirements Conformance The circuitry described for the RMCS is completely independent of the circuitry controlling the scram valves. This separation of the scram and normal rod control functions prevents failures in the reactor manual control circuitry from affecting the scram circuitry. The scram circuitry is discussed in Section 7.2. Because each control rod is controlled as an individual unit, a failure that energizes any of the insert or withdraw solenoid valves affects only one control rod. The effectiveness of a reactor scram is not impaired if any one control rod malfunctions. It can be concluded that no single failure in the RMCS can prevent a reactor scram, and that repair, adjustment, or maintenance of RMCS components does not affect the scram circuitry.

Chapter 15 examines the various failure mode considerations for this system. The expected and abnormal transients and accident events analyzed envelop the NSOA associated with this system's components. These include the following:

a. Control rod withdrawal errors
b. Control rod-drop accident The following are cited:
a. The RMCS is not required for plant safety functions. This system has no function associated with any DBA.
b. This system is not used for plant shutdown resulting from an accident or nonstandard operational conditions.
c. The function of the RMCS is to control core reactivity and thus power level.

Interlocks from many different sources are incorporated to prevent the spurious operation of control rod drives or undesirable rod patterns throughout all ranges of operation.

d. This system contains no components, circuits, or instruments required for reactor trip. There are no operator manual controls that can prevent scram.
e. The requirements for the portions of the RMCS that interface with any safety system function include tolerance to single failures and component quality.
f. The rod withdrawal block trip function prevents an operator from carrying out actions that, if unchecked, could result in a RPS action (scram). A fixed margin separates the rod withdrawal block setpoints and the scram setpoints in IRM and APRM. There are no safety considerations.
g. No specific regulatory requirements apply to the rod block trip function. The circuits are designed to be normally energized (trip on loss of power) and are single active component failure tolerant. The equipment is designed to prevent the rod block trip circuitry from affecting the RPS trips in the IRM and APRM channels through the use of separate trip circuits and relays. IEEE standards do not apply because rod CHAPTER 07 7.7-100 REV. 20, SEPTEMBER 2020

LGS UFSAR block trips are not required for any postulated design basis accident or for safe shutdown.

7.7.2.2.2 RMCS Specific Regulatory Requirements Conformance The following statements represent the RMCS level of compliance with these applicable requirements.

7.7.2.2.2.1 RMCS Conformance to 10CFR50, Appendix A, General Design Criteria 7.7.2.2.2.1.1 RMCS - GDC 24 - Separation of Protection and Control Systems No part of the RMCS is required for scram. The rod block functions provided by the NMS are the only instances where the RMCS uses any instruments or devices related to RPS functions. The rod block signals received from the NMS prevent improper rod motion before limits causing reactor scram are reached. Common APRM and IRM, detectors are used, but the signal is electrically isolated before its use in the RMCS (See Sections 7.6.1.4 and 7.6.2.4 for a description of this interface). This isolation is achieved through two separate relay trip units that prevent any feedback from the RMCS to the RPS. Single failure of a control component therefore does not degrade the RPS.

7.7.2.2.2.1.2 RMCS - GDC 26 - Reactivity Control System Redundancy and Capability The RMCS is one of the two independent reactivity control systems as required by this GDC.

7.7.2.2.3 RMCS Power Generation Design Basis The RMCS conforms to the provisions of the power generation design basis as discussed in the following.

a. The RMCS design limits the potential for inadvertent rod withdrawal, which could lead to RPS protective action by requiring deliberate operator action to effect rod withdrawal. The control buttons for rod withdrawal are spring-returned to the off position; two buttons must be held depressed for continuous rod withdrawal. Rod selection circuitry design prevents the withdrawal of more than one rod at a time.
b. Rod block interlocks in the RMCS design inhibit erroneous control rod motion to prevent circumstances that require an RPS protective action. Pertinent interlocks include the SDV high water level trip, APRM Simulated Thermal Power upscale trip, and IRM upscale trip (Section 7.7.1.2.3.2.3).
c. Rod block interlocks in the RMCS design inhibit erroneous control rod motion that could result in local fuel damage or undesirable core reactivity conditions. Pertinent interlocks include mode switch in SHUTDOWN, the RWM function, RBM (NMS) upscale and APRM recirculation flow upscale (Section 7.7.1.2.3.2.3).
d. Rod block interlocks inhibit rod movement whenever instrumentation is incapable of monitoring the core response or when protective system operating bypasses are in effect. Pertinent interlocks include APRM and RBM downscale or inoperative; APRM less than minimum number of LPRMs; IRM and SRM not fully inserted, CHAPTER 07 7.7-101 REV. 20, SEPTEMBER 2020

LGS UFSAR downscale, or inoperative; rod position indicator malfunction; and SDV high level scram bypassed.

e. The RMCS design provides the operator with the information necessary to achieve prescribed control rod patterns and to provide information pertinent to the position and motion of the control rods. Each rod is instrumented to show its position if fully inserted or if fully withdrawn. The rod pattern controller provides information to the operator concerning rod inhibits, modes, and position.

7.7.2.3 Recirculation Flow Control System - Instrumentation and Controls 7.7.2.3.1 RFCS General Functional Requirements Conformance The ASD has no mechanical inertia and will not supply any coast-down power to the pump motor after loss of ac power to the ASD or any other condition which trips the ASD. However, upon loss of the speed demand signal, the ASD will enforce a speed hold and continue to supply power to the pump motor to maintain the speed prior to the loss of control signal.

The analyses in Chapter 15 examine various failure modes and show that no malfunction in the RFCS can cause enough of a transient to damage the fuel barrier or exceed the nuclear system pressure limits as required by the safety design basis. Therefore, the system is classified as not required for safety.

7.7.2.3.2 RFCS Specific Regulatory Requirements Conformance The RFCS is designed in compliance with the following GDC of Appendix A to 10CFR50; IEEE 279, Paragraph 4.7; and regulatory guides.

7.7.2.3.2.1 RFCS Conformance to 10CFR50, Appendix A, General Design Criteria 7.7.2.3.2.1.1 RFCS - GDC 12 - Suppression of Reactor Power Oscillations See Section 7.1.2.6.7 for compliance discussion.

7.7.2.3.2.1.2 RFCS - GDC 13 - Instrumentation and Control The RFCS is capable of monitoring and controlling its variables over their anticipated range for normal operation and for anticipated operational occurrences.

7.7.2.3.3 RFCS Power Generation Design Basis The reactor RFCS conforms to the power generation design basis as follows:

The RFCS is designed to control the reactor power level, over a limited range, by controlling the flow rate of the reactor recirculating water. The control involves varying the speed of the recirculation pumps by supplying a demand signal to the ASD controller..

7.7.2.4 Feedwater Control System - Instrumentation and Controls 7.7.2.4.1 FCS General Functional Requirements Conformance CHAPTER 07 7.7-102 REV. 20, SEPTEMBER 2020

LGS UFSAR The FCS is a power generation system designed to maintain proper vessel water level. Interlocks are provided to lock the flow changing capabilities in the as-is condition if there is a control signal failure. If the vessel level rises too high, the feedwater pumps and plant main turbine are tripped.

This is an equipment protective action that results in reactor shutdown by the RPS system as outlined in Section 7.2. Lowering the vessel level also causes the RPS to shut down the reactor.

The analyses in Chapter 15 examine the various failure mode considerations for this system relative to plant safety and operational effects. The analyses show that malfunctions of the FCS cannot cause enough of a transient to damage the fuel barrier or exceed the nuclear system pressure limits as required by the safety design basis. Therefore, the system is classified as not required for safety.

7.7.2.4.2 FCS Specific Regulatory Requirements Conformance The FCS is not a safety-related system and is not required for safe shutdown of the plant, nor is it required during or after accident conditions.

No specific regulatory requirements are imposed on the system; however, the following general considerations are offered.

See Section 7.7.1.4 for a discussion of compliance details to the General Design Criteria.

7.7.2.4.3 FCS Power Generation Design Basis The FCS, along with portions of RVI, regulates the feedwater flow over the entire power range of the reactor to maintain adequate water level in the reactor vessel, such that unnecessary initiation of safety-related systems because of low water level are prevented. The FCS instrumentation therefore augments safety-related information so that the operator can startup, operate, and shut down the reactor in an efficient manner.

7.7.2.5 Pressure Regulator and Turbine-Generator System - Instrumentation and Controls 7.7.2.5.1 PRTGS General Functional Requirements Conformance Turbine speed and acceleration are controlled by a pressure regulator that controls steam throttle valve position to maintain constant reactor pressure. The turbine speed governor overrides the pressure regulator during an increase of system frequency or loss of generator load. Excess steam is automatically bypassed directly to the main condenser by the pressure-controlled bypass valves.

The analyses described in Chapter 15 examine the various failure mode considerations for this system relative to plant safety and operational effects. The analysis shows that malfunctions of the PRTGS cannot cause enough of a transient to damage the fuel barrier or exceed the nuclear system pressure limits as required by the safety design basis. Therefore the system is classified as not required for safety.

7.7.2.5.2 PRTGS Specific Regulatory Requirements Conformance CHAPTER 07 7.7-103 REV. 20, SEPTEMBER 2020

LGS UFSAR No specific regulatory requirements are imposed on the PRTGS. However, the following general considerations should be noted.

See Section 7.7.1.5 for a discussion of compliance details to the General Design Criteria.

7.7.2.5.3 PRTGS General The turbine-generator control system is not a safety-related system. The equipment protection systems provided as an integral part of the turbine-generator equipment override the turbine-generator control system. If there is a turbine-generator trip due to an equipment protection action, the control valve fast closure and the stop valve closure inputs to the RPS initiate reactor scram (Sections 7.2.1.1.4.2.d and 7.2.1.1.4.2.e).

Pressure regulator malfunctions that lead to low turbine inlet pressure are detected by pressure sensors provided in the PCRVICS which in turn initiate closure of the main steam line isolation valves (Section 7.3.1.1.2.4.5). Similarly, high turbine inlet pressure causes the RPS to detect high reactor pressure, which initiates reactor scram (Section 7.2.1.1.4.2.b).

Control malfunction (e.g., pressure regulation malfunction - upscale) that results in high flow through the turbine control valves and the bypass valves is detected by main steam flow sensors provided in the PCRVICS, which then initiate closure of the main steam line isolation valves (Section 7.3.1.1.2.4.5) and a subsequent reactor scram (Section 7.2.1.1.4.f).

Interfaces between the subject nonsafety-related system and safety-related systems (RPS and PCRVICS) are designed so that failure of the nonsafety-related components does not negate the necessary safety system functions.

7.7.2.5.4 PRTGS Power Generation Design Basis See Section 7.1.2.1.10.3.

7.7.2.6 Neutron Monitoring System - Instrumentation and Controls 7.7.2.6.1 Source Range Monitor System 7.7.2.6.1.1 SRM General Functional Requirements Conformance The arrangement of the neutron sources and startup chambers in the reactor is shown in Figure 7.6-1. This arrangement produces at least 3 counts/second in the SRM using the sensitivity noted in Section 7.7.1.6.1 and the design source strength at initial reactor startup. With the discriminator setting adjusted to produce the specified sensitivity, the signal-to-noise count ratio is above the 2:1 design basis for cold startup.

If the multiplication of one section of the core increases to put that section of the reactor on a 20 second period, the nearest SRM chamber shows an increase in count rate. In general, at least one detector indicates the change in multiplication.

Normal startup procedures ensure that withdrawal of control rods is distributed about the core to prevent excessive multiplication in any one section of the core.

CHAPTER 07 7.7-104 REV. 20, SEPTEMBER 2020

LGS UFSAR Hence, each SRM chamber can respond in some degree during the initial rod withdrawal. During startup withdrawal, one of the four control rods adjacent to each SRM chamber and one control rod adjacent to each neutron source is withdrawn before the reactor is critical. This procedure reduces source and detector shadowing by the control blade and ensures increases in the detector signals as the core average neutron multiplication increases. Detector shadowing caused by the control blade reduces the SRMs ability to detect neutrons.

Examination of the sensitivity of the SRM detectors and their operating ranges of 106 counts/second indicates that the IRM is on scale before the SRM reaches full-scale (Figure 7.6-8).

7.7.2.6.1.2 SRM Specific Regulatory Requirements Conformance There are no specific regulatory or IEEE requirements for the SRM system.

7.7.2.6.1.3 SRM Power Generation Design Basis The SRM design conforms to the provisions of the power generation design basis as follows:

a. The SRM is designed so that over the operating range the ratio of signal-to-noise does not fall below a factor of two. Before initial power operation and with the control rods fully inserted, a count rate of at least 3 counts/ second will be detected from the installed source.
b. Under the worst possible startup rod withdrawal conditions, the SRM is capable of indicating a measurable increase in output signal from at least one detecting channel before the reactor period is less than 20 seconds.
c. During normal reactor startup the SRM is capable of indicating a substantial increase in output signals with the maximum permissible number of SRM channels out-of-service.
d. The indication on the upper scale of the SRM overlaps the indication on the lower scale of the IRM during reactor startup. The upper limit of the SRM range is 1x109 nv nominal, and the lower limit of the IRM range is 1x108 nv.
e. The SRM signal conditioning equipment processes the input to provide an output corresponding to the reactor period. This output is displayed in the control room and is also provided to remote recorders.
f. The SRM system has trips for SRM upscale, downscale, and instrument inoperable.

With the reactor mode switch in STARTUP or REFUEL, any of these trips will initiate a rod block.

g. The SRM is designed to function in the maximum environmental conditions specified in Section 3.11.
h. Loss of single power bus will de-energize only half of the SRM channels and will not disable the monitoring and alarm functions of the remaining channels.

7.7.2.6.2 Rod Block Monitor System CHAPTER 07 7.7-105 REV. 20, SEPTEMBER 2020

LGS UFSAR 7.7.2.6.2.1 RBM General Functional Requirements Conformance Motion of a control rod causes the LPRMs adjacent to the control rod to respond strongly to the change in power in the region of the rod in motion. Figures 7.7-15 and 7.7-16 illustrate the calculated response of the two RBMs to the full withdrawal of a selected control rod from a region in which the design limits on power and flow exist. Curves are shown for the response when all LPRMs are operable and for cases when certain LPRMs are inoperable.

Because MCPR cannot reach 1.0 until the control rod is withdrawn through greater than half its stroke, the highest rod block setpoint halts rod motion well before local fuel damage can occur.

This is true even with the adjacent and nearest LPRM detector assemblies failed.

7.7.2.6.2.2 RBM Specific Regulatory Requirements Conformance IEEE standards and regulatory guides do not apply to the RBM system because it is not a protection system.

7.7.2.6.2.2.1 RBM Conformance to 10CFR50, Appendix A, General Design Criteria 7.7.2.6.2.2.1.1 RBM - GDC 24 - Separation of Protection and Control Systems The RBM provides an interlocking function in the control rod withdrawal portion of the CRD RMCS.

This design is separated from the protective functions in the plant to ensure their independence.

The RBM is designed to prevent inadvertent control rod withdrawal given an imposed single failure within the RBM. One of the two RBM channels is sufficient to provide an appropriate control rod withdrawal block.

7.7.2.6.2.3 RBM Power Generation Design Basis The RBM design conforms to the provisions of the power generation design basis as follows:

a. The RBM inhibits erroneous control rod motion that could result in local fuel damage. An RBM upscale trip initiates the rod block. A rod block is also initiated by an RBM inoperative trip for any position of the reactor mode switch or RBM downscale with the mode switch in RUN, unless the RBM is correctly bypassed.

One of the two RBM channels can be bypassed without loss of subsystem function because a trip signal from either RBM channel initiates a rod block.

b. RBM signal level is displayed to the operator and indicates the change in relative local power level.

7.7.2.6.3 Traversing Incore Probe System 7.7.2.6.3.1 TIP General Functional Requirements Conformance An adequate number of TIP machines is supplied to ensure that each LPRM assembly can be probed by a TIP and that one LPRM (the central one) can be probed by every TIP to allow intercalibration. Typical TIPs are tested to prove linearity. The system has been field tested in an operating reactor to ensure reproducibility for repetitive measurements. The mechanical CHAPTER 07 7.7-106 REV. 20, SEPTEMBER 2020

LGS UFSAR equipment has undergone life testing under simulated operating conditions to ensure that all specifications can be met. The system design allows semiautomatic operation for LPRM calibration and Plant Monitoring System use. The TIP machines can be operated manually for pointwise flux mapping.

7.7.2.6.3.2 TIP Specific Regulatory Requirements Conformance There are no specific regulatory requirements for the TIP system.

7.7.2.7 Deleted 7.7.2.8 Reactor Water Cleanup System - Instrumentation and Controls 7.7.2.8.1 RWCU General Functional Requirements Conformance The RWCU system is designed and supplied for plant equipment protection and operator information only. None of this instrumentation and control is required to initiate or control any safety-related system.

A high differential pressure across the filter/demineralizer or its discharge strainer automatically closes the unit's outlet valve after sounding an alarm. The holding pump starts whenever there is low flow through a filter/demineralizer. The precoat pump will not start or stop when the level in the precoat tank is low.

7.7.2.8.2 RWCU Specific Regulatory Requirements Conformance 7.7.2.8.2.1 RWCU Conformance to Regulatory Guides 7.7.2.8.2.1.1 RWCU - Regulatory Guide 1.56 (1978) - Maintenance of Water Purity in Boiling Water Reactors The RWCU system provides the recorded conductivity measurements and alarms of influents and effluents of the demineralizers and records of the flow rate through each demineralizer as recommended in Regulatory Guide 1.56.

7.7.2.8.2.2 RWCU Conformance to 10CFR50, Appendix A, General Design Criteria 7.7.2.8.2.2.1 RWCU - GDC 63 - Monitoring Fuel and Waste Storage To meet the requirements of GDC 63, monitors are placed in fuel and waste storage areas to give a continuous display of gamma levels and to alarm if a level exceeds the preselected level.

7.7.2.8.3 RWCU Power Generation Design Basis The RWCU system design conforms to the power generation design basis as discussed in the following:

a. The RWCU system continuously processes reactor water to maintain water purity within specified limits. The system can be operated during normal plant operation, or during startup, shutdown, and refueling. The installed filter/demineralizer CHAPTER 07 7.7-107 REV. 20, SEPTEMBER 2020

LGS UFSAR removes fission products, corrosion products, and other soluble and insoluble impurities.

b. Excess water can be removed from the reactor via the RWCU system to the main condenser, to the condensate storage tank, or to the equipment drain collection tank.

7.7.2.9 Process Radiation Monitoring Systems - Instrumentation and Controls 7.7.2.9.1 South Stack Effluent Radiation Monitoring System 7.7.2.9.1.1 SSE-RMS General Functional Requirements Conformance The SSE-RMS is designed to draw two representative, isokinetic samples from each of the two unitized ducts in the south stack and to monitor these samples for radioactive particulates, iodines, and noble gases. These ducts provide exhaust venting for the reactor enclosure ventilation system. The monitoring systems are completely redundant and independent to ensure that monitoring capability is continuous during periods of maintenance. However, the system is not qualified for seismic Category I or IEEE Class 1E. If there is an accident, the south stack is isolated, and the SSE-RMS is not functional.

7.7.2.9.1.2 SSE-RMS Specific Regulatory Requirements Conformance 7.7.2.9.1.2.1 SSE-RMS Conformance to Regulatory Guides 7.7.2.9.1.2.1.1 SSE-RMS - Regulatory Guide 1.21 (1974) - Measuring, Evaluating, and Reporting Radioactivity in Solid Wastes and Releases of Radioactive Materials in Liquid and Gaseous Effluents from Light-Water-Cooled Nuclear Power Plants The SSE-RMS is designed to comply with Regulatory Guide 1.21. Data generated by this instrument is transmitted to the RMMS (Section 11.5.6) to be incorporated into periodic Regulatory Guide 1.21 reports. Alarms are annunciated in the control room if radioactive concentrations exceed preset limits, but no control action is provided.

7.7.2.9.1.2.2 SSE-RMS Conformance to 10CFR50, Appendix A, General Design Criteria 7.7.2.9.1.2.2.1 SSE-RMS - GDC 13 - Instrumentation and Control Instrumentation is provided to monitor radioactive effluents over their anticipated ranges for normal operation as well as anticipated operational occurrences.

7.7.2.9.1.2.2.2 SSE-RMS - GDC 64 - Monitoring Radioactivity Releases Means are provided for monitoring effluent discharge paths for radioactivity that may be released from normal operations and anticipated operational occurrences.

7.7.2.9.1.2.3 SSE-RMS Conformance to 10CFR50, Appendix I The SSE-RMS affords instrumentation with state-of-the-art sensitivity and quick response times to provide the operator with information that will ensure that radioactive effluents are as low as is CHAPTER 07 7.7-108 REV. 20, SEPTEMBER 2020

LGS UFSAR reasonably practicable. In-place detectors for particulates, iodines, and noble gases provide real-time outputs in the control room to alert personnel to the escape of radioactive material in its early stages.

7.7.2.9.2 Radwaste Equipment Rooms Ventilation Exhaust Radiation Monitoring System -

Instrumentation and Controls 7.7.2.9.2.1 RERV-RMS Power Generation Design Basis This system is designed to provide a display and alarm capability in the control room for concentrations of radioactive material in the ventilation exhaust from the charcoal offgas treatment compartments. This monitoring system indicates leakage of radioactive material from the charcoal beds into the surrounding air.

7.7.2.9.2.2 RERV-RMS General Functional Requirements Conformance This is an off-line process radiation monitoring system that measures the intensity of gamma radioactivity using a detector in a sampling chamber. High/downscale trip alarms are transmitted to annunciators in the control room.

7.7.2.9.2.3 RERV-RMS Specific Regulatory Requirements Conformance Because the only purpose of this system is leak detection, it serves no safety-related function. The system is not required to meet any specific IEEE standards or regulatory guides. This system is used to detect a malfunction of the charcoal offgas treatment system, complying with GDC 60, because the charcoal treatment system controls releases of radioactive materials to the environment.

7.7.2.9.3 Charcoal Treatment System Process Exhaust Radiation Monitoring System -

Instrumentation and Controls 7.7.2.9.3.1 CTSP-RMS Power Generation Design Basis This system is designed to provide a display and alarm capability in the control room for concentrations of radioactive material in the effluents from the charcoal offgas treatment system.

This monitoring system provides an indication of the effectiveness of the offgas treatment system in arresting radioactive discharges to the north stack.

7.7.2.9.3.2 CTSP-RMS General Functional Requirements Conformance This is an off-line process radiation monitoring system that measures the intensity of gamma radioactivity by means of a detector in a sampling chamber. High/downscale trip alarms are transmitted to annunciators in the control room.

7.7.2.9.3.3 CTSP-RMS Specific Regulatory Requirements Conformance There are no specific IEEE standards or regulatory guides that this system is required to meet.

The system conforms to GDC 60, providing the operator with information on the effectiveness of the charcoal effluent treatment system in controlling releases of radioactive materials to the environment.

CHAPTER 07 7.7-109 REV. 20, SEPTEMBER 2020

LGS UFSAR 7.7.2.9.4 Recombiner Rooms and Hydrogen Analyzer Compartments Exhaust Radiation Monitoring System - Instrumentation and Controls 7.7.2.9.4.1 RRHAC-RMS - Power Generation Design Basis This system is designed to provide a display and alarm capability in the control room if radioactive gaseous effluents leak from the recombiners, hydrogen analyzers, or equipment drain sumps. It can also localize the source of such leakage.

7.7.2.9.4.2 RRHAC-RMS General Functional Requirements Conformance This is an off-line process radiation monitoring system that measures concentrations of radioactive gaseous effluents in the ventilation ducting from the recombiner compartments, hydrogen analyzer compartments, and equipment drain sump. High/downscale trip alarms are transmitted to an annunciator in the control room. A hand selector switch actuating a set of solenoid valves is provided for determining the source of the high radiation indication.

7.7.2.9.4.3 RRHAC-RMS Specific Regulatory Requirements Conformance There are no specific IEEE standards or regulatory guides that this system is required to meet.

7.7.2.9.5 Steam Exhauster Discharge and Vacuum Pump Exhaust Radiation Monitoring System -

Instrumentation and Controls 7.7.2.9.5.1 SEDVP-RMS Power Generation Design Basis This monitoring system is designed to detect the presence of short half-life radioactive isotopes leaking from the steam seals of the turbines or excessive radioactivity in the mechanical vacuum pump exhaust.

7.7.2.9.5.2 SEDVP-RMS General Functional Requirements Conformance This is an off-line process radiation monitoring system that measures the intensity of gamma radioactivity by a detector in a sampling chamber.

7.7.2.9.5.3 SEDVP-RMS Specific Regulatory Requirements Conformance There are no specific IEEE standards or regulatory guides that this system is required to meet. It is a power generation system and serves no safety-related function.

7.7.2.9.6 Radwaste Enclosure Ventilation Radiation Monitoring System - Instrumentation and Controls 7.7.2.9.6.1 REV-RMS Power Generation Design Basis This system is designed to provide a display and alarm capability in the control room if unacceptable radioactive concentrations leak into the radwaste enclosure ventilation exhaust.

7.7.2.9.6.2 REV-RMS General Functional Requirements Conformance This is an off-line process radiation monitoring system that measures the concentration of gamma radioactivity in the ventilation exhaust from the radwaste enclosure by means of a detector in a CHAPTER 07 7.7-110 REV. 20, SEPTEMBER 2020

LGS UFSAR shielded sampling chamber. Evidence of leakage in this system is to be used for diagnostic purposes.

7.7.2.9.6.3 REV-RMS Specific Regulatory Requirements Conformance This is a power generation system and serves no safety-related function. There are no specific IEEE standards or regulatory guides that this system is required to meet.

7.7.2.9.7 Air Ejector Offgas Effluent Radiation Monitoring System - Instrumentation and Controls 7.7.2.9.7.1 AEO-RMS General Functional Requirements Conformance The AEO-RMS monitors the offgas effluents downstream of the recombiner and provides indication and alarm under appropriate OUT OF ACCEPTABLE RANGE radiation levels.

The AEO-RMS monitors have characteristics sufficient to provide accurate indication of radioactivity in the air ejector offgas. Sufficient redundancy is provided to allow maintenance on one channel without losing the system indications.

7.7.2.9.7.2 AEO-RMS Specific Regulatory Requirements Conformance No specific regulatory guides or IEEE standards apply to this system.

7.7.2.9.7.2.1 AEO-RMS Conformance to 10CFR50, Appendix A, General Design Criteria 7.7.2.9.7.2.1.1 AEO-RMS - GDC 13 - Instrumentation and Control The system conforms to GDC 13 in that the instruments used cover the anticipated range of radiation under normal operating conditions with enough margin to include postulated accident conditions.

7.7.2.9.7.2.1.2 AEO-RMS - GDC 64 - Monitoring Radioactivity Releases The system conforms to GDC 64 in that it monitors radioactive releases resulting from normal operations including anticipated operational occurrences.

7.7.2.9.8 Primary Containment Leak Detection Radiation Monitoring System - Instrumentation and Controls 7.7.2.9.8.1 PCLD-RMS General Functional Requirements Conformance The PCLD-RMS supplements the leak detection instrumentation for the primary containment. This monitoring system continually samples the containment (drywell) atmosphere for radioactive noble gases. It is designed to respond to increases in the concentration of radioactivity in the containment atmosphere.

This system is not intended for use during or after a LOCA. In this eventuality, the PCLD-RMS will be automatically isolated from the containment. The PCLD-RMS is a power generation system, and is not safety-related.

CHAPTER 07 7.7-111 REV. 20, SEPTEMBER 2020

LGS UFSAR 7.7.2.9.8.2 PCLD-RMS Specific Regulatory Requirements Conformance 7.7.2.9.8.2.1 PCLD-RMS Conformance to Regulatory Guides 7.7.2.9.8.2.1.1 PCLD-RMS - Regulatory Guide 1.45 (1973) - Reactor Coolant Pressure Boundary Leakage Detection Systems The PCLD-RMS is designed to provide one of the three independent means of detecting RCPB leakage. The systems provided meet the intent of Regulatory Guide 1.45 and meet or exceed the recommendations of ANSI/ISA S67.03.

7.7.2.9.8.2.2 PCLD-RMS 10CFR50, Appendix A, General Design Criteria 7.7.2.9.8.2.2.1 PCLD-RMS - GDC 2 - Design Bases for Protection Against Natural Phenomena The PCLD-RMS is not designed to continue its leak detection capability after a SSE. The other RCPB leak detection systems are designed to withstand an SSE (sump level) or an OBE (drywell chiller drains).

7.7.2.9.8.2.2.2 PCLD-RMS - GDC 13 - Instrumentation and Control The PCLD-RMS is designed to monitor for containment internal leakage over normal operation and anticipated occurrences, but not for accidents that result in containment isolation.

7.7.2.9.8.2.2.3 PCLD-RMS - GDC 64 - Monitoring Radioactivity Releases The PCLD-RMS provides a means of monitoring the containment atmosphere for radioactivity released from the RCPB during normal operations and anticipated operational occurrences.

7.7.2.9.9 Hot Maintenance Shop Ventilation Exhaust Radiation Monitoring System -

Instrumentation and Controls 7.7.2.9.9.1 HMS-RMS General Functional Requirements Conformance The HMS-RMS monitors the effluents from the hot maintenance shop exhaust filters in the ventilation duct. All instrumentation is local. This system provides to the RMMS an ongoing record of the number of concentration of particulates and iodines discharged from the hot maintenance shop to the environment. Under normal operating conditions, this radioactive effluent is insignificant. If there is an anomalous occurrence, an alarm annunciator alerts hot maintenance personnel of a malfunction.

7.7.2.9.9.2 HMS-RMS Specific Regulatory Requirements Conformance 7.7.2.9.9.2.1 HMS-RMS Conformance to Regulatory Guides 7.7.2.9.9.2.1.1 HMS-RMS - Regulatory Guide 1.21 (1974) - Measuring, Evaluating, and Reporting Radioactivity in Solid Wastes and Releases of Radioactive Materials in Liquid and Gaseous Effluents from Light-Water-Cooled Nuclear Power Plants CHAPTER 07 7.7-112 REV. 20, SEPTEMBER 2020

LGS UFSAR The HMS-RMS is designed to comply with Regulatory Guide 1.21. Data generated by this instrument are transmitted to the RMMS (Section 11.5.6) to be incorporated into periodic Regulatory Guide 1.21 reports.

7.7.2.9.9.2.2 HMS-RMS Conformance to 10CFR50, Appendix A, General Design Criteria 7.7.2.9.9.2.2.1 HMS-RMS - GDC 13 - Instrumentation and Control Instrumentation is provided to monitor radioactive effluents over their anticipated ranges for normal operation as well as for anticipated operational occurrences.

7.7.2.9.9.2.2.2 HMS-RMS - GDC 64 - Monitoring Radioactivity Releases Means are provided for monitoring an effluent discharge path for radioactivity that may be released from normal operations and anticipated operational occurrences.

7.7.2.9.10 Liquid Radwaste Discharge Radiation Monitoring System - Instrumentation and Controls 7.7.2.9.10.1 LRD-RMS Power Generation Design Basis This system is designed to trip the valves in the liquid radwaste discharge line when the concentration of radioactive effluents in the liquid radwaste discharge is unacceptable. The monitoring system also provides ongoing records of radioactive releases, together with displays in the radwaste control room and alarm annunciation in the main control room.

7.7.2.9.10.2 LRD-RMS General Functional Requirements Conformance This is an off-line process radiation monitoring system that measures the gross concentration of radioactive isotopes in the liquid radwaste discharge. Combined with flow rate data, the radiation data are used to keep an inventory of the total liquid radioactive discharge from the plant. When concentrations of the discharges exceed permissible limits, these discharges are automatically stopped.

7.7.2.9.10.3 LRD-RMS Specific Regulatory Requirements Conformance No specific IEEE standards apply to this system.

7.7.2.9.10.3.1 LRD-RMS Conformance to Regulatory Guides 7.7.2.9.10.3.1.1 LRD-RMS - Regulatory Guide 1.21 (1974) - Measuring, Evaluating, and Reporting Radioactivity in Solid Wastes and Releases of Radioactive Materials in Liquid and Gaseous Effluents from Light-Water-Cooled Nuclear Power Plants The system conforms to Regulatory Guide 1.21 in maintaining records of discharges of radioactive effluents from the plant and in holding the discharges within permissible limits.

7.7.2.9.10.3.2 LRD-RMS Conformance to 10CFR50, Appendix A, General Design Criteria CHAPTER 07 7.7-113 REV. 20, SEPTEMBER 2020

LGS UFSAR 7.7.2.9.10.3.2.1 LRD-RMS - GDC 60 - Control of Releases of Radioactive Materials to the Environment The system controls releases of radioactive materials to the environment.

7.7.2.9.10.3.2.2 LRD-RMS - GDC 64 - Monitoring Radioactive Releases The system provides a means of monitoring an effluent discharge path for radioactivity that may be released from normal operations and anticipated operational occurrences.

7.7.2.9.11 Service Water Radiation Monitoring System - Instrumentation and Controls 7.7.2.9.11.1 SW-RMS Power Generation Design Basis This system is designed to provide a display and alarm capability in the control room if there is leakage of radioactive material into the plant service water. Concentrations of radioactive materials in the service water are detected in the discharge to the cooling tower. A display, an ongoing record, and an alarm annunciation capability are provided in the control room.

7.7.2.9.11.2 SW-RMS General Functional Requirements Conformance This is an off-line process radiation monitoring system that measures the concentration of radioactive materials in the service water discharges. High/downscale trip alarms are transmitted to an annunciator in the control room. This information is used for diagnostic purposes.

7.7.2.9.11.3 SW-RMS Specific Regulatory Requirements Conformance There are no specific IEEE standards or regulatory guides that this system is required to meet. It partially satisfies GDC 64, in that it monitors radioactive releases to the cooling tower resulting from normal operations and anticipated malfunctions.

7.7.2.9.12 Reactor Enclosure Cooling Water Radiation Monitoring System - Instrumentation and Controls 7.7.2.9.12.1 RECW-RMS Power Generation Design Basis This system is designed to provide a display and alarm capability in the control room if there is leakage of radioactive material into the RECW. Such a condition could occur only if a leak occurs in the interface between the reactor enclosure cooling water system and the service water system, and if radioactive contamination is already present in the plant service water system.

Consequently, the probability of a high radiation alarm is remote.

7.7.2.9.12.2 RECW-RMS General Functional Requirements Conformance This is an off-line process radiation monitoring system that measures the concentration of radioactive materials in the cooling water. High/downscale alarm trips are transmitted to an annunciator in the control room. This information is used for diagnostic purposes.

7.7.2.9.12.3 RECW-RMS Specific Regulatory Requirements Conformance There are no specific IEEE standards or regulatory guides that this system is required to meet.

CHAPTER 07 7.7-114 REV. 20, SEPTEMBER 2020

LGS UFSAR 7.7.2.10 Area Radiation Monitoring System - Instrumentation and Controls 7.7.2.10.1 ARMS General Functional Requirements Conformance

a. Sensor/converters monitor the gamma radiation levels at the selected locations and send a dc signal, proportional to the gamma level, to the control room where indicator/trip units display the signal.
b. Wherever it is necessary to have the local indication and alarm, an auxiliary unit and a horn are installed at the desired location. The auxiliary unit receives the signal from the indicator/trip unit both for display and alarm. The local alarm consists of an indicator and a red light. The light and horn are energized if the alarm level is exceeded.

7.7.2.10.2 ARMS Specific Regulatory Requirements Conformance 7.7.2.10.2.1 ARMS Conformance to Regulatory Guides 7.7.2.10.2.1.1 ARMS - Regulatory Guide 1.97 (1980) - Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident Section 7.5.2.5.1.1.2 contains a discussion of the degree of conformance.

7.7.2.10.2.2 ARMS Conformance to 10CFR50, Appendix A, General Design Criteria 7.7.2.10.2.2.1 ARMS - GDC 63 - Monitoring Fuel and Waste Storage To meet the requirements of GDC 63, monitors are placed in fuel and waste storage areas to give a continuous display of gamma levels and to alarm if the levels exceed preselected heights.

7.7.2.11 Gaseous Radwaste System - Instrumentation and Controls 7.7.2.11.1 GRS General Functional Requirements Conformance This is not a safety-related system.

The offgas flow recorder is provided to keep a record of all discharge volumes.

All instrumentation with connections to the offgas process lines is purchased and installed not to exceed a maximum leak rate of 1x10-6 atm-cc/sec to limit release of radioactive gases other than through the controlled process system release point after treatment.

Chapter 15 examines the various failure mode considerations for this system. These include:

a. Failure of the charcoal bed
b. Failure of the delay piping 7.7.2.11.2 GRS Specific Regulatory Requirements Conformance CHAPTER 07 7.7-115 REV. 20, SEPTEMBER 2020

LGS UFSAR 7.7.2.11.2.1 GRS Conformance to Regulatory Guides 7.7.2.11.2.1.1 GRS - Regulatory Guide 1.21 (1974) - Measuring, Evaluating, and Reporting Radioactivity in Solid Wastes and Releases of Radioactive Materials in Liquid and Gaseous Effluents from Light-Water- Cooled Nuclear Power Plants The stack measurements (which include the offgas system effluents) provide the methods of measuring the effluents that are required to be reported.

7.7.2.12 Liquid Radwaste System - Instrumentation and Controls 7.7.2.12.1 LRS General Functional Requirements Conformance The liquid radwaste flow for discharge to the blowdown line is flow controlled and monitored for activity level. The discharge flow shutoff valves are operated by key-lock switches that require plant supervisory control of any releases. The flow is recorded in the radwaste control room.

The packaged wastes are stored in a plant storage area set aside for this purpose. The radioactivity and quantity is the responsibility of plant supervisory personnel.

7.7.2.12.2 LRS Specific Regulatory Requirements Conformance 7.7.2.12.2.1 LRS Conformance to Regulatory Guides 7.7.2.12.2.1.1 LRS - Regulatory Guide 1.21 (1974) - Measuring, Evaluating, and Reporting Radioactivity in Solid Wastes and Releases of Radioactive Materials in Liquid and Gaseous Effluents from Light-Water- Cooled Nuclear Power Plants The radwaste flow measurements provide the method of measuring effluents that are required to be reported.

7.7.2.13 Solid Radwaste System - Instrumentation and Controls 7.7.2.13.1 SRS General Functional Requirements Conformance The solid radwaste handling system is not a safety-related system. Therefore, the instrumentation supplied is for system equipment protection and for operator information only.

7.7.2.13.2 SRS Specific Regulatory Requirements Conformance No specific regulatory requirements are imposed on the subject system instrumentation and controls.

7.7.2.14 Fuel Pool Cooling and Cleanup System - Instrumentation and Controls 7.7.2.14.1 FPCC General Functional Requirements Conformance CHAPTER 07 7.7-116 REV. 20, SEPTEMBER 2020

LGS UFSAR The FPCC system is not a safety-related system. Indicators and alarms are not needed to prevent an accident and are only used as an operator aid for plant equipment protection and for operator information.

The primary function of the FPCC system is to remove decay heat from the spent fuel storage pool water to ensure adequate cooling of irradiated stored assemblies. The FPCC system also purifies the spent fuel storage pool water. It maintains water clarity for fuel handling operations and fills and drains the fuel transfer canal.

The pool water temperature is controlled manually; automatic temperature control is not required.

Once the desired cooler duty has been established, the cooling load changes gradually with time; only when spent fuel is added to or taken from the pool does the cooling load change rapidly. The two fuel pool cooling system pumps are controlled from the control room. All other functions of the FPCC system are accomplished by local manipulation of valves and control of the equipment.

The instrumentation in the FPCC system provides measurements that are used to indicate and alarm as indicated in Section 7.7.1.14.

The FPCC system is monitored for conductivity, temperature, pool level, flow rate, and leakage.

The conductivity measurement provides the operator with the information required to ensure that impurities in the water are maintained at acceptable levels.

Low flow (pump discharge pressure) and temperature monitoring provide the operator with the information required to ensure that the desired temperature is not exceeded and that filtering is maintained. Pool level and leakage monitoring provide the operator with information ensuring the maintenance of adequate shielding and cooling.

The FPCC system is an independent system during normal operation. Evaporative losses in the system can be replaced from the Demin Water System or by the condensate storage system. If the heat load becomes excessive, the shutdown cooling portion of the RHR system can be operated in parallel with the FPCC system to remove the excess heat, when in the shutdown mode.

7.7.2.14.2 FPCC Specific Regulatory Requirements Conformance 7.7.2.14.2.1 FPCC Conformance to 10CFR50, Appendix A, General Design Criteria 7.7.2.14.2.1.1 FPCC - GDC 61 - Fuel Storage and Handling and Radioactivity Control The fuel pool cooling and cleanup system filters and cools the spent fuel pool water to remove residual heat and maintain pool water clarity for refueling operations. The controls and instrumentation designed for this system provide monitoring to ensure proper operation of the fuel pool cooling and cleanup system so that conductivity, temperature, levels, flow, and leakage are maintained at normal levels.

7.7.2.15 Refueling Interlocks - Instrumentation and Controls 7.7.2.15.1 RI General Functional Requirements Conformance CHAPTER 07 7.7-117 REV. 20, SEPTEMBER 2020

LGS UFSAR The refueling interlocks, in combination with core nuclear design and refueling procedures, limit the probability of an inadvertent criticality. The nuclear characteristics of the core ensure that the reactor is subcritical even when the highest worth control rod is fully withdrawn. Refueling procedures are written to avoid situations in which inadvertent criticality is possible. The combination of refueling interlocks for control rods and refueling platform interlocks provides redundant methods of preventing inadvertent criticality even after procedural violations. The interlocks on hoists provide yet another method of avoiding inadvertent criticality.

Table 7.7-3 illustrates the effectiveness of the refueling interlocks. This table considers various operational situations involving rod movement, hoist load conditions, refueling platform movement and position, and mode switch manipulation. The initial conditions in situations 4 and 5 appear to contradict the action of refueling interlocks, because the initial conditions indicate that more than one control rod is withdrawn, yet the mode switch is in REFUEL. Such initial conditions are possible if the rods are withdrawn when the mode switch is in STARTUP, and then the mode switch is turned to REFUEL. In all cases, correct operation of the refueling interlock prevents either the operation of loaded refueling equipment over the core when any control rod is withdrawn, or the withdrawal of any control rod when fuel-loaded refueling equipment is operating over the core. In addition, when the mode switch is in REFUEL, only one rod can be withdrawn; selection of a second rod initiates a rod block.

7.7.2.15.2 RI Specific Regulatory Requirements Conformance No specific regulatory requirements apply to refueling interlocks. The refueling interlocks are designed to be normally energized (fail-safe) and single failure tolerant of equipment failures.

IEEE standards do not apply because the refueling interlocks are not required for any postulated DBA or for safe shutdown. The interlocks are required only for the refueling mode of plant operation. The requirements of 10CFR50, Appendix B, are met in the manner set forth in Chapter 17.

No specific GDC requirements apply to this system.

7.7.2.16 Leak Detection System - Instrumentation and Controls 7.7.2.16.1 LDS General Functional Requirements Conformance Following are the analyses to demonstrate how various general function requirements listed under the leak detection system (Section 7.7.1.16) are satisfied.

7.7.2.16.2 LDS Specific Regulatory Requirements Conformance 7.7.2.16.2.1 LDS Conformance to Regulatory Guides 7.7.2.16.2.1.1 LDS - Regulatory Guide 1.45 (1973) - Reactor Coolant Pressure Boundary Leakage Detection Systems Leakage into the primary reactor containment from identified sources such as valve stem packing, recirculation pump seal, head seal, etc., is separated so that flow rates are monitored separately from unidentified leakages and total flow rates can be established and monitored as described in Section 5.2.5.2.1.3.

CHAPTER 07 7.7-118 REV. 20, SEPTEMBER 2020

LGS UFSAR Similarly, the flow rate for unidentified leakage is monitored to detect a pipe break in the primary containment.

Condensation from the drywell air coolers is monitored to provide a diverse means of detecting RCPB leakage. This monitoring provision is described in Section 5.2.5.2.1.4.

Noble gas radioactivity monitoring is also provided to supplement the flow measurement methods of RCPB leak detection, as described in Section 5.2.5.2.1.5.

The RCPB leak detection systems and equipment provided at LGS meet the intent of Regulatory Guide 1.45 and meet or exceed the recommendations of ANSI/ISA S67.03.

7.7.2.16.2.2 LDS Conformance to 10CFR50, Appendix A, General Design Criteria 7.7.2.16.2.2.1 LDS - GDC 13 - Instrumentation and Control The leak detection sensors and associated electronics are designed to monitor reactor coolant leakage over all expected ranges required for the safety of the plant.

7.7.2.16.2.2.2 LDS - GDC 19 - Control Room Controls and instrumentation are provided in the control room.

7.7.2.16.2.2.3 LDS - GDC 30 - Quality of Reactor Coolant Pressure Boundary The system provides the means for detecting and locating the source of the reactor coolant leakage. This also applies to the sump drywell, recirculating pump, and SRV leak detection equipment.

7.7.2.16.2.2.4 LDS - GDC 34 - Residual Heat Removal Leak detection is provided for the RHR shutdown cooling lines penetrating the drywell.

7.7.2.16.2.2.5 LDS - GDC 54 - Piping Systems Penetrating Containment Leak detection is provided for the RHR shutdown cooling lines penetrating the containment. Sump fill rate monitoring provides leak detection for other pipes penetrating the containment and reactor enclosures.

7.7.2.16.2.3 LDS Conformance to Industry Codes and Standards 7.7.2.16.2.3.1 LDS - IEEE 344 (1971) - Guide for Seismic Qualification of Class 1 Electric Equipment for Nuclear Power Generating Stations The RHR and the drywell leak detection systems comply with this standard. See Section 3.10.

7.7.2.16.2.3.2 LDS - ANSI/ISA S67.03 CHAPTER 07 7.7-119 REV. 20, SEPTEMBER 2020

LGS UFSAR The RCPB leak detection systems and equipment provided at LGS meet or exceed the recommendations of this standard.

7.7.2.17 Emergency Response Facilities Data System - Instrumentation and Controls 7.7.2.17.1 ERFDS General Functional Requirements Conformance The ERFDS is designed to provide the operator with certain categories of information. The system augments information from other systems so that the operator can startup, operate at power, and shutdown in an efficient manner. This system is not required to initiate or control any safety-related system.

7.7.2.17.2 ERFDS Specific Regulatory Requirements Conformance The ERFDS is not classified as a safety system and does not initiate any safety-related functions.

The ERFDS has provisions for appropriate isolation from interfacing equipment classified as Class 1E, and the maintenance of the ERFDS is possible without affecting the function of the interfacing Class 1E systems.

The following summarizes the requirements given in section 4.1 of NUREG-0737, Supplement 1 and generic design implementation characteristics which are fully applicable to LGS:

a. "Displays shall be available..."

Reliability has been designed into the ERFDS by providing

1. redundant and validated signals
2. on-line failure diagnostics
3. processor backup
4. use of quality components "Provide concise displays of critical plant variables..."

This requirement is met by all top level displays discussed in Section 7.7.1.17.

b. "Continuous display of SPDS parameters shall be located conveniently to the control room operator..."

The installed ERFDS configuration ensures that this requirement will be met.

Graphic display consoles are located in the LGS control room. These consoles provide continuous display of SPDS parameters.

c. "The SPDS shall be suitably isolated from...safety systems."

This concern is addressed in Section 7.7.1.17 2.4.

d. "Prompt implementation of SPDS...information."

CHAPTER 07 7.7-120 REV. 20, SEPTEMBER 2020

LGS UFSAR The schedule for implementation of the LGS SPDS is provided in Section 1.13, Item I.D.1. The selection of parameters included in the ERFDS was based on the guidance provided in NUREG-0696.

e. "The SPDS shall be designed to incorporate accepted human factors principles..."

This concern is addressed in Section 7.7.1.17.2 1.

f. "The minimum information to be provided shall be sufficient to provide information to plant operators about:"

"(i) Reactivity Control" Reactor power is one of the EPG control parameters. Its control is addressed by the RPV Control display which includes the reactor power trend plot and the SCRAM event target. Expanded trend information is also given by the full page reactor power trend plot.

"(ii) Reactor core cooling and heat removal from the primary system" Core cooling and heat removal are evidence by EPG control parameters of RPV water level and pressure. The RPV Control Display provides trend and value information for these parameters as well as for RPV temperature when the reactor is shutdown. It also provides event targets (e.g., SRV, MSIV). Expanded trend information is also given by the full page trend plot displays.

"(iii) Reactor coolant system integrity" A breach of the reactor coolant boundary would be evident by the reactor and containment response (the latter for breaches in the primary containment). The initial alert would be evident on the Critical Plant Variables Display and/or the RPV Control and Containment Control Displays. Sufficient information to mitigate the consequences of a RCS breach in accordance with the LGS EOPs is given on the RPV Control and Containment Control Displays through indication of RPV water level, RPV pressure, SRV positions, drywell temperature, and drywell pressure.

"(iv) Radioactivity Control" The LGS ERFDS monitors selected process and area radiation monitors in the plant. However, the computer-based system with primary responsibility for providing the process, area, and stack effluent radiation data required for radioactivity control is the RMMS. An RMMS color graphic CRT with keyboard is located in close proximity (within approximately 6 feet) of the ERFDS CRT in the control room. This provides control room personnel with concise displays of radioactivity control data in close proximity to the SPDS displays provided by the ERFDS. A description of the RMMS is given in Section 11.5.6.

CHAPTER 07 7.7-121 REV. 20, SEPTEMBER 2020

LGS UFSAR

"(v) Containment conditions" Trends and values of containment control parameters and event indications that affect containment conditions (e.g., SRV OPEN, ISOLATION VALVE CLOSURE), are given on the Containment Control Display. Key top level information is given on the Critical Variables Display and expanded trend information on the full page trend plot displays.

7.7.2.18 Containment Instrument Gas System - Instrumentation and Controls 7.7.2.18.1 CIGS General Functional Requirements Conformance Following are analyses that demonstrate how the CIGS satisfies applicable plant design requirements.

7.7.2.18.2 CIGS Specific Regulatory Requirements Conformance 7.7.2.18.2.1 CIGS Conformance to Regulatory Guides 7.7.2.18.2.1.1 CIGS - Regulatory Guide 1.30 (1972) - Quality Assurance Requirements for the Installation, Inspection, and Testing of Instrumentation and Electric Equipment (Safety Guide 30)

See Section 8.1.6.1.

7.7.2.18.2.2 CIGS Conformance to 10CFR50, Appendix A, General Design Criteria 7.7.2.18.2.2.1 CIGS - GDC 13 - Instrumentation and Control Instrumentation is provided to monitor the operation of the CIGS over its anticipated range of operating conditions. Manual operation of the system is possible from the local control panels.

7.7.2.18.2.2.2 CIGS - GDC 56 - Primary Containment Isolation The suction line has two automatic isolation valves: one inside and one outside containment. The CIGS supply lines have an automatic isolation valve outside containment and a check valve inside containment.

7.7.2.18.2.3 CIGS Conformance to Industry Codes and Standards 7.7.2.18.2.3.1 CIGS - IEEE 336 (1971) - Installation, Inspection, and Testing Requirements for Instrumentation and Electric Equipment During the Construction of Nuclear Power Generating Stations See Section 8.1.6.1.

7.7.2.18.2.3.1.2 CIGS - IEEE 384 (1974) - Criteria for Separation of Class 1E Equipment and Circuits See Section 7.1.2.7.9.

CHAPTER 07 7.7-122 REV. 20, SEPTEMBER 2020

LGS UFSAR 7.7.2.19 Fire Protection and Suppression System - Instrumentation and Controls 7.7.2.19.1 FPSS General Functional Requirements Conformance A local annunciator panel located in the auxiliary equipment room will receive a signal from the smoke detectors located in the raised floor. This panel will indicate which PGCC floor section has alarmed. The plant operator will decide, in case of fire, either to use a portable fire extinguisher through removable floor plates or to manually initiate the Halon system from a remote station.

Automatic actuation is initiated by thermal detection.

7.7.2.19.2 FPSS Specific Regulatory Requirements Conformance The PGCC design complies with the applicable requirements of the standards described in the GE topical report NEDO-10466-A, "Power Generation Control Complex."

7.7.2.20 Nonsafety-Related Equipment Area Cooling Ventilation Systems - Instrumentation and Controls 7.7.2.20.1 Power Generation Design Basis Instruments and controls are provided to ensure adequate ventilation for equipment and personnel located in the areas serviced by the nonsafety-related equipment area cooling ventilation systems during normal plant operation.

7.7.2.20.2 General Functional Requirements Conformance 7.7.2.20.2.1 Reactor Enclosure Ventilation System Instrumentation and controls are designed to provide the operator with information necessary to operate the reactor enclosure ventilation system in an efficient manner during normal plant operation. None of this instrumentation is required to initiate or control any safety-related systems because, after a DBA, operation is terminated upon initiation of the REIS.

7.7.2.20.2.2 Turbine Enclosure Ventilation System Instrumentation and controls are designed to provide the operator with the information necessary to operate the turbine enclosure ventilation system efficiently during normal plant operation. The turbine enclosure ventilation system is not safety-related and does not operate after a DBA. See Section 9.4.4.5 for instrumentation requirements.

7.7.2.20.2.3 Radwaste Enclosure and Chem. Lab. Expansion Ventilation System Instrumentation and controls are designed to provide the operator with the information necessary to operate the radwaste enclosure and chem. lab. expansion ventilation system efficiently during normal plant operation. The radwaste enclosure ventilation system is not safety-related and does not operate after a DBA. See Section 9.4.3.5 for instrumentation requirements.

7.7.2.20.2.4 Hot Maintenance Shop Ventilation System CHAPTER 07 7.7-123 REV. 20, SEPTEMBER 2020

LGS UFSAR Instrumentation and controls are designed to provide the operator with the information necessary to operate the hot maintenance shop ventilation system efficiently during normal plant operation.

The hot maintenance shop ventilation system is not safety-related and does not operate after a DBA. See Section 9.4.8.5 for instrumentation requirements.

7.7.2.20.2.5 Miscellaneous Enclosure Ventilation System Instrumentation and controls are designed to provide the operator with the information necessary to operate the miscellaneous enclosure ventilation systems efficiently during normal plant operation. The miscellaneous enclosure ventilation systems are not safety-related and do not operate after a DBA. See Section 9.4.9.5 for instrumentation requirements.

7.7.2.20.3 Specific Regulatory Requirements Conformance There are no specific IEEE standards or regulatory guides with which these systems must comply.

7.7.2.21 Rod Worth Minimizer - Instrumentation and Controls 7.7.2.21.1 RWM General Functional Requirements Conformance 7.7.2.21.1.1 RWM Power Generation Design Basis The RWM monitors and enforces operator adherence to established startup, shutdown, and low power level control rod sequences. The startup or shutdown of the reactor can continue without the RWM when a second licensed operator or a technically qualified member of the unit technical staff is present to check rod movements.

7.7.2.21.2 RWM Specific Regulatory Requirements Conformance The RWM is not classified as a safety-related system and is not required to initiate any safety-related functions.

7.7.2.21.2.1 RWM Conformance to Regulatory Guides 7.7.2.21.2.1.1 RWM - Regulatory Guide 1.29 (1978) - Seismic Design Classification The RWM operator display panel meets the applicable portions of the regulatory guide.

7.7.2.21.2.2 RWM Conformance to Industry Codes and Standards 7.7.2.21.2.2.1 RWM - IEEE 388 (1974) - Standard for Type Test of Class 1E Electric Cables, Field Splices, and Connections for Nuclear Power Generating Stations The RWM operator display panel and interconnecting cable meet the applicable portions of the regulatory guide.

7.7.2.22 Plant Monitoring System - Instrumentation and Controls 7.7.2.22.1 PMS General Functional Requirements Conformance CHAPTER 07 7.7-124 REV. 20, SEPTEMBER 2020

LGS UFSAR The plant monitoring system is designed to provide the operator with certain categories of information. The system augments information from other systems so that the operator can start up, operate at power, and shut down in an efficient manner. This system is not required to initiate or control any safety-related system. The analyses for the RWM function are contained in Section 7.7.1.21.

7.7.2.22.2 PMS Specific Regulatory Requirements Conformance The PMS is not classified as a safety system and does not initiate any safety-related functions.

The PMS has provisions for appropriate isolation from interfacing equipment classified as Class 1E, and the maintenance of the PMS is possible without affecting the function of the interfacing 1E systems.

The PMS data acquisition system employs signal isolation and fiber optics cables designed to prevent electrical interference between the PMS and Class 1E safety systems. Thus the PMS conforms with all electrical isolation requirements.

7.7.2.22.2.1 PMS Conformance to Industry Codes and Standards 7.7.2.22.2.1.1 PMS - IEEE 323 (1974) - IEEE Trial Use Standard: General Guide for Qualifying Class 1 Electric Equipment for Nuclear Power Generating Stations These DAS I/O modules are not Class 1E qualified but the Scientech I/C Isolators conform to this standard. Class 1E qualifications and conformance to the IEEE Standards applies only to those modules that provide isolation between Class 1E and Non-Class 1E circuits. (NUREG-0588 Category I) 7.7.2.22.2.1.2 PMS - IEEE 344 (1975) - Guide for Seismic Qualification of Class 1 Electric Equipment for Nuclear Power Generating Stations These DAS I/O modules are not Class 1E but the Scientech I/C Isolators conform to this standard.

Class 1E qualifications and conformance to the IEEE Standards applies only to those modules that provide isolation between Class 1E and Non-Class 1E circuits. See Section 3.10.

7.7.2.22.2.1.3 PMS - IEEE 383 (1974) - Guide for Type Test of Class 1 Electric Cables, Field Splices, and Connections for Nuclear Power Generating Stations These DAS I/O modules are not Class 1E qualified but the Scientech l/C Isolators conform to this standard. Class 1E qualifications and conformance to the IEEE Standards applies only to those modules that provide isolation between Class 1E and Non-Class 1E circuits.

7.7.2.22.2.1.4 PMS - IEEE 384 (1981) - Criteria for Independence of Class 1E Equipment and Circuits These DAS I/O modules are not Class 1E qualified but the Scientech l/C Isolators conform to this standard. No fault on the isolator output wiring can affect any isolator input signal. Class 1E qualifications and conformance to the IEEE Standards applies only to those modules that provide isolation between Class 1E and Non-Class 1E circuits.

CHAPTER 07 7.7-125 REV. 20, SEPTEMBER 2020

LGS UFSAR Table 7.7-1 CRD HYDRAULIC SYSTEM PROCESS INDICATORS MEASURED VARIABLE INSTRUMENT TYPE Total system flow Flow indicator Drive water pump suction Annunciator pressure Drive water filter Annunciator differential pressure Cooling water header Pressure indicator differential pressure Charging water header Annunciator and indicator pressure Drive water flow rate Flow indicator Cooling water header flow rate Flow indicator CRD temperature Annunciator SDV not drained Annunciator Scram valve pilot air header Annunciator pressure Drive water header Pressure indicator differential pressure Drive water pump suction filter Annunciator differential pressure CHAPTER 07 7.7-126 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 7.7-2 LOCATIONS FOR AREA RADIATION MONITORING SENSORS CHANNEL EQUIPMENT MONITOR RANGE LOCAL NUMBER NUMBER DESCRIPTION OF LOCATION (mR/hr) ALARM 01 RE01-M1-1N001 RCIC pump compartment .01-104 YES 01 RE01-M1-2N001 RCIC pump compartment .01-104 YES 02 RE02-M1-1N001 HPCI pump-room el 177' .01-104 YES 02 RE02-M1-2N001 HPCI pump-room el 177' .01-104 YES 03 RE03-M1-1N001 Reactor enclosure sumps el 177' .01-104 YES 03 RE03-M1-2N001 Reactor enclosure sumps el 177' .01-104 YES 04 RE04-M1-1N001 CRD pumps area el 200' .01-104 YES 04 RE04-M1-2N001 CRD pump area el 200' .01-104 YES 05 RE05-M1-1N001 Turbine auxiliary bay hallway el 200' .01-104 YES 05 RE05-M1-2N001 Turbine auxiliary bay hallway el 200' .01-104 YES 06 RE06-M1-1N001 Isolation valve compartment el 201' .01-104 YES 06 RE06-M1-2N001 Isolation valve compartment el 201' .01-104 YES 07 RE07-M1-1N001 Condensate pump-room el 189' .01-104 YES 07 RE07-M1-2N001 Condensate pump-room el 189' .01-104 YES 08 RE08-M1-1N001 RHR division I room el 201' .01-104 YES 08 RE08-M1-2N001 RHR division I room el 201' .01-104 YES 09 RE09-M1-1N001 RHR division II room el 201' .01-104 YES CHAPTER 07 7.7-127 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 7.7-2 (Cont'd)

CHANNEL EQUIPMENT MONITOR RANGE LOCAL NUMBER NUMBER DESCRIPTION OF LOCATION (mR/hr) ALARM 09 RE09-M1-2N001 RHR division II room el 201' .01-104 YES 10 RE10-M1-1N001 Steam vent area stairwell el 217' .01-104 YES 10 RE10-M1-2N001 Steam vent area stairwell el 217' .01-104 YES 11 RE11-M1-1N001 Reactor enclosure, railroad access airlock el 217' .01-104 YES 11 RE11-M1-2N001 Reactor enclosure, railroad access airlock el 217' .01-104 YES 12 RE12-M1-1N001 Hallway, condensate filter/

demineralizers el 217' .01-104 YES 12 RE12-M1-2N001 Hallway, condensate filter/

demineralizers el 217' .01-104 YES 13 RE13-M1-1N001 Turbine Enclosure, condenser area el 217' 1.0-106 YES 13 RE13-M1-2N001 Turbine enclosure, condenser area el 217' 1.0-106 YES 14 RE14-M1-1N001 Reactor drywell el 253' 1.0-106 YES 14 RE14-M1-2N001 Reactor drywell el 253' 1.0-106 YES 15 RE15-M1-1N001 Reactor enclosure, east el 253' .01-104 YES 15 RE15-M1-2N001 Reactor enclosure, east el 253' .01-104 YES CHAPTER 07 7.7-128 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 7.7-2 (Cont'd)

CHANNEL EQUIPMENT MONITOR RANGE LOCAL NUMBER NUMBER DESCRIPTION OF LOCATION (mR/hr) ALARM 16 RE16-M1-1N001 Reactor Enclosure, west el 253' .01-104 YES 16 RE16-M1-2N001 Reactor enclosure, west el 253' .01-104 YES 17 RE17-M1-1N001 Neutron monitoring system area el 253' .01-104 YES 17 RE17-M1-2N001 Neutron monitoring system area el 253' .01-104 YES 18 RE18-M1-1N001 Neutron monitoring drive mechanism .01-104 YES 18 RE18-M1-2N001 Neutron monitoring drive mechanism .01-104 YES 19 RE19-M1-1N001 Turbine auxiliary bay hallway east el 239' 1.0-106 YES 19 RE19-M1-2N001 Turbine auxiliary bay hallway east el 279' 1.0-106 YES 20 RE20-M1-1N001 Turbine auxiliary bay hallway west el 239' 1.0-106 YES 20 RE20-M1-2N001 Turbine auxiliary bay hallway west el 239' 1.0-106 YES 21 RE21-M1-1N001 RWCU heat exchanger area el 283' .01-104 YES 21 RE21-M1-2N001 RWCU heat exchanger area el 283' .01-104 YES CHAPTER 07 7.7-129 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 7.7-2 (Cont'd)

CHANNEL EQUIPMENT MONITOR RANGE LOCAL NUMBER NUMBER DESCRIPTION OF LOCATION (mR/hr) ALARM 22 RE22-M1-1N001 RWCU pump area el 283' .01-104 YES 22 RE22-M1-2N001 RWCU pump area el 283' .01-104 YES 23 RE23-M1-1N001 SLCS area el 283' .01-104 YES 23 RE23-M1-2N001 SLCS area el 283' .01-104 YES 24 RE24-M1-1N001 RWCU instrument rack el 283' .01-104 YES 24 RE24-M1-2N001 RWCU instrument rack area el 283' .01-104 YES 25 RE25-M1-1N001 Turbine auxiliary bay el 269' .01-104 YES 25 RE25-M1-2N001 Turbine auxiliary bay el 269' .01-104 YES 26 RE26-M1-1N001 Turbine enclosure washdown area el 269' .01-104 YES 26 RE26-M1-2N001 Turbine enclosure washdown area el 269' .01-104 YES 27 RE27-M1-1N001 RWCU filter area el 313' .01-104 YES 27 RE27-M1-2N001 RWCU filter area el 313' .01-104 YES CHAPTER 07 7.7-130 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 7.7-2 (Cont'd)

CHANNEL EQUIPMENT MONITOR RANGE LOCAL NUMBER NUMBER DESCRIPTION OF LOCATION (mR/hr) ALARM 28 RE28-M1-1N001 Turbine enclosure equipment compartment .01-104 YES exhaust filters area el 302' 28 RE28-M1-2N001 Turbine enclosure equipment compartment .01-104 YES exhaust filters area el 302' 29 RE29-M1-1N001 Drywell head lay-down area el 352' .01-104 YES 29 RE29-M1-2N001 Drywell head lay-down area el 352' .01-104 YES 30 RE30-M1-1N001 Steam separator area el 352' .01-104 YES 30 RE30-M1-2N001 Steam separator area el 352' .01-104 YES 31 RE31-M1-1N001 Spent fuel pool el 352' .01-104 YES 31 RE31-M1-2N001 Spent fuel pool el 352' .01-104 YES 32 RE32-M1-1N001 New fuel storage vault el 352' .01-104 YES 32 RE32-M1-2N001 New fuel storage vault el 352' .01-104 YES 33 RE33-M1-1N001 Pool plug lay-down area el 352' .01-104 YES 33 RE33-M1-2N001 Pool plug lay-down area el 352' .01-104 YES 34 RE34-M1-1N001 H2/02 analyzers area el 200' .01-104 YES 34 RE34-M1-2N001 H2/02 analyzers area el 200' .01-104 YES CHAPTER 07 7.7-131 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 7.7-2 (Cont'd)

CHANNEL EQUIPMENT MONITOR RANGE LOCAL NUMBER NUMBER DESCRIPTION OF LOCATION (mR/hr) ALARM 35 RE35-M1-1N001 Gaseous radwaste recombiner hallway, el. 180' .01-104 YES 35 RE35-M1-2N001 Gaseous radwaste recombiner hallway, El. 180' .01-104 YES 36 RE36-M1-0N001 Unit 1 deep bed demin area, .01-104 YES el. 217 41 RE41-M1-0N001 RWCU Sludge discharge mixing pump-room el. 162' .01-104 YES 42 RE42-M1-0N001 Radwaste enclosure hallway el. 162' .01-104 YES 43 RE43-M1-0N001 Concentrate storage discharge pump-room el. 191' .01-104 YES 44 RE44-M1-0N001 Laundry drain processing room El. 191' .01-104 YES 45 RE45-M1-0N001 Floor drain filter holding pump-room el. 191' .01-104 YES 46 RE46-M1-0N001 Fuel pool holding pump-room el. 191' .01-104 YES 47 RE47-M1-0N001 Precoat tank & pump-room el. 191' .01-104 YES 48 RE48-M1-0N001 Remote shutdown control area el. 289' .01-104 YES 49 RE49-M1-0N001 Radwaste cask loading area el. 217' .01-104 YES CHAPTER 07 7.7-132 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 7.7-2 (Cont'd)

CHANNEL EQUIPMENT MONITOR RANGE LOCAL NUMBER NUMBER DESCRIPTION OF LOCATION (mR/hr) ALARM 50 RE50-M1-0N001 Railroad car airlock el 217' .01-104 YES 51 RE51-M1-0N001 Radwaste enclosure hallway el 217' .01-104 YES 52 RE52-M1-0N001 Hot Maintenance Shop el 217' .01-104 YES 53 RE53-M1-0N001 Entrance, turbine enclosure railroad el 217' .01-104 YES 54 RE54-M1-0N001 Radwaste enclosure el 239' .01-104 YES 55 RE55-M1-0N001 Radwaste exhaust fan area el 257' .01-104 YES 56 RE56-M1-0N001 Control room el 269' .01-104 YES 57 RE57-M1-0N001 Turbine area operating floor el 269' .01-104 YES 58 RE58-M1-0N001 Standby-gas treatment filter room el 332' .01-104 YES 59 RE59-M1-0N001 Source storage & calibration room, .01-104 YES el 230'

- RIAH-TA-001(1) (Local) A turbine enclosure crane el 310' .01-104 YES

- RIAH-TA-002(1) (Local) B turbine enclosure crane El 310' .01-104 YES

- RIAH-TA-003(1) (Local) C turbine enclosure crane el 310' .01-104 YES (1)

Local monitor only - requires local power supply (monitor and power supply supplied as part of BOP scope)

CHAPTER 07 7.7-133 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 7.7-2 (Cont'd)

CHANNEL EQUIPMENT MONITOR RANGE LOCAL NUMBER NUMBER DESCRIPTION OF LOCATION (mR/hr) ALARM

- RIAH-TA-101(1) (Local) Refueling platform, el 352' .01-104 YES

- RIAH-TA-201(1) (Local) Refueling platform, el 352' .01-104 YES

- RIAH-TA-102(1) (Local) Turb.deep bed VSL area, el 217 .01-102 YES

- RIAH-TA-103(1) (Local) Turb.deep bed VSL area, el 217 .01-102 YES

-- RIAH-TA-203(1) (Local) Turb.deep bed VSL area, el 217 .01-102 YES

- RIAH-TA-025(1) (Local) Source storage 1 calibration RM, el 191 .1-104 YES CHAPTER 07 7.7-134 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 7.7-3 REFUELING INTERLOCK EFFECTIVENESS REFUELING PLATFORM HOISTS MODE SITUATION POSITION FG CONTROL RODS SWITCH ATTEMPT RESULT 1 Not near core UL All rods in Refuel Move refueling No restrictions platform over core 2 Not near core UL All rods in Refuel Withdraw rods Cannot withdraw more than one rod 3 Not near core UL One rod withdrawn Refuel Move refueling No restrictions platform over core 4 Not near core L One or more rods Refuel Move refueling Platform stopped withdrawn platform over core before over core 5 Not near core UL More than one Startup Move refueling Platform stopped rod withdrawn platform over core before over core 6 Over core UL All rods in Refuel Withdraw rods Cannot withdraw more than one rod 7 Over core Either Hoist L All rods in Refuel Withdraw rods Rod block 8 Not near core UL All rods in Refuel Withdraw rods Rod block 9 Not near core UL All rods in Refuel Operate service No restrictions platform hoist 10 Not near core UL One rod withdrawn Refuel Operate service Hoist operation platform hoist Prevented 11 Not near core UL All rods in Startup Move refueling Platform stopped platform over core before over core 12 Not near core UL All rods in Startup Operate service No restrictions platform hoist 13 Not near core UL One rod withdrawn Startup Operate service Hoist operation platform hoist prevented 14 Not near core UL All rods in Startup Withdraw rods Rod block 15 Not near core UL All rods in Startup Withdraw rods No restrictions 16 Over core UL All rods in Startup Withdraw rods Rod block 17 Over core UL One or more Refuel Lift a fuel (Main) rods withdrawn assembly with Hoist Operation FG prevented FG - Fuel Grapple UL - Unloaded L - Fuel Loaded CHAPTER 07 7.7-135 REV. 15, SEPTEMBER 2010

LGS UFSAR Table 7.7-4 SRM SYSTEM TRIPS TRIP FUNCTION NORMAL SETPOINT TRIP ACTION SRM upscale 105 counts/second Rod block, amber light display annunciator SRM instrument inoperative (1) Rod block, amber light display annunciator Detector retraction 100 counts/second Bypass detector full-in permissive (SRM downscale) limit switch when above preset limit, annunciator, green light display, rod block when below preset limit with IRM range switches on first two ranges.

SRM period 50 sec Annunciator, amber light display SRM downscale 3 counts/second Rod block, annunciator, white light display SRM bypassed White light display (1) SRM is inoperative if module interlock chain is broken, OPERATE-CALIBRATE switch is not in operate position or detector polarizing voltage is below 90% of normal operating voltage (operationally determined).

CHAPTER 07 7.7-136 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 7.7-5 RBM SYSTEM TRIPS NOMINAL TRIP FUNCTION SETPOINT TRIP ACTION RBM upscale To be provided in Rod block, annunciator, Technical Specifications amber light display, RBM ODA RBM inoperative (1)

Rod block, annunciator, amber light display, RBM ODA RBM downscale To be provided in Rod block, annunciator, Technical Specifications white light display, RBM ODA RBM bypassed Manual switch or White light display, peripheral rod selected RBM ODA or APRM reference below 30%

(1)

RBM is inoperative if module interlock chain is broken, chassis mode switch is not in the OPERATE position, less than 50% of available LPRM signals are above 3% threshold, or internal self-test logic indicates trouble.

NOTE: These values are for unit cycle 6 only. The Core Operating Limit Report specifies these values for each cycle.

CHAPTER 07 7.7-137 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 7.7-6 ANALYSIS METHODOLOGY A. Michelson Concern (Reference line failure + single additional failure in a protective channel not dependent on the failed sensing line).

1. Determine the logic for combining channel trips to achieve protective actions.
2. Identify each case where a reactor vessel water level tap or sensing line failure concurrent with an additional random signal electrical failure induces a transient and precludes the automatic operation of a RPS and/or ESF system.
3. For each case identified, demonstrate how the redundancy or diversity of the plant design provides the reactor protection or safety system operation within acceptable limits. Typically, at this point all but one or two cases identified in Step 2 can be eliminated as concerns due to the redundancy of protection and/or ESF systems.

For the one or two remaining cases (worst failure combination scenarios), transient analyses are performed to demonstrate that plant safety is not compromised.

4. Where manual action is required by the operators, identify the instrumentation and time available for the operator to take corrective action.

Note: For LGS, which has 4 instrumentation divisions, automatic protection is available under all scenarios so Step 4 is not required.

CHAPTER 07 7.7-138 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 7.7-7 LGS WATER LEVEL CONDENSING CHAMBER D004A D004B D004C D004D INSTRUMENTATION DIVISION: ECCS 1 2 3 4 RPS/NSSSS IA IB IIA IIB RRCS 1A 2A 1B 2B POWER SUPPLY ECCS(E21-K602) A B C D RPS/NSSSS(B21-K613) A B A B RRCS A B - -

LEVEL TRIP TRANSMITTER TRIP UNIT SYSTEM POINT NUMBER(1) NUMBER(1) SYSTEM LOGIC RPS 3 N080 N680 (A + C) .

(B + D)

NSSSS 3 N080 N680 (A . B) + (C . D)

NSSSS 2 N081 N681 (A . B) + (C . D)

NSSSS (MSIV) 1 N081 N684 (A + (C .

(B + D)

RHR (A) & CS (A) 1 N091 N691 (A . E)

RHR (B) & CS (B) 1 N091 N691 (B . F)

RHR (C) & CS (C) 1 N091 N691 (C . G)

RHR (D) & CS (D) 1 N091 N691 (D . H)

ADS (A) 1/3 N091(2) N691(2) (A1.E1.A3)

ADS (C) 1/3 N091(2) N691(2) (C1.G1.C3)

HPCI 2 N091 N692B, F (B + D) .

N091 N692D, H (F + H) .

CHAPTER 07 7.7-139 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 7.7-7 (Cont'd)

CONDENSING CHAMBER D004A D004B D004C D004D INSTRUMENTATION DIVISION: ECCS 1 2 3 4 RPS/NSSSS IA IB IIA IIB RRCS 1A 2A 1B 2B POWER SUPPLY ECCS(E21-K602) A B C D RPS/NSSSS(B21-K613) A B A B RRCS A B - -

LEVEL TRIP TRANSMITTER TRIP UNIT SYSTEM POINT NUMBER(1) NUMBER(1) SYSTEM LOGIC RCIC 2 N091A, E N692A, E (A + A') . } A & E = N692A, E N097A, E N697A, E (E + E') A' & E' = N697A, E HPCI Trip 8 N091 N693 (B +F) . (D + H)

RCIC Trip 8 N091A, E N693A, E (A + A') . } A & E = N693A, E (E + E') A & E' = N698A, E RPT 2 N402 Analog trip module (A . E) + (B . F)

ARI 2 N402 Analog trip module (A . E) + (B . F)

(1)

All instrument numbers are prefixed with B21- unless otherwise noted.

(2)

Level 3 signal for ADS logic is from transmitter B21-N095 and trip unit B21-N695.

CHAPTER 07 7.7-140 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 7.7-8 TREND PLOT LIMIT TAGS FOR RPV CONTROL DISPLAY CONTROL PARAMETERS STATIC LIMIT DYNAMIC LIMITS RPV water level Trip Hi, Scram Lo, ADS Level TAF None RPV pressure Scram Hi, 100% BPV(1) SRV TPLL Reactor power APRM DNSCL None RPV temperature None None (1)

Indicates a permissive limit CHAPTER 07 7.7-141 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 7.7-9 CONTAINMENT CONTROL DISPLAY TREND PLOT LIMITS CONTROL PARAMETERS STATIC LIMIT DYNAMIC LIMIT Containment Level None Drywell Pressure Oper Hi Spray Wetwell Pressure Oper Hi Pressurization Limit-A, Pressurization Limit-B, Pressure-Suppression, Spray Drywell Temperature Design, Oper Hi RPV Sat Suppression Pool Scram Temp, Oper Hi Temperature Suppression Pool Oper Hi, Oper Lo, SRV Tail Pipe Level Level VAC BRKR Limit CHAPTER 07 7.7-142 REV. 13, SEPTEMBER 2006

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