ML21123A303
| ML21123A303 | |
| Person / Time | |
|---|---|
| Site: | Turkey Point  |
| Issue date: | 05/03/2021 |
| From: | NRC/RGN-II |
| To: | Florida Power & Light Co |
| References | |
| Download: ML21123A303 (125) | |
Text
SRO EXAM KEY ML21123A303 RO 01 C RO 02 D RO 03 B RO 04 D RO 05 A RO 06 B RO 07 B RO 08 C RO 09 B RO 10 A RO 11 B RO 12 A RO 13 B RO 14 B RO 15 A RO 16 C RO 17 C RO 18 C RO 19 A RO 20 B RO 21 A RO 22 C RO 23 C RO 24 D RO 25 C RO 26 C RO 27 C RO 28 B RO 29 A RO 30 A RO 31 D RO 32 D RO 33 D RO 34 A RO 35 C RO 36 B RO 37 B RO 38 D RO 39 A RO 40 D RO 41 A RO 42 D RO 43 A RO 44 C RO 45 B RO 46 C RO 47 C RO 48 B
SRO EXAM KEY RO 49 B RO 50 B RO 51 C RO 52 A RO 53 B RO 54 B RO 55 C RO 56 A RO 57 A RO 58 C RO 59 D RO 60 B RO 61 D RO 62 B RO 63 C RO 64 C RO 65 A RO 66 C RO 67 B RO 68 D RO 69 B RO 70 D RO 71 A RO 72 C RO 73 B RO 74 C RO 75 C SRO 1 / 76 D SRO 2 / 77 B SRO 3 / 78 C SRO 4 / 79 B SRO 5 / 80 B SRO 6 / 81 C SRO 7 / 82 D SRO 8 / 83 B SRO 9 / 84 B SRO 10 / 85 C SRO 11 / 86 A SRO 12 / 87 C SRO 13 / 88 C SRO 14 / 89 C SRO 15 / 90 A SRO 16 / 91 D SRO 17 / 92 D SRO 18 / 93 C SRO 19 / 94 C SRO 20 / 95 C SRO 21 / 96 A
SRO EXAM KEY SRO 22 / 97 B SRO 23 / 98 D SRO 24 / 99 A SRO 25 / 100 B
L-21-1 NRC EXAM SECURE INFORMATION RO Question # 1 Given the following:
Unit 3 reactor and turbine power are stable at 20%.
3-SMI-049.01A, Reactor Protection System Logic Test, is in progress for the test of the A RPS Train.
The Power Range Hi Nuclear Flux (Hi Setpoint) logic test is in progress.
Subsequently:
EHC Supply Header Pressure lowers to 900 psig.
The RO observes the following on the console:
Which one of the following completes the statements below?
IAW with the above light indications the control rods (1) expected to have fallen into the core.
The crew (2) required to enter 3-EOP-E-0, Reactor Trip or Safety Injection, NEXT.
A. (1) are (2) is B. (1) are (2) is NOT C. (1) are NOT (2) is D. (1) are NOT (2) is NOT Page 1
L-21-1 NRC EXAM SECURE INFORMATION RO Question # 2 Given the following:
Unit 4 experiences a Steam Space LOCA from 100% power.
The crew is performing 4-EOP-E-1, Loss of Reactor or Secondary Coolant.
RCS pressure is 1300 psig and stable.
Containment temperature is 195°F and rising.
Which one of the following completes the statements below?
The crew will evaluate RCS pressure using the (1) pressure indicators.
IAW 4-EOP-E-1, RCS pressure SI termination criterion (2) satisfied at this time.
A. (1) PZR (2) is B. (1) PZR (2) is NOT C. (1) RCS wide range (2) is D. (1) RCS wide range (2) is NOT Page 2
L-21-1 NRC EXAM SECURE INFORMATION RO Question # 3 Given the following:
3-EOP-ES-1.2, Post LOCA Cooldown and Depressurization, is in progress.
RCS pressure is 1400 psig.
CET temperatures are 550°F.
Containment temperature is 140°F.
Which one of the following completes the statements below?
IAW 3-EOP-ES-1.2, CET Subcooling is monitored on the foldout page to determine if (1) , the current CET subcooling (2) this foldout page item to be addressed IMMEDIATELY.
A. (1) SI re-initiation is required (2) requires B. (1) SI re-initiation is required (2) does NOT require C. (1) voiding will occur during depressurization (2) requires D. (1) voiding will occur during depressurization (2) does NOT require Page 3
L-21-1 NRC EXAM SECURE INFORMATION RO Question # 4 Given the following:
Unit 3 experiences a LOCA.
The 3B RHR Pump breaker trips and can NOT be closed.
3-EOP-ES-1.3,Transfer to Cold Leg Recirculation, is in progress.
When the crew attempts to open the RHR Suction To Containment Recirc Sump Valves, the following occurs:
MOV-3-860A, Cntmt South Recirc Sump Isolation, does NOT open.
MOV-3-860B, Cntmt North Recirc Sump Isolation, opens.
MOV-3-861A, Cntmt South Recirc Sump Isolation, does NOT open.
MOV-3-861B, Cntmt North Recirc Sump Isolation, opens.
Which one of the following identifies the MINIMUM required LOCAL action(s), if any, to establish a suction flowpath to the 3A RHR Pump?
A. Open both MOV-3-860A AND MOV-3-861A B. ONLY open MOV-3-861A C. ONLY open MOV-3-860A D. A flowpath already exists NO further action is required Page 4
L-21-1 NRC EXAM SECURE INFORMATION RO Question # 5 Given the following:
Unit 3 is at 55% power.
Subsequently:
3A RCP trips Which one of the following completes the statements below?
Unit 3 reactor (1) .
When establishing SG pressure control IAW 3-EOP-ES-0.1, Reactor Trip Response, the RO will expect that (2) .
A. (1) is expected to trip AUTOMATICALLY (2) 3B and 3C SG pressures will be greater than 3A SG pressure B. (1) is expected to trip AUTOMATICALLY (2) 3A, 3B and 3C SG pressures will be equal C. (1) is required to be MANUALLY tripped (2) 3B and 3C SG pressures will be greater than 3A SG pressure D. (1) is required to be MANUALLY tripped (2) 3A, 3B and 3C SG pressures will be equal Page 5
L-21-1 NRC EXAM SECURE INFORMATION RO Question # 6 Given the following:
Unit 3 is at 100% power.
All Charging has been lost.
3-ONOP-047.1, Loss of Charging Flow in Modes 1 Through 4, is in progress.
Which one of the following completes the statements below?
Maintaining CCW cooling flow to the Thermal Barrier HX will prevent the RCP (1) from degrading.
IAW 3-ONOP-047.1, if CCW to the thermal barrier is lost, it (2) be desired that the crew performs a 3-GOP-100, Fast Load Reduction.
A. (1) seals (2) will B. (1) seals (2) will NOT C. (1) lower radial bearing (2) will D. (1) lower radial bearing (2) will NOT Page 6
L-21-1 NRC EXAM SECURE INFORMATION RO Question # 7 Given the following:
The crew has entered 4-ONOP-046.1, Emergency Boration.
Neither of Unit 4's boric acid transfer pumps will start.
Which of the following completes the statement below?
IAW 4-ONOP-046.1, the RO is required NEXT to _____ .
A. open FCV-4-113A, Boric Acid to Blender B. close LCV-4-115C, VCT to Charging Pump Suction C. direct an operator to open 4-358, Bypass Valve for RWST to Charging Pump Suction D. direct an operator to locally open 4-356, Manual Emergency Boration Valve Page 7
L-21-1 NRC EXAM SECURE INFORMATION RO Question # 8 Given the following:
Unit 3 experiences a total loss of CCW.
Which one of the following completes the statements below?
IAW 3-ONOP-030, CCW Malfunction, the crew is required to (1) .
Once CCW flow is restored, a MAXIMUM of (2) heatloads from NCC/CRDMs (counts as one load), ECCs and RHR HXs are allowed to be supplied flow from CCW.
A. (1) ONLY trip the Reactor (2) 5 B. (1) ONLY trip the Reactor (2) 6 C. (1) trip the reactor and the RCPs (2) 5 D. (1) trip the reactor and the RCPs (2) 6 Page 8
L-21-1 NRC EXAM SECURE INFORMATION RO Question # 9 Given the following:
Unit 3 is at 100% power.
Subsequently:
A PZR pressure control malfunction causes a reactor trip and safety injection.
The crew has completed the actions of 3-EOP-E-0, Reactor Trip or Safety Injection, and is evaluating the Critical Safety Functions.
Which one of the following completes the statements below?
IAW 3-EOP-F-0, Critical Safety Function Status Trees, the FIRST step to evaluate the Integrity CSF is to determine if (1) lowered at least (2) in the last 60 minutes.
A. (1) RCS Wide Range Tcold (2) 90°F B. (1) RCS Wide Range Tcold (2) 100°F C. (1) Core Exit Thermocouples (2) 90°F D. (1) Core Exit Thermocouples (2) 100°F Page 9
L-21-1 NRC EXAM SECURE INFORMATION RO Question # 10 Given the following:
Unit 3 is at 60% power.
Subsequently:
Unit 3 experiences a transient that requires the crew to MANUALLY trip the reactor.
The reactor does NOT trip.
3-EOP-FR-S.1, Response to Nuclear Power Generation/ATWS, is in progress.
Which one of the following completes the statements below?
The RO will FIRST (1) .
IAW 3-EOP-FR-S.1, the turbine trip will be confirmed by observation of (2) on TCS.
A. (1) commence control rod insertion in MANUAL (2) turbine valve positions B. (1) commence control rod insertion in MANUAL (2) turbine trip header pressures C. (1) direct the turbine operator to LOCALLY trip the Reactor (2) turbine valve positions D. (1) direct the turbine operator to LOCALLY trip the Reactor (2) turbine trip header pressures Page 10
L-21-1 NRC EXAM SECURE INFORMATION RO Question # 11 Given the following:
Unit 3 experiences a fault on 3C SG.
3-EOP-E-0, Reactor Trip or Safety Injection, is in progress.
Which one of the following completes the statements below?
IAW 3-EOP-E-0, the crew will evaluate in order to determine a transition to 3-EOP-E-2, Faulted Steam Generator Isolation.
A. SG level B. SG pressure C. Containment pressure D. RCS loop T Page 11
L-21-1 NRC EXAM SECURE INFORMATION RO Question # 12 Given the following:
Unit 3 experiences a Reactor trip.
3-EOP-E-0, Reactor Trip or Safety Injection, is in progress.
Post trip steam generator trends on Unit 3 are as follows:
ALSO PROVIDED IN LARGER FORMAT ON THE NEXT PAGE Which one of the following identifies the initiating event?
A. One main feed regulation valve failed closed.
B. A feedwater isolation occurred.
C. One steam dump to condenser failed open.
D. A steamline isolation occurred.
Page 12
L-21-1 NRC EXAM SECURE INFORMATION Page 13
L-21-1 NRC EXAM SECURE INFORMATION RO Question # 13 Given the following:
Both units experience a loss of all AC.
Neither units A nor B 4kV buses can be PROMPTLY re-energized from the control room.
3/4-EOP-ECA-0.0, Loss of all AC power, are in progress.
Which one of the following completes the statements below?
IAW 3/4-EOP-ECA-0.0, the crew is required to shed Vital DC loads within a MAXIMUM of (1) minutes, this allows for powering of the (2) System during a Station Blackout.
A. (1) 90 (2) DCS B. (1) 90 (2) QSPDS C. (1) 120 (2) DCS D. (1) 120 (2) QSPDS Page 14
L-21-1 NRC EXAM SECURE INFORMATION RO Question # 14 Given the following:
Unit 3 is at 100% power.
Subsequently:
Unit 3 experiences a LOOP.
Which one of the following completes the statements below?
IAW 3-EOP-E-0, Reactor Trip or Safety Injection, the RO will realign the 3D 4KV Bus by FIRST manipulating breaker (1) in order to (2) .
A. (1) 3AB19, Feeder to 4KV Bus 3D (2) generate a bus clearing signal B. (1) 3AB19, Feeder to 4KV Bus 3D (2) satisfy an interlock C. (1) 3AA17, Feeder to 4KV Bus 3D (2) generate a bus clearing signal D. (1) 3AA17, Feeder to 4KV Bus 3D (2) satisfy an interlock Page 15
L-21-1 NRC EXAM SECURE INFORMATION RO Question # 15 Given the following:
Unit 3 is in MODE 1.
During rounds the RO was checking the following indication:
ALSO PROVIDED IN LARGER FORMAT ON THE NEXT PAGE Subsequently:
3P08, 120V Vital AC Instrumentation Panel, de-energizes.
Which one of the following completes the statements below IAW 3-ONOP-003.8, Loss of 120V Vital Instrument Panel 3P08?
(1) of the 3A SG NR level instruments will be lost.
(2) of the 3A SG pressure instruments will be lost.
A. (1) one (2) one B. (1) one (2) none C. (1) none (2) one D. (1) none (2) none Page 16
L-21-1 NRC EXAM SECURE INFORMATION Page 17
L-21-1 NRC EXAM SECURE INFORMATION RO Question # 16 Given the following:
Unit 3 is at 100% power.
Unit 3 RO is starting 3A ICW Pump and securing 3B.
Subsequently, upon stopping the 3B ICW Pump:
ANN I4/4 ICW HEADER A/B LO PRESS, alarms.
ICW Pump indication in the control room is as follows:
Which one of the following completes the statements below?
The crew will implement 3-ONOP-019, Intake Cooling Water Malfunction, in order to address a (1) condition on one of the ICW Pumps.
The RO will confirm the ANN I4/4 by observation of (2) on VPA.
A. 1) sheared shaft
- 2) ICW HEADER pressure ONLY B. 1) sheared shaft
- 2) ICW HEADER pressure and flow C. 1) check valve failure
- 2) ICW HEADER pressure ONLY D. 1) check valve failure
- 2) ICW HEADER pressure and flow Page 18
L-21-1 NRC EXAM SECURE INFORMATION RO Question # 17 Given the following:
3CD Instrument Air Compressor is OOS for engine overhaul.
Subsequently:
An electrical fault occurs in the switchyard that results in a loss of offsite power to both Units.
3-ONOP-013, Loss of Instrument Air, is in progress.
PI-3-1444, Instrument Air Pressure, reads 73 psig and lowering.
PI-4-1444, Instrument Air Pressure, reads 108 psig and stable.
Which one of the following completes the statements below IAW 3-ONOP-013?
The crew will operate Unit 3 AFW (1) FCVs in manual.
The crew will operate Unit 4 AFW (2) FCVs in manual.
A. (1) Train 1 (2) Train 1 B. (1) Train 1 (2) Train 2 C. (1) Train 2 (2) Train 1 D. (1) Train 2 (2) Train 2 Page 19
L-21-1 NRC EXAM SECURE INFORMATION RO Question # 18 Given the following:
Unit 3 experienced a reactor trip, due to a loss of main feedwater.
The AFW pumps did NOT start AUTOMATICALLY and attempts to start them were NOT successful.
The crew initiated a feed and bleed of the RCS IAW 3-EOP-FR-H.1, Response to Loss of Secondary Heat Sink.
Subsequently:
One of the AFW pumps is now available to feed the S/Gs.
SG wide range levels are as follows:
3A - 7%
3B - 8%
3C - 8%
IAW 3-EOP-FR-H.1, which one of the following describes the MAXIMUM allowable AFW flow rate, for present plant conditions?
A. The MAXIMUM attainable flow rate, until the RCS feed-and-bleed criteria are no longer met B. The MAXIMUM flow rate that will maintain RCS cooldown rates below 100°F per hour C. A MAXIMUM flow rate of 100 gpm per S/G D. A MAXIMUM total flow rate of 400 gpm Page 20
L-21-1 NRC EXAM SECURE INFORMATION RO Question # 19 Given the following:
Unit 3 is at 88% power.
Two (2) Control Bank D control rods are mechanically bound at 196 steps.
All other Control Bank D rods are at 192 steps.
Which one of the following completes the statements below?
IAW TS 3.1.3.1, Movable Control Assemblies -Group Height, the crew is required to (1) within one hour.
IAW 3-ONOP-028, Reactor Control System Malfunction, immediate operator actions, the crew (2) required to verify the Rod Motion Selector Switch to the MANUAL position.
A. (1) perform 0-OSP-028.8, Shutdown Margin Calculation (2) is B. (1) perform 0-OSP-028.8, Shutdown Margin Calculation (2) is NOT C. (1) perform 3-OSP-059.10, Quadrant Power Tilt Ratio Calculation (2) is D. (1) perform 3-OSP-059.10, Quadrant Power Tilt Ratio Calculation (2) is NOT Page 21
L-21-1 NRC EXAM SECURE INFORMATION RO Question # 20 Given the following:
Unit 3 is at 100% power.
3C Charging Pump is running.
PZR LEVEL CONTROL CHANNEL SELECT, is selected as shown below:
Subsequently:
LT-3-461, Pressurizer Level Transmitter, fails LOW.
The crew just entered 3-ONOP-041.6, Pressurizer Level Control Malfunction.
Which one of the following completes the statements below?
IAW 3-ONOP-041.6, the RO will NEXT place the PZR LEVEL CONTROL CHANNEL SELECT switch to (1) .
CVCS Normal Letdown (2) expected to have AUTOMATICALLY isolated.
A. (1) position 1 (2) is NOT B. (1) position 1 (2) is C. (1) position 2 (2) is NOT D. (1) position 2 (2) is Page 22
L-21-1 NRC EXAM SECURE INFORMATION RO Question # 21 Given the following:
A reactor startup is in progress.
The RO has just completed withdrawing Control Rods and is checking nuclear instrumentation.
POWER ABOVE P-6 status light is lit.
Both SRNIs indicate 5 x 104 cps.
Subsequently:
N-3-32 drifts to 102 cps and is declared INOPERABLE.
Which one of the following completes the statements below?
Intermediate Range NIs (1) used to validate SRNI indication.
IAW 3-ONOP-059.5, Source Range Nuclear Instrumentation Malfunction, the crew (2) required to place N-3-32 Level Trip switch in Bypass.
A. 1) are on scale and will be
- 2) is B. 1) are on scale and will be
- 2) is NOT C. 1) are NOT on scale and thus can NOT be
- 2) is D. 1) are NOT on scale and thus can NOT be
- 2) is NOT Page 23
L-21-1 NRC EXAM SECURE INFORMATION RO Question # 22 Given the following:
Unit 3 is at 6% power and rising IAW 3-GOP-301, Hot Standby to Power Operation.
Subsequently:
NC35F, Hi Flux Trip Bistable Status Light, illuminates.
ANN B5/1, INTERM RANGE HI FLUX ROD STOP, alarms.
ANN C6/5, INTERM RANGE HI FLUX TRIP, alarms.
Unit 3 is at 6.5% power.
Which one of the following identifies the expected crew response IAW 3-ONOP-059.7, Intermediate Range Nuclear Instrumentation Malfunction?
A. Stabilize power at 6.5% until the N-3-35, IRNI, is bypassed B. Stabilize power at 6.5% until the N-3-35, IRNI, is restored C. Trip the reactor and enter EOP network D. Verify IRNI RPS trip is blocked and continue with up power Page 24
L-21-1 NRC EXAM SECURE INFORMATION RO Question # 23 Given the following:
A release of Gas Decay Tank A is in progress IAW 0-NOP-061.14A, Waste Gas Disposal System Controlled Release of Gas Decay Tank A.
Subsequently, R-14, Plant Vent Gaseous Effluent Monitor, HIGH ALARM actuates.
The A Gas Decay Tank Pressure continues to lower.
Which one of the following completes the statements below IAW 3-ONOP-067, Radioactive Effluent Release?
The release flow path will be isolated (1) .
The A GDT (2) required to be sampled before recommencing the gas release.
A. (1) MANUALLY from the Control Room (2) is B. (1) MANUALLY from the Control Room (2) is NOT C. (1) LOCALLY in the Aux Building (2) is D. (1) LOCALLY in the Aux Building (2) is NOT Page 25
L-21-1 NRC EXAM SECURE INFORMATION RO Question # 24 Given the following:
Unit 3 experiences a reactor trip from 100% power.
ANN H1/5, CHRMS HI RADIATION, alarms.
Which one of the following completes the statements below?
IAW 3-ARP-097.CR.H, Control Room Response -Panel H, the RO will validate the ANN H1/5 alarm on an instrument located on (1) .
The TOP of the scale, on RAD-6311A, Containment High Radiation Monitor, is (2) .
A. (1) VPB (2) 1.0x105 R/Hr B. (1) VPB (2) 1.0x108 R/Hr C. (1) 3QR81 and 3QR82 instrument racks (2) 1.0x105 R/Hr D. (1) 3QR81 and 3QR82 instrument racks (2) 1.0x108 R/Hr Page 26
L-21-1 NRC EXAM SECURE INFORMATION RO Question # 25 Given the following:
An alarm is received on Fire Alarm Operator Workstation C41.
Alarming detectors are identified as:
DET108-11, DET108-12 (Train A Inverter Room)
DET108-21, DET108-22 (Train A Inverter Room)
Which one of the following completes the statements below?
IAW 0-ONOP-016.8, Response to a Fire/Smoke Detection System Alarm, immediate operator actions the crew will (1) .
The Inverter Room fire suppression is expected to actuate (2) the receipt of the local fire alarms.
A. (1) Acknowledge and reset the alarms (2) 60 seconds after B. (1) Acknowledge and reset the alarms (2) concurrently with C. (1) Acknowledge but will NOT reset fire alarms (2) 60 seconds after D. (1) Acknowledge but will NOT reset fire alarms (2) concurrently with Page 27
L-21-1 NRC EXAM SECURE INFORMATION RO Question # 26 Given the following:
Unit 3 experiences a Large Break LOCA.
The crew has entered 3-EOP-FR-P.1, Response to Imminent Pressurized Thermal Shock.
IAW 3-EOP-FR-P.1, which one of the following identifies the parameter that will be evaluated to determine whether 3-EOP-FR-P.1 will be performed or if the crew will return to procedure in effect?
A. RVLMS level B. RCS subcooling C. RHR injection flow D. RCS temperature Page 28
L-21-1 NRC EXAM SECURE INFORMATION RO Question # 27 Given the following:
Unit 4 has experienced a LOOP.
4 KV buses are energized by their respective EDGs.
4-EOP-ES-0.2, Natural Circulation Cooldown, is in progress.
Which one of the following completes the statements below IAW 4-EOP-ES-0.2?
The CRDM fans (1) .
Running of CRDM fan(s) will (2) .
A. (1) were expected to AUTOMATICALLY start (2) aid in removal of heat from the upper head of the reactor vessel B. (1) were expected to AUTOMATICALLY start (2) allow a higher/faster cooldown rate C. (1) will be MANUALLY started (2) aid in removal of heat from the upper head of the reactor vessel D. (1) will be MANUALLY started (2) allow a higher/faster cooldown rate Page 29
L-21-1 NRC EXAM SECURE INFORMATION RO Question # 28 Given the following:
Unit 3 is at 100% power.
Subsequently:
ANN H9/1, RCP A MOTOR BEARING HI TEMP, alarms.
3A RCP Motor Bearing Temperature indicates 205°F and slowly rising.
3-ONOP-041.1, Reactor Coolant Pump Off-Normal, is in progress.
Which one of the following identifies the MINIMUM required crew response IAW 3-ONOP-041.1?
A. Reduce power to less than 20%, trip the reactor and then trip 3A RCP.
B. Trip the reactor, and then trip the 3A RCP.
C. Reduce power to less than P-10, trip the reactor and then trip 3A RCP.
D. Notify engineering and increase monitoring of 3A RCP operation.
Page 30
L-21-1 NRC EXAM SECURE INFORMATION RO Question # 29 Given the following:
Unit 4 is at 100% power.
Subsequently:
ANN H8/6, CCW HEAD TANK HI/LO LEVEL, alarms.
Which one of the following completes the statements below?
IAW ANN H8/6 the CVCS systems Seal Water Heat Exchanger tube leakage is verified by .
A. unexplained reactor power rise.
B. lowering VCT level.
C. lowering PRZ level.
D. elevated molybdates levels in the RCS.
Page 31
L-21-1 NRC EXAM SECURE INFORMATION RO Question # 30 Given the following:
Charging Pump Controller configuration is as follows:
Subsequently:
Power is lost to 3P09, Vital AC Panel.
Which one of the following completes the statements below?
3A Charging Pump Controller Automatic operation (1) be affected by the loss of 3P09.
3B Charging Pump Controller Automatic operation (2) be affected by the loss of 3P09.
A. (1) will (2) will B. (1) will (2) will NOT C. (1) will NOT (2) will D. (1) will NOT (2) will NOT Page 32
L-21-1 NRC EXAM SECURE INFORMATION RO Question # 31 Given the following:
Unit 3 is in MODE 4.
3B RHR Cooling Train is in service.
Subsequently:
A tube leak occurs in the 3B RHR Exchanger.
Which one of the following identifies (1) a symptom of the tube leak and (2) assuming no operator action, the response of the RHR Hx Bypass Flow Valve, FCV-3-605?
A. (1) CCW Head Tank level lowers (2) Closes B. (1) CCW Head Tank level lowers (2) Opens C. (1) RCS level lowers (2) Closes D. (1) RCS level lowers (2) Opens Page 33
L-21-1 NRC EXAM SECURE INFORMATION RO Question # 32 Given the following:
Unit 3 experiences a LOCA.
Due to ECCS component failures 3-EOP-ECA-1.1, Loss of Emergency Coolant Recirculation, is in progress.
3A Containment Spray Pump is running with suction aligned to the RWST.
3B Containment Spray Pump is in Pull-To-Lock.
3C and 3A ECCs are running.
RWST level is 70,000 gallons.
Containment pressure is 25 psig.
Which one of the following completes the statements below?
IAW 3-EOP-ECA-1.1, the crew (1) required to start the Normal Containment Coolers.
IAW 3-EOP-ECA-1.1, the crew will determine that the MINIMUM required Containment Spray Pumps in operation is (2) .
REFERENCE PROVIDED A. (1) is (2) 1 B. (1) is (2) 0 C. (1) is NOT (2) 1 D. (1) is NOT (2) 0 Page 34
L-21-1 NRC EXAM SECURE INFORMATION RO Question # 33 Given the following:
Unit 3 experienced a spurious SI.
3-EOP-E-0, Reactor Trip or Safety Injection, is in progress.
RCS pressure is 1939 psig.
FI-3-943, SI Cold Leg Flow, reads 0 gpm.
FI-3-940, SI Hot Leg Flow, reads 0 gpm.
PI-3-943, SI Cold leg Pressure, reads 0 psig.
PI-3-940, SI Hot leg Pressure, reads 1500 psig.
3A and 3B HHSI Pumps are NOT running.
4A and 4B HHSI Pumps are running.
Which one of the following completes the statements below?
The indication for (1) is NOT expected for the given plant conditions, this condition (2) be resolved by starting 3A and 3B HHSI Pumps IAW 3-EOP-E-0, Reactor Trip of Safety Injection.
A. (1) FI-3-943, SI Cold Leg Flow (2) will B. (1) FI-3-943, SI Cold Leg Flow (2) will NOT C. (1) PI-3-943, SI Cold leg Pressure (2) will D. (1) PI-3-943, SI Cold leg Pressure (2) will NOT Page 35
L-21-1 NRC EXAM SECURE INFORMATION RO Question # 34 Given the following:
Unit 3 is at 100% power.
Unit 3 is experiencing elevated RCS leak rate.
PRT level is 55% and slowly rising.
3-ONOP-041.3, Excessive Reactor Coolant System Leakage, is in progress.
Which one of the following completes the statement below?
IAW 3-ONOP-041.3, the crew will identify as the possible cause of the PRT indications.
A. RV-3-551C, PZR Safety Valve leaking by its seat B. RV-3-304, Excess Letdown Relief valve leaking by its seat C. Reactor Vessel Flange leakage D. Increased RCP seal leakoff Page 36
L-21-1 NRC EXAM SECURE INFORMATION RO Question # 35 Which one of the following completes the statements below?
IAW 3-NOP-041.02, Pressurizer Operation, the PZR bubble will begin to form when PZR conditions are (1) at approximately 330 psig.
IAW 3-NOP-041.02, Pressurizer Operation, in order to avoid an unexpected rise in PRT parameters (2) is/are required to be operated carefully.
A. (1) 425°F (2) PC-3-444J, PRESSURIZER PRESSURE CONTROLLER B. (1) 425°F (2) PRESSURIZER BACKUP GROUP HTRs C. (1) 430°F (2) PC-3-444J, PRESSURIZER PRESSURE CONTROLLER D. (1) 430°F (2) PRESSURIZER BACKUP GROUP HTRs Page 37
L-21-1 NRC EXAM SECURE INFORMATION RO Question # 36 Given the following:
Unit 3 is at 100% power.
3D 4Kv bus is aligned to the 3A 4KV bus.
3A CCW Pump is running.
Subsequently:
Unit 3 experiences a LOOP.
3B EDG locks out.
Which one of the following completes the statements below?
Once sequencing is complete the 3A CCW Pump is expected to be (1) and the 3C CCW Pump is expected to be (2) .
A. (1) running (1) running B. (1) running (2) stopped C. (1) stopped (2) running D. (1) stopped (2) stopped Page 38
L-21-1 NRC EXAM SECURE INFORMATION RO Question # 37 Given the following:
Unit 3 is operating at 50% power.
ALL PRZ pressure controls are in AUTO.
PT-3-444, Pressurizer Pressure Transmitter, fails LOW.
Which one of the following completes the statement below?
While the PRZ Pressure Controls remain in AUTO, ONLY PRZ PORV (1) will relieve PRZ pressure.
Assuming no operator action, the plant is expected to (2) .
A. (1) PCV-3-456 (2) trip B. (1) PCV-3-456 (2) remain at power C. (1) PCV-3-455C (2) trip D. (1) PCV-3-455C (2) remain at power Page 39
L-21-1 NRC EXAM SECURE INFORMATION RO Question # 38 Given the following:
Unit 3 is at 100% power Which one of the following completes the statements below?
The de-energization of a MINIMUM of (1) MG set(s) will cause rods to insert into the core.
The MG sets (2) receive a DIRECT trip signal from RPS system on a reactor trip.
A. (1) 1 (2) will B. (1) 1 (2) will NOT C. (1) 2 (2) will D. (1) 2 (2) will NOT Page 40
L-21-1 NRC EXAM SECURE INFORMATION RO Question # 39 Given the following:
Unit 4 is at 50% power.
PT-4-464A, Steam Header Pressure, fails high.
The bi-stables for the failed channel are tripped IAW 4-ONOP-049.1, Deviation or Failure of Safety Related or Reactor Protection Channels.
Which one of the following completes the statements below?
A High Steam Line delta P Safety Injection will be triggered if (1) of the remaining Steam Header Pressure Transmitters fail (2) .
A. (1) 1/2 (2) HIGH B. (1) 1/2 (2) LOW C. (1) 2/2 (2) HIGH D. (1) 2/2 (2) LOW Page 41
L-21-1 NRC EXAM SECURE INFORMATION RO Question # 40 Given the following:
Unit 4 is at 100% power.
Subsequently:
A fault occurs on the 4C SG inside containment.
A Reactor Trip and Safety injection occur.
3B Containment Spray Pump will NOT start Containment temperature is 197°F and lowering.
Containment pressure is 25 psig and lowering.
Which one of the following completes the statements below?
R-3-11, Containment Particulate Monitor, (1) designed to operate under the given conditions.
N-3-43, Power Range NI, (2) designed to operate under the given conditions.
A. (1) is (2) is B. (1) is (2) is NOT C. (1) is NOT (2) is D. (1) is NOT (2) is NOT Page 42
L-21-1 NRC EXAM SECURE INFORMATION RO Question # 41 Given the following:
Unit 3 experiences a LOCA.
3C ECC sustains physical damage and fails to start.
Which one of the following completes the statements below?
3B ECC (1) .
3A ECC (2) .
A. (1) is expected to AUTOMATICALLY start (2) is expected to AUTOMATICALLY start B. (1) is expected to AUTOMATICALLY start (2) is required to be MANUALLY started C. (1) is required to be MANUALLY started (2) is expected to AUTOMATICALLY start D. (1) is required to be MANUALLY started (2) is required to be MANUALLY started Page 43
L-21-1 NRC EXAM SECURE INFORMATION RO Question # 42 Given the following:
Unit 3 experiences a Steam Generator Fault concurrently with a LOOP at 100% power.
3C RCP Breaker will NOT open.
3B SG is completely depressurized.
Which one of the following completes the statements below?
(1) Containment Spray Pump is expected to have electrical power available to it from a 480V (2) .
A. (1) 3B (2) MCC B. (1) 3B (2) Load Center C. (1) 3A (2) MCC D. (1) 3A (2) Load Center Page 44
L-21-1 NRC EXAM SECURE INFORMATION RO Question # 43 Given the following:
Unit 3 is in MODE 6.
Unit 4 is at 100% power.
Subsequently:
4A SG experiences a fault and it completely depressurizes.
Which one of the following identifies the effect on the AFW system?
A. Steam Supply remains available to ALL AFW Pumps B. Steam Supply is lost to the A AFW Pump ONLY C. Steam Supply is lost to the B AFW Pump ONLY D. Steam Supply is lost to the C AFW Pump ONLY Page 45
L-21-1 NRC EXAM SECURE INFORMATION RO Question # 44 Given the following:
Unit 3 is at 70% power.
FT-3-476, 3A SG feed flow transmitter, is selected for control.
FT-3-476 slowly fails HIGH, ANN C6/1, SG A LEVEL DEVIATION / CNTRL TROUBLE, alarms.
Which one of the following completes the statements below?
Actual 3A SG level will initially (1) .
IAW 3-ARP-097.CR, Control Room Response - Panel C, the RO will NEXT (2) .
A. (1) rise (2) place 3A SG Feedwater controller in MANUAL and lower feed flow to return SG level to 50%
B. (1) rise (2) select FT-3-477, 3A SG feed flow transmitter, for control and verify 3A SG level automatically returns to 50%
C. (1) lower (2) place 3A SG Feedwater controller in MANUAL and raise feed flow to return SG level to 50%
D. (1) lower (2) select FT-3-477, 3A SG feed flow transmitter, for control and verify 3A SG level automatically returns to 50%
Page 46
L-21-1 NRC EXAM SECURE INFORMATION RO Question # 45 Which one of the following completes the statements below?
Flow coming from the AFW Pumps discharges (1) of the Feedwater Isolation valves.
Flow coming from the Standby Steam Generators Feed Pumps discharges (2) of the Feedwater Isolation valves.
A. (1) DOWNSTREAM (2) DOWNSTREAM B. (1) DOWNSTREAM (2) UPSTREAM C. (1) UPSTREAM (2) DOWNSTREAM D. (1) UPSTREAM (2) UPSTREAM Page 47
L-21-1 NRC EXAM SECURE INFORMATION RO Question # 46 Given the following:
Unit 3 tripped from 100% power.
Unit 4 is at 100% power.
B AFW Pump is running.
A and C AFW Pumps will NOT start.
All Unit 3 SG NR levels are off scale low.
Which one of the following completes the statement below?
In order to prevent a Red Path on Heatsink Critical Safety Function, the RO is required to maintain a MIMIMUM AFW flow of to Unit 3.
A. 270 gpm B. 340 gpm C. 400 gpm D. 810 gpm Page 48
L-21-1 NRC EXAM SECURE INFORMATION RO Question # 47 Given the following:
Unit 3 is at 100% power.
Subsequently:
Unit 3 has experiences a LOOP.
3A EDG locks out.
Which one of the following completes the statements below (Consider each statement separately)?
If before the LOOP, 3P07, Vital Instrument Panel, was powered by the 3Y01A, CVT for Inverter 3Y01, then 3P07 (1) expected to have power to it.
If before the LOOP, 3P07, Vital Instrument Panel, was powered by the A Spare inverter, then 3P07 (2) expected to have power to it.
A. (1) is (2) is B. (1) is (2) is NOT C. (1) is NOT (2) is D. (1) is NOT (2) is NOT Page 49
L-21-1 NRC EXAM SECURE INFORMATION RO Question # 48 Given the following:
A loss of Vital DC bus 3D23 occurs while at 10% power.
Which one of the following describes the DIRECT effect on the RTBs?
A. 3A RTB opens due to loss of power to the undervoltage trip coil B. 3B RTB opens due to loss of power to the undervoltage trip coil C. 3A RTB opens due to loss of power to the shunt trip coil D. 3B RTB opens due to loss of power to the shunt trip coil Page 50
L-21-1 NRC EXAM SECURE INFORMATION RO Question # 49 In accordance with 0-NOP-003.01, 125V Vital DC System, which one of the choices below completes the following statements?
If two battery chargers are connected to the battery bank, each battery charger is required to have a MINIMUM output of (1) amps or a notification to the Shift Manager is required.
IAW 0-NOP-003.01, When placing a Battery Charger in service the required output voltage range is (2) VDC.
A. (1) 10 (2) 125-129 B. (1) 10 (2) 131-140 C. (1) 20 (2) 125-129 D. (1) 20 (2) 131-140 Page 51
L-21-1 NRC EXAM SECURE INFORMATION RO Question # 50 Given the following:
Unit 3 is at 100% power.
Which one of the following completes the statements below?
If the 3A EDG Master Control Switch is taken to LOCAL, a control room annunciator (1) be triggered; in this configuration, Unit 3 (2) be required to enter into a Tech Spec Action Statement.
A. (1) will (2) will B. (1) will (2) will NOT C. (1) will NOT (2) will D. (1) will NOT (2) will NOT Page 52
L-21-1 NRC EXAM SECURE INFORMATION RO Question # 51 Given the following:
Unit 3 is at 100% power.
Subsequently:
ANN H1/6, PRMS CHANNEL FAILURE, alarms.
R-3-20, Reactor Coolant Letdown Monitor, amber FAIL light is lit.
The detector for R-3-20, Reactor Coolant Letdown Monitor, is failed low.
Which one of the following completes the statements below?
CVCS Normal Letdown flow (1) AUTOMATICALLY isolate.
The crew is required to enter (2) .
A. (1) will (2) 3-ONOP-067, Radioactive Effluent Release B. (1) will (2) 3-ONOP-041.4, Excessive Reactor Coolant System Activity C. (1) will NOT (2) 3-ONOP-067, Radioactive Effluent Release D. (1) will NOT (2) 3-ONOP-041.4, Excessive Reactor Coolant System Activity Page 53
L-21-1 NRC EXAM SECURE INFORMATION RO Question # 52 Given the following:
Unit 3 experiences an SI from 100% power.
Which one of the following completes the statements below?
POV-3-4882 and POV-3-4883, ICW/TPCW isolation valves, (1) AUTOMATICALLY CLOSE.
At this time CCW flow is being supplied to a MINIMUM of (2) of the emergency CCW heat loads.
A. (1) will (2) two B. (1) will (2) three C. (1) will NOT (2) two D. (1) will NOT (2) three Page 54
L-21-1 NRC EXAM SECURE INFORMATION RO Question # 53 Given the following:
Both units are at 100% power.
Subsequently:
Instrument Air Pressures begin to lower on both units.
All Instrument Air Compressors are OFF and will NOT start.
3-ONOP-013 and 4-ONOP-013, Loss of Instrument Air, are in progress.
PI-3-1444, Instrument Air Pressure, reads 73 psig.
PI-4-1444, Instrument Air Pressure, reads 77 psig.
Which one of the following completes the statement below?
Unit 3 running charging pump is expected to go to (1) .
The Service Air System will be (2) aligned to the Instrument Air System.
A. (1) high speed (2) AUTOMATICALLY B. (1) high speed (2) MANUALLY C. (1) low speed (2) AUTOMATICALLY D. (1) low speed (2) MANUALLY Page 55
L-21-1 NRC EXAM SECURE INFORMATION RO Question # 54 Given the following:
Unit 4 is at 100% power.
Subsequently:
Unit 4 experiences a LOCA.
Containment pressure is 15 psig and rising.
MOV-4-626, RCP Thermal Barrier CCW Outlet, is OPEN MOV-4-6386, Seal Return Isolation, is OPEN Which one of the following completes the statements below?
IAW Attachment 3 of 4-EOP-E-0, Reactor Trip of Safety Injection, MOV-4-626 (1) required to be CLOSED.
IAW Attachment 3 of 4-EOP-E-0, Reactor Trip of Safety Injection, MOV-4-6386 (2) required to be CLOSED.
A. (1) is NOT (2) is NOT B. (1) is NOT (2) is C. (1) is (2) is NOT D. (1) is (2) is Page 56
L-21-1 NRC EXAM SECURE INFORMATION RO Question # 55 Which one of the following completes the statements below?
The RO is able to remotely monitor from the control room the position of (1) door(s).
The Personnel Hatch Doors are designed to swing OPEN towards the (2) .
A. (1) ONLY the Containment Personnel Hatch (2) INSIDE of Containment B. (1) ONLY the Containment Personnel Hatch (2) OUTSIDE of Containment C. (1) the Containment Personnel Hatch and Emergency Escape Hatch (2) INSIDE of Containment D. (1) the Containment Personnel Hatch and Emergency Escape Hatch (2) OUTSIDE of Containment Page 57
L-21-1 NRC EXAM SECURE INFORMATION RO Question # 56 Given the following:
N-3-43. Power Range Channel III, and Tave-Tref recorder indication is:
Subsequently RO observes the following:
Which one of the following identifies the event in progress?
A. A continuous rod insertion B. A continuous turbine load reduction C. An inadvertent RCS boration D. A turbine control valve failing open Page 58
L-21-1 NRC EXAM SECURE INFORMATION RO Question # 57 Given the following:
Unit 3 Reactor Startup is in progress.
Power level is 1 x 10-8 amps.
Subsequently:
120V Vital Instrument Panel 3P06 is lost.
Which one of the following completes the statements below?
Power has been lost to (1) , and the reactor (2) expected to trip.
A. (1) IR NI-35 (2) is B. (1) IR NI-35 (2) is NOT C. (1) IR NI-36 (2) is D. (1) IR NI-36 (2) is NOT Page 59
L-21-1 NRC EXAM SECURE INFORMATION RO Question # 58 Given the following:
Unit 3 is at 100% power.
FIC-3-478A, 3A Feed Water Control Valve Primary Controller, experiences a short circuit.
Which one of the following describes the effect this fault will have on the Reactor Protection System?
The FIC-3-478A controller fault is expected to .
A. directly feed back into the protection circuit, preventing the affected RPS channel(s) from tripping B. directly feed back into the protection circuit, causing the affected RPS channel(s) to trip C. NOT directly feed back into the protection circuit due to use of isolation devices D. NOT directly feed back into the protection circuit since separate transmitters are used Page 60
L-21-1 NRC EXAM SECURE INFORMATION RO Question # 59 Which one of the following completes the statements below?
(1) can be relied upon in a post-accident condition, IAW 0-ADM-209, Equipment Tagging and Labeling, this instrument is required to have a (2) label.
A. (1) PI-3-444, Pressurizer Pressure (2) blue B. (1) PI-3-444, Pressurizer Pressure (2) purple C. (1) TI-3-410A, Loop A T-cold Wide Range (2) blue D. (1) TI-3-410A, Loop A T-cold Wide Range (2) purple Page 61
L-21-1 NRC EXAM SECURE INFORMATION RO Question # 60 Given the following:
A core re-load is in progress on Unit 3.
Which one of the following completes the statements below?
The radioactivity level of the Unit 3 Spent Fuel Pool ventilation exhaust, is able to be monitored from the control room (1) .
If a high radiation condition exists on the instrument that monitors Unit 3 SFP ventilation exhaust, the control room (2) be AUTOMATICALLY placed in the ventilation recirculation mode.
A. (1) ONLY on DCS (2) will B. (1) ONLY on DCS (2) will NOT C. (1) on DCS or on 3QR66, PRMS Rack (2) will D. (1) on DCS or on 3QR66, PRMS Rack (2) will NOT Page 62
L-21-1 NRC EXAM SECURE INFORMATION RO Question # 61 Given the following:
Unit 3 is at 100% power.
Subsequently:
Unit 3 experiences a transient that causes SG pressures to rise.
2 Safety valves open on 3A SG.
Which one of the following completes the statements below?
Containment pressure (1) expected to rise as a direct result of 3A SG Safety Valves opening.
If one of the 3A SG Safety Valves fails to re-seat, the crew (2) be able to isolate the failed SG safety valve.
A. (1) is (2) will B. (1) is (2) will NOT C. (1) is NOT (2) will D. (1) is NOT (2) will NOT Page 63
L-21-1 NRC EXAM SECURE INFORMATION RO Question # 62 Given the following:
Unit 3 is at 100% power.
Subsequenly:
RD-3-19, Steam Generator Blowdown Radiation Monitor, fails HIGH.
Which one of the following completes the statements below?
FCV-3-6278 A,B,C Blowdown Flow Control Valves, (1) close AUTOMATICALLY.
CV-3-6275 A,B,C Blowdown Isolation Valves, (2) close AUTOMATICALLY.
A. (1) will (2) will B. (1) will (2) will NOT C. (1) will NOT (2) will D. (1) will NOT (2) will NOT Page 64
L-21-1 NRC EXAM SECURE INFORMATION RO Question # 63 Given the following:
A containment purge is in progress on Unit 3 using 4V20, Unit 4 Containment-Purge Exhaust Fan.
Subsequently:
ANN H1/4, PRMS HI RADIATION, actuates.
The red high-alarm light for RD-3-11, Containment Air Particulate Radiation Monitor, is lit.
Which one of the following completes the statement below?
The purge isolation valves are expected to (1) and 4V20 will be (2) .
A. (1) remain open (2) MANUALLY secured B. (1) remain open (2) AUTOMATICALLY tripped C. (1) AUTOMATICALLY close (2) MANUALLY secured D. (1) AUTOMATICALLY close (2) AUTOMATICALLY tripped Page 65
L-21-1 NRC EXAM SECURE INFORMATION RO Question # 64 Which one of the following completes the statements below?
The RO will check the readout of RAI-6642 and RAI-6643 Control Room Intake Radiation Monitors, on the (1) side of the control room.
The Control Room will AUTOMATICALLY realign for recirc when (2) Control Room Intake Radiation Monitor(s) exceed(s) the actuation setpoint.
A. (1) Unit 3 (2) 1/2 B. (1) Unit 3 (2) 2/2 C. (1) Unit 4 (2) 1/2 D. (1) Unit 4 (2) 2/2 Page 66
L-21-1 NRC EXAM SECURE INFORMATION RO Question # 65 Given the following:
Unit 4 ANN I6/6, XFMR/H2 SEAL OIL DELUGE OPERATING, alarms.
ANN X5/3, SERVICE WTR/ FIRE PUMP/ RWT / WTP TROUBLE, alarms.
A fire is reported on the Unit 4 Main Transformer.
Which one of the following completes the statements below?
The fire pumps receive an AUTOMATIC start signal from a (1) transmitter in the fire main header.
The ARP for ANN X5/3, SERVICE WTR/ FIRE PUMP/ RWT / WTP TROUBLE, (2) direct to perform 0-ARP-097.WTP, Water Treatment Plant Annunciator Response.
A. (1) pressure (2) will B. (1) pressure (2) will NOT C. (1) temperature (2) will D. (1) temperature (2) will NOT Page 67
L-21-1 NRC EXAM SECURE INFORMATION RO Question # 66 Which one of the following completes the statements below?
In accordance with 0-ADM-200, Operations Management Manual:
Night Orders (1) ONLY allowed to be issued by the Operations Director.
Night Orders expire after a MINIMUM of (2) days.
A. (1) are (2) 30 B. (1) are (2) 90 C. (1) are NOT (2) 30 D. (1) are NOT (2) 90 Page 68
L-21-1 NRC EXAM SECURE INFORMATION RO Question # 67 IAW OP-AA-100-1000, Conduct of Operations, which one of the following completes the statements below?
The RO at the controls is required to perform a panel walkdown of key control parameters at least every (1) minutes.
The RO at the controls (2) allowed to initiate a condition report from NAMS for equipment deficiencies.
A. (1) 15 (2) is B. (1) 15 (2) is NOT C. (1) 30 (2) is D. (1) 30 (2) is NOT Page 69
L-21-1 NRC EXAM SECURE INFORMATION RO Question # 68 Given the following:
Unit 3 experiences an event.
3-EOP-E-1, Loss of Reactor or Secondary Coolant, is in progress.
The US assigns the RO to monitor two Continuous Action steps from 3-EOP-E-1.
Subsequently:
The crew transitions to 3-EOP-FR-C.1, Response to Inadequate Core Cooling.
Which one of the following completes the statements below?
The Continuous Actions (1) summarized in the Foldout Page of 3-EOP-E-1.
The Continuous Actions from 3-EOP-E-1 (2) be applicable upon entering 3-EOP-FR-C.1.
A. (1) are (2) will B. (1) are (2) will NOT C. (1) are NOT (2) will D. (1) are NOT (2) will NOT Page 70
L-21-1 NRC EXAM SECURE INFORMATION RO Question # 69 Given the following:
During the performance of an operations surveillance procedure the operator encounters an old setpoint that has changed due to a recent system modification.
Which one of the following completes the statement below?
IAW AD-AA-100-1006, Procedure and Work Instruction Use and Adherence, the crew is expected to .
A. initiate a procedure change while continuing with the work B. stop the work and perform a procedure change before continuing C. stop the work and have a supervisor make pen and ink changes before continuing D. continue with the work and note any deviations for procedure enhancement following completion Page 71
L-21-1 NRC EXAM SECURE INFORMATION RO Question # 70 Given the following:
An infrequently performed evolution is to be performed on Unit 3.
Which one of the following completes the statements below?
IAW OP-AA-1000, Conduct of Infrequently Performed Tests or Evolutions, the (1) is responsible for selecting the criteria for test termination.
IAW OP-AA-1000, in the case that the test will place the plant in an unacceptable condition, the Shift Manager (2) required to have IPTE Manager concurrence in order to terminate the test.
A. (1) Shift Manager (2) is B. (1) Shift Manager (2) is NOT C. (1) IPTE Manager (2) is D. (1) IPTE Manager (2) is NOT Page 72
L-21-1 NRC EXAM SECURE INFORMATION RO Question # 71 Given the following:
Unit 3 is at 100%.
Containment radiation monitors indicate rising trends.
R-3-11 and R-3-12, Containment Air Particulate / Gas Radiation Monitors are in HIGH alarm.
Which one of the following completes the statement below?
IAW 3-ONOP-067, Radioactive Effluent Release, to check the channel operability of R-11/12, the operator will (1) .
Subsequently:
The SNPO reports a failure at the local RM-80 skid.
For the failure of BOTH R-3-11 and R-3-12 a Tech Spec action statement (2) required to be entered.
A. (1) depress the CHECK SOURCE pushbutton and ensure that the readout rises slightly (2) is B. (1) depress the CHECK SOURCE pushbutton and ensure that the readout rises slightly (2) is NOT C. (1) depress the FAIL/TEST pushbutton and ensure that the readout equals 288K or 289K (2) is D. (1) depress the FAIL/TEST pushbutton and ensure that the readout equals 288K or 289K (2) is NOT Page 73
L-21-1 NRC EXAM SECURE INFORMATION RO Question # 72 Which one of the following identifies the TEDE Federal Limit?
A. 1 Rem B. 2 Rem C. 5 Rem D. 10 Rem Page 74
L-21-1 NRC EXAM SECURE INFORMATION RO Question # 73 Given the following:
- The crew is responding to a LOCA on Unit 3.
- The five hottest CETs read between 1250°F and 1300°F.
- The containment pressure is 57 psig.
- Power range NIs are off-scale low.
- Gammametrics read 0.5% power.
- The intermediate range NIs are increasing slowly.
- The crew is evaluating step 15 of 3-EOP-E-0, Check RCS is Intact.
The crew will transition to _____ .
A. 3-EOP-E-1, Loss of Reactor or Secondary Coolant B. 3-EOP-FR-C.1, Inadequate Core Cooling C. 3-EOP-FR-S.1, Response to Nuclear Power Generation/ATWS D. 3-EOP-FR-Z.1, Response to High Containment Pressure Page 75
L-21-1 NRC EXAM SECURE INFORMATION RO Question # 74 Given the following:
Unit 3 is at 100% power.
Work is ongoing in the Unit 3 480V Load Center Rooms.
The foreman requests the fire door between A & B and C & D Load Centers be opened to allow better air circulation for the workers comfort.
A Fire Protection Impairment has NOT been issued.
Which one of the following identifies the policy concerning the propping open of fire doors in accordance with 0-ADM-016, Fire Protection Program?
A. May be open without compensatory actions as long as work is ongoing.
B. May be open if an hourly roving fire watch is provided.
C. Cannot be opened solely for comfort of personnel.
D. Cannot be opened for more than 30 minutes.
Page 76
L-21-1 NRC EXAM SECURE INFORMATION RO Question # 75 IAW 0-ADM-211, Emergency and Off-Normal Operating Procedure Usage, a crew brief is expected to take place prior to transitioning to which one of the following procedures?
A. 3-EOP-ES-1.3, Transfer To Cold Leg Recirculation B. 3-EOP-FR-P.1, Response To Imminent Pressurized Thermal Shock Condition C. 3-EOP-ES-1.2, Post LOCA Cooldown and Depressurization D. 3-EOP-E-3, Steam Generator Tube Rupture Page 77
L-21-1 NRC EXAM SECURE INFORMATION SRO Question # 76 Given the following:
Unit 3 experiences a 3C SG tube rupture coincident with a LOOP.
3-EOP-E-3, Steam Generator Tube Rupture, is in progress.
Normal letdown has NOT been established.
Which one of the following completes the statements below?
In order to control RCS pressure the RO will manipulate (1) .
If a second SG level starts rising uncontrollably the US will (2) .
A. (1) Aux Spray (2) transition to 3-EOP-ECA-3.1, SGTR With Loss Of Reactor Coolant -
Subcooled Recovery Desired B. (1) Aux Spray (2) stabilize the plant and return to step 1 of 3-EOP-E-3 C. (1) One PORV (2) transition to 3-EOP-ECA-3.1, SGTR With Loss Of Reactor Coolant -
Subcooled Recovery Desired D. (1) One PORV (2) stabilize the plant and return to step 1 of 3-EOP-E-3 Page 78
L-21-1 NRC EXAM SECURE INFORMATION SRO Question # 77 Given the following:
Unit 3 trips from 100% power.
The following indications are observed in the control room.
The left half of all alarming annunciators on Unit 3 are DARK.
ANN G4/1, ANNUNCIATOR POWER FAILURE, is LIT.
MOV-3-1405, 3C STM to AFW Pumps, position indicating lights are DARK.
MOV-6459A, A AFW T&T Valve, position indicating lights are DARK.
Which one of the following completes the statements below?
The US is required to implement (1) .
IAW LI-AA-102-1001, Regulatory Reporting, the NRC is required to be notified within a MAXIMUM of (2) ?
REFERENCE PROVIDED A. (1) 3-ONOP-003.4, Loss of DC Busses 3D01 and 3D01A (3A)
(2) 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> B. (1) 3-ONOP-003.4, Loss of DC Busses 3D01 and 3D01A (3A)
(2) 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> C. (1) 3-ONOP-003.5, Loss of DC Busses 3D23 and 3D23A (3B)
(2) 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> D. (1) 3-ONOP-003.5, Loss of DC Busses 3D23 and 3D23A (3B)
(2) 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> Page 79
L-21-1 NRC EXAM SECURE INFORMATION SRO Question # 78 Given the following:
Both Units are at 100% power.
Both Units MVAR loads suddenly start fluctuating.
Subsequently:
Switchyard voltage is 235.0 KV.
Unit 3 Main Generator parameters are:
22 kV 860 MWe 350 MVAR in the LEAD Hydrogen pressure is 65 psig Which one of the following completes the statements below?
The US will direct maintaining Main Generator reactive load below a MAXIMUM of (1) .
The US will direct steps from (2) to mitigate this event.
REFERENCE PROVIDED A. (1) 300 MVAR in the LEAD (2) 3-ONOP-090, Abnormal Generator MW/MVAR Oscillation B. (1) 300 MVAR in the LEAD (2) 0-ONOP-004.6, Degraded Switchyard Voltage C. (1) 200 MVAR in the LEAD (2) 3-ONOP-090, Abnormal Generator MW/MVAR Oscillation D. (1) 200 MVAR in the LEAD (2) 0-ONOP-004.6, Degraded Switchyard Voltage Page 80
L-21-1 NRC EXAM SECURE INFORMATION SRO Question # 79 Given the following:
Unit 4 experiences a reactor trip and safety injection.
The crew is on the last step of 4-EOP-ECA-1.2, LOCA Outside Containment, with the following conditions:
4A Charging Pump is running.
RCS pressure is 1300 psig and rising.
PRZ Level is off-scale low.
Which one of the following completes the statement below?
The US is required to transition NEXT to .
A. 4-EOP-ES-0.0, Rediagnosis B. 4-EOP-E-1, Loss of Reactor or Secondary Coolant C. 4-EOP-ES-1.2, Post LOCA Cooldown and Depressurization D. 4-EOP-ECA-1.1, Loss of Emergency Coolant Recirculation Page 81
L-21-1 NRC EXAM SECURE INFORMATION SRO Question # 80 Given the following:
Unit 3 is at 12% power.
Unit 4 is at 100% power.
The SM declares both 3A HHSI and 3B HHSI Pump INOPERABLE due to installation of incorrect bearing materials.
Replacement parts will be available in 6 days.
Which one of the following completes the statements below?
Assuming RICT is NOT implemented, the MINIMUM Tech Spec required action is to be in MODE 3 within (1) .
If Unit 3 experiences a LOCA the US (2) be required to implement 3-EOP-ECA-1.1, Loss of Emergency Coolant Recirculation.
REFERENCE PROVIDED A. (1) 84 hours9.722222e-4 days <br />0.0233 hours <br />1.388889e-4 weeks <br />3.1962e-5 months <br /> (2) will B. (1) 84 hours9.722222e-4 days <br />0.0233 hours <br />1.388889e-4 weeks <br />3.1962e-5 months <br /> (2) will NOT C. (1) 90 hours0.00104 days <br />0.025 hours <br />1.488095e-4 weeks <br />3.4245e-5 months <br /> (2) will D. (1) 90 hours0.00104 days <br />0.025 hours <br />1.488095e-4 weeks <br />3.4245e-5 months <br /> (2) will NOT Page 82
L-21-1 NRC EXAM SECURE INFORMATION SRO Question # 81 Given the following:
Unit 3 trips from 100% power.
The crew transitions from 3-EOP-ES-0.1, Reactor Trip Response, to 3-EOP-FR-H.1, Response to Loss of Secondary Heat Sink.
RCS bleed and feed is established.
SG NR levels are off scale low.
Subsequently:
ANN A7/1, PRT HI/LO LEVEL HI PRESS/TEMP, is in alarm.
Which one of the following identifies the NEXT required action?
The crew (1) required to reduce the bleed flow in order to avoid containment pressurization.
If the BOP establishes flow from the A Standby SG Feed Pump the US (2) required to IMMEDIATELY, transition out of to 3-EOP-FR-H.1.
A. (1) is (2) is NOT B. (1) is (2) is C. (1) is NOT (2) is NOT D. (1) is NOT (2) is Page 83
L-21-1 NRC EXAM SECURE INFORMATION SRO Question # 82 Given the following:
Unit 3 is at 92% power and stable.
A MANUAL rod withdraw was completed 70 minutes ago.
Control bank D demand is 202 steps.
Control bank D RPIs are as follows:
H-8 indicates 210 steps M-8 indicates 185 steps D-8 indicates 206 steps H-12 indicates 183 steps H-4 indicates 210 steps Which one of the following completes the statements below?
In accordance with Tech Specs, (1) rod(s) is (are) positioned outside of the allowed rod misalignment.
The US will NEXT transition to (2) .
A. (1) one (2) 3-ONOP-028.2, RCC Position Indication Malfunction B. (1) one (2) 3-ONOP-028.1, RCC Misalignment C. (1) two (2) 3-ONOP-028.2, RCC Position Indication Malfunction D. (1) two (2) 3-ONOP-028.1, RCC Misalignment Page 84
L-21-1 NRC EXAM SECURE INFORMATION SRO Question # 83 Given the following:
Unit 3 is shutdown.
RCS Tave is 360°F.
3A Boric Acid Transfer Pump is OOS COLD SHUTDOWN boron concentration is NOT yet achieved.
Subsequently:
Unit 3 experiences an uncontrolled rise in source range counts.
The US announces the entry into 3-ONOP-046.1, Emergency Boration.
The 3B Boric Acid Transfer Pump fails to start.
Which one of the following completes the statements below?
IAW 3-ONOP-046.1, the crew will monitor (1) in order to establish the MINIMUM required boration flow.
Assuming plant conditions remain unchanged, IAW Tech Specs, the crew is required to be in MODE 5 within a MAXIMUM of (2) hours.
REFERENCE PROVIDED A. (1) FI-3-122, Charging Flow (2) 108 B. (1) FI-3-122, Charging Flow (2) 180 C. (1) FI-3-110, Emergency Borate Flow (2) 108 D. (1) FI-3-110, Emergency Borate Flow (2) 180 Page 85
L-21-1 NRC EXAM SECURE INFORMATION SRO Question # 84 Given the following conditions on the NIS during a Unit 3 startup:
(N-3-35 indicates actual power)
Which one of the following completes the statements below?
N-3-36 is (1) .
IAW Tech Specs and 3-ONOP-059.7, Intermediate Range Nuclear Instrumentation Malfunction, if N-3-36 is declared INOPERABLE, the crew (2) allowed to raise power on Unit 3 to 15%.
A. (1) undercompensated (2) is B. (1) undercompensated (2) is NOT C. (1) overcompensated (2) is D. (1) overcompensated (2) is NOT Page 86
L-21-1 NRC EXAM SECURE INFORMATION SRO Question # 85 Given the following:
Unit 3 is at 100% power.
3-OSP-206.2, Quarterly Inservice Valve Testing, is in progress.
CV-3-2826 Testing is SAT.
CV-3-2819 stroke time exceeds the required action time and is declared INOPERABLE.
The following is a diagram of penetration P-63:
ALSO PROVIDED IN LARGER FORMAT ON THE NEXT PAGE Which one of the following identifies MINIMUM required Tech Spec action(s)?
REFERENCE PROVIDED A. ONLY maintain CV-3-2826 OPERABLE B. ONLY close CV-3-2826 within four hours C. ONLY close CV-3-2819 and remove its fuses within four hours D. ONLY close CV-3-2826 and CV-3-2819 within four hours Page 87
L-21-1 NRC EXAM SECURE INFORMATION Page 88
L-21-1 NRC EXAM SECURE INFORMATION SRO Question # 86 Given the following:
3D 4kV bus is OOS on a clearance.
3B RHR pump is in service.
RCS temperature is 127°F.
Refueling cavity is full RCS time to boil is greater than 30 minutes Subsequently:
The switchyard is de-energized.
3B EDG fails to start.
3A EDG is manually loaded onto the 3A 4kV bus after a 5 minute delay.
Switchyard expected time of restoration is 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
Which one of the following completes the statement below?
The Emergency Coordinator will declare an (1) IAW 0-EPIP-20101 and the crew will reinitiate core cooling by (2) .
REFERENCE PROVIDED A. (1) Unusual Event (2) loading an RHR pump on an EDG in accordance with 3-ONOP-004, Loss of Offsite Power B. (1) Unusual Event (2) establishing a secondary heat sink with a SSGFP in accordance 3-ONOP-050, Loss of RHR C. (1) Alert (2) loading an RHR pump on an EDG in accordance with 3-ONOP-004, Loss of Offsite Power D. (1) Alert (2) establishing a secondary heat sink with a SSGFP in accordance 3-ONOP-050, Loss of RHR Page 89
L-21-1 NRC EXAM SECURE INFORMATION SRO Question # 87 Given the following:
Unit 3 is in MODE 4 The 3A train of RHR is in service in the cooldown mode.
3A and 3B CCW Pumps are running.
Subsequently:
ANN H8/4, CCW PP SUCTION HI TEMP, alarms.
Which one of the following completes the statements below?
IAW 3-ARP-097.CR.H, Control Room Response Panel H, the crew (1) required to reduce the RCS cooldown rate.
IAW 3-ARP-097.CR.H, Control Room Response Panel H, the US will refer to and perform applicable actions of (2) .
A. (1) is NOT (2) 3-ONOP-019, Intake Cooling Water Malfunction B. (1) is NOT (2) 3-ONOP-050, Loss of RHR C. (1) is (2) 3-ONOP-019, Intake Cooling Water Malfunction D. (1) is (2) 3-ONOP-050, Loss of RHR Page 90
L-21-1 NRC EXAM SECURE INFORMATION SRO Question # 88 Given the following:
Unit 3 experienced a LOCA today.
Containment pressured reached a MAXIMUM of 34 psig.
Subsequently:
3-EOP-ES-1.3, Transfer to Cold Leg Recirculation, is in progress.
Containment pressure is 17 psig.
Which one of the following completes the statements below?
The crew will NEXT transition to (1) .
(2) Containment Spray Pump is expected to be in operation at the time of this transition.
A. (1) 3-EOP-ES-1.4, Transfer to Hot Leg Recirculation (2) ONE B. (1) 3-EOP-ES-1.4, Transfer to Hot Leg Recirculation (2) NO C. (1) 3-EOP-E-1, Loss of Reactor or Secondary Coolant (2) ONE D. (1) 3-EOP-E-1, Loss of Reactor or Secondary Coolant (2) NO Page 91
L-21-1 NRC EXAM SECURE INFORMATION SRO Question # 89 Given the following:
Unit 3 is in MODE 3 returning from a refueling outage.
Engineering has discovered a manufacturing flaw in the internal wiring of the 3A 4KV Bus which was overhauled during the outage.
Which one of the following completes the statements below?
In the current MODE, the (1) 4KV Buses is/are required to be OPERABLE per Tech Specs, the basis of this requirement (2) to supply safety related equipment required for safe shutdown of the facility.
A. (1) 3A OR 3B (2) is B. (1) 3A OR 3B (2) is NOT C. (1) 3A AND 3B (2) is D. (1) 3A AND 3B (2) is NOT Page 92
L-21-1 NRC EXAM SECURE INFORMATION SRO Question # 90 Given the following:
Unit 4 is at 100% power.
ANN F8/2, EDG A TROUBLE, is in alarm.
The A and B Air Receivers for the 4A EDG are at 155 psig.
The C and D Air Receivers for the 4A EDG are at 200 psig.
The associated electric air compressor has started, but will NOT load.
Which one of the following completes the statements below?
Under the current conditions, IAW UFSAR and Tech Specs the 4A EDG (1) be considered as OPERABLE.
IAW with 4-ARP-097.DG, Diesel Generator Panel Annunciator Response, the crew (2) implement 4-OP-023, Emergency Diesel Generator in order to restore EDG air receiver pressures.
A. (1) will (2) will B. (1) will (2) will NOT C. (1) will NOT (2) will D. (1) will NOT (2) will NOT Page 93
L-21-1 NRC EXAM SECURE INFORMATION SRO Question # 91 Given the following:
Unit 3 is operating at 25% power during a plant shutdown.
Subsequently:
LT-3-459A, PZR Level Prot/Cont, fails LOW.
ANN A9/3, PZR CONTROL HI/LO LEVEL, alarms.
ANN G1/2, CHARGING PUMP HI SPEED, alarms.
Charging flow is 100 GPM and rising.
Which one of the following completes the statements below?
Assuming No Operator Action, a Reactor Protection Trip associated with PZR level (1) AUTOMATICALLY occur.
IAW Tech Specs 3.3.1, LT-3-459A associated bistable(s) is(are) required to be placed in the tripped condition within a MAXIMUM of (2) hours.
A. (1) will NOT (2) 4 B. (1) will NOT (2) 6 C. (1) will (2) 4 D. (1) will (2) 6 Page 94
L-21-1 NRC EXAM SECURE INFORMATION SRO Question # 92 Given the following:
Both Units are at 100% power.
A liquid release is in progress from the A Waste Monitor Tank.
4A2 Circulating Water Pump is OOS.
Subsequently:
ANN H1/4, PRMS HI RADIATION, alarms on Unit 3.
The following indication is present on R-3-18 for 22 minutes while A Waste monitor Tank continues to lower:
ALSO PROVIDED AS A REFERENCE IN LARGER FORMAT Which one of the following completes the statements below?
(1) failed to AUTOMATICALLY isolate the liquid release.
The Emergency Coordinator will declare (2) .
REFERENCE PROVIDED A. (1) SV-3-1413 and SV-3-1414, Radwaste Discharge to Seal Well Solenoids (2) an UNNUSUAL EVENT B. (1) SV-3-1413 and SV-3-1414, Radwaste Discharge to Seal Well Solenoids (2) an ALERT C. (1) RCV-018, Liquid Waste Effluent Isolation Valve (2) an UNNUSUAL EVENT D. (1) RCV-018, Liquid Waste Effluent Isolation Valve (2) an ALERT Page 95
L-21-1 NRC EXAM SECURE INFORMATION Page 96
L-21-1 NRC EXAM SECURE INFORMATION SRO Question # 93 Given the following:
Chemistry is preparing a Gas Decay Tank Release Permit IAW 0-NCOP-004, Section 7.2.
R-14, Plant Vent Gaseous Monitor, is OOS.
Which one of the following completes the statements below?
IAW Tech Specs, with R-14 OOS, Unit 3 (1) required to be in an action statement.
With R-14 OOS (2) .
A. (1) is (2) two independent samples of the tank to be release shall be analyzed B. (1) is (2) the STA shall independently verify the gas release permit prior to submission for approval C. (1) is NOT (2) two independent samples of the tank to be release shall be analyzed D. (1) is NOT (2) the STA shall independently verify the gas release permit prior to submission for approval Page 97
L-21-1 NRC EXAM SECURE INFORMATION SRO Question # 94 Which one of the following completes the statements below IAW 0-EPIP-2101, Duties of Emergency Coordinator?
The Emergency Coordinator or his/ her designee is required to make a Plant Page announcement (1) emergency declarations. Page Volume Boost (1) required to be used during these announcements.
A. (1) ONLY for ALERT or higher (2) is B. (1) ONLY for ALERT or higher (2) is NOT C. (1) for ALL (2) is D. (1) for ALL (2) is NOT Page 98
L-21-1 NRC EXAM SECURE INFORMATION SRO Question # 95 Given the following plant conditions:
Shift complement is at MINIMUM allowed by Tech Specs.
One RO becomes seriously ill and must be taken to the hospital.
There are four hours left until shift change.
Which one of the following describes the required actions IAW Tech Specs?
A. Action must be taken within one hour to identify a relief operator who will arrive within the following three hours.
B. Immediate action must be taken to ensure a relief operator arrives within one hour.
C. Immediate Action must be taken to ensure a relief operator arrives within two hours.
D. No action is required since turnover will occur within four hours.
Page 99
L-21-1 NRC EXAM SECURE INFORMATION SRO Question # 96 Which one of the following completes the statements below?
IAW MA-AA-202, Work Order Execution Process, which one of the following is allowed during an emergency?
SM is allowed to verbally direct work without an approved work order:
A. with no additional approval required.
B. if approval from ONLY the Ops Director is obtained.
C. if approval from ONLY the Maintenance Director is obtained.
D. if approval from BOTH Ops Director and Maintenance Director is obtained Page 100
L-21-1 NRC EXAM SECURE INFORMATION SRO Question # 97 A required annunciator must be defeated in accordance with 0-OSP-200.5, Miscellaneous Tests, Checks, and Operating Evolutions, Attachment 1.
Which one of the following identifies (1) the MAXIMUM time the annunciator may be defeated prior to performing a PCR to the ARP, and (2) the MAXIMUM time allowed for an out of service annunciator prior to performing a 10CFR50.59 applicability/screening review?
A. (1) 7 days (2) 7 days B. (1) 7 days (2) 60 days C. (1) 60 days (2) 7 days D. (1) 60 days (2) 60 days Page 101
L-21-1 NRC EXAM SECURE INFORMATION SRO Question # 98 Given the following:
At 08:00 a.m. Unit 3 experiences a Steam Generator Tube Rupture on 3B SG.
At 08:05 a.m., the EC makes an EAL declaration.
3-EOP-E-3, Steam Generator Tube Rupture, is in progress.
Which one of the following completes the statements below?
IAW 0-EPIP-20201, Duties of Emergency Coordinator, State and Local counties are required to be notified at the LATEST by (1) .
IAW 3-EOP-E-3, Steam Generator Tube Rupture, CV-3-1607, 3B SG Steam Dump to Atmospheric Valve, will be (2) .
A. (1) 08:15 a.m.
(2) CLOSED with controller in Manual B. (1) 08:15 a.m.
(2) Set at 1060 psig with controller in AUTO C. (1) 08:20 a.m.
(2) CLOSED with controller in Manual D. (1) 08:20 a.m.
(2) Set at 1060 psig with controller in AUTO Page 102
L-21-1 NRC EXAM SECURE INFORMATION SRO Question # 99 Given the following:
Unit 3 is in MODE 1.
All Circulating Water pumps are running.
PRMS R-18 Liquid Release Monitor is functional.
The Radioactive Liquid Release Permit and 0-NCOP-003, Recirculation and Sampling Verification Sheet has been reviewed and signed by the Radiochemist to release liquid from the RECYCLE MONITOR TANK A.
Only one Circulating Water pump was required by the release permit Which one of the following completes the statements below?
IAW 0-NCOP-003, Preparation of Liquid Release, the (1) or his/her designee is required to approve the Liquid Release Permit.
If, after completion of the release, it is discovered that R-18 count rate setpoint was a factor of 10 too high and actual count rate did exceed the required R-18 ALARM setpoint, then (2) .
A. (1) Shift Manager (2) the Shift Manager is required to consult 0-EPIP-20101, Duties of the Emergency Coordinator B. (1) Shift Manager (2) Chemistry is required to recalculate the alarm setpoint assuming 3 running Circulating Water pumps C. (1) Chemistry Supervisor (2) the Shift Manager is required to consult 0-EPIP-20101, Duties of the Emergency Coordinator D. (1) Chemistry Supervisor (2) Chemistry is required to recalculate the alarm setpoint assuming 3 running Circulating Water pumps Page 103
L-21-1 NRC EXAM SECURE INFORMATION SRO Question # 100 Given the following:
The emergency response organization (ERO) is activated due to an imminent airborne threat that has been confirmed by Plant Security.
The Shift Manager Becomes incapacitated due to a medical condition.
Which one of the following identifies the NEXT person in order of succession that will assume Emergency Coordinator Responsibilities, IAW 0-EPIP-20101, Duties of Emergency Coordinator?
NOTE: Assume all individuals are Emergency Coordinator qualified.
A. Maintenance Director B. Unit Supervisor C. OPS Director D. OPS AOM Page 104
L-21-1 NRC Exam Reference List RO & SRO References Steam Tables 3-EOP-ECA-1.1 Table SRO Only References LI-AA-102-1001 Unit 3 PCB Section 4 Figure 2 TS 3.1.2.1 TS 3.1.2.2 TS 3.5.2 TS 3.6.4 EAL Tables
REVISION NO.: PROCEDURE TITLE: PAGE:
8A LOSS OF EMERGENCY COOLANT RECIRCULATION 8 of 120 PROCEDURE NO.:
3-EOP-ECA-1.1 TURKEY POINT UNIT 3 STEP ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED
- 4. Determine Containment Spray Requirements (Suction From RWST)
- a. Containment Spray Pump Suction - a. Go to Step 5.
ALIGNED TO RWST
- b. Determine number of Containment Spray Pumps required from Table 1:
Table 1 Containment Spray Pumps Required EMERGENCY SPRAY CONTAINMENT CONTAINMENT RWST LEVEL PUMPS PRESSURE COOLERS REQUIRED RUNNING GREATER THAN
2 55 PSIG GREATER THAN 0 OR 1 2 BETWEEN 20 PSIG 155,000 GALLONS AND 55 PSIG 2 1 LESS THAN 20 PSIG ---- 0 GREATER THAN
2 55 PSIG BETWEEN 60,000 GALLONS 0 OR 1 1 BETWEEN 20 PSIG AND AND 55 PSIG 155,000 GALLONS 2 0 LESS THAN 20 PSIG ---- 0 LESS THAN N/A ---- 0 60,000 GALLONS
REVISION NO.: PROCEDURE TITLE: PAGE:
30 REGULATORY REPORTING 16 of 108 PROCEDURE NO.:
LI-AA-102-1001 NUCLEAR FLEET ADMINISTRATIVE ATTACHMENT 1 REPORTABLE EVENTS (Page 1 of 8)
Declaration of an Emergency Class (See NUREG-1022 Section 3.1.1) 1 Hour Report § 50.72(a)(1)(i) The declaration of any of the Emergency Classes specified in the licensees approved Emergency Plan.
Plant Shutdown Required by Technical Specifications (See NUREG-1022 Section 3.2.1) 4 Hour Report § 50.72(b)(2)(i) The initiation of any 60 Day LER § 50.73(a)(2)(i)(A) The completion nuclear plant shutdown required by the plants of any nuclear plant shutdown required by the Technical Specifications. plants Technical Specifications.
Operation or Condition Prohibited by Technical Specifications (See NUREG-1022 Section 3.2.2) 60 Day LER § 50.73(a)(2)(i)(B) Any operation or condition which was prohibited by the plants Technical Specifications except when:
(1) The Technical Specification is administrative in nature; (2) The event consisted solely of a case of a late surveillance test where the oversight was corrected, the test was performed, and the equipment was found to be capable of performing its specified safety functions; or (3) The Technical Specification was revised prior to discovery of the event such that the operation or condition was no longer prohibited at the time of discovery of the event.
REVISION NO.: PROCEDURE TITLE: PAGE:
30 REGULATORY REPORTING 17 of 108 PROCEDURE NO.:
LI-AA-102-1001 NUCLEAR FLEET ADMINISTRATIVE ATTACHMENT 1 REPORTABLE EVENTS (Page 2 of 8)
Deviation from Technical Specifications Authorized under § 50.54(x)
(See NUREG-1022 Section 3.2.3) 1 Hour Report§ 50.72(b)(1) ... any deviation from the 60 Day LER § 50.73(a)(2)(i)(C) Any deviation from plants Technical Specifications authorized pursuant to § the plants Technical Specifications authorized 50.54(x) of this part. pursuant to § 50.54(x) of this part.
Degraded or Unanalyzed Condition (See NUREG-1022 Section 3.2.4) 8 Hour Report § 50.72(b)(3)(ii) Any event or condition 60 Day LER 50.73(a)(2)(ii) Any event or condition that results in: that resulted in:
(A) The condition of the nuclear power plant, (A) The condition of the nuclear power plant, including its principal safety barriers, being including its principal safety barriers, being seriously degraded; or seriously degraded; or (B) The nuclear power plant being in an (B) The nuclear power plant being in an unanalyzed condition that significantly unanalyzed condition that significantly degrades plant safety. degraded plant safety.
External Threat or Hampering (See NUREG-1022 Section 3.2.5) 60 Day LER § 50.73(a)(2)(iii) Any natural phenomenon or other external condition that posed an actual threat to the safety of the nuclear power plant or significantly hampered site personnel in the performance of duties necessary for the safe operation of the nuclear power plant.
REVISION NO.: PROCEDURE TITLE: PAGE:
30 REGULATORY REPORTING 18 of 108 PROCEDURE NO.:
LI-AA-102-1001 NUCLEAR FLEET ADMINISTRATIVE ATTACHMENT 1 REPORTABLE EVENTS (Page 3 of 8)
System Actuation (See NUREG-1022 Section 3.2.6) 4 Hour Report § 50.72(b)(2)(iv)(A) Any event that results or should have resulted in emergency core cooling system (ECCS) discharge into the reactor coolant system as a result of a valid signal except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation.
4 Hour Report § 50.72(b)(2)(iv)(B) Any event or condition that results in actuation of the reactor protection system (RPS) when the reactor is critical except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation.
8 Hour Report § 50.72(b)(3)(iv)(A) Any event or 60 Day LER § 50.73(a)(2)(iv)(A) Any event or condition that results in valid actuation of any of the condition that resulted in manual or automatic systems listed in paragraph (b)(3)(iv)(B) of this section, actuation of any of the systems listed in paragraph except when the actuation results from and is part of a (a)(2)(iv)(B) of this section, except when:
pre-planned sequence during testing or reactor operation. (1) The actuation resulted from and was part of a pre-planned sequence during testing or reactor operation; or (2) The actuation was invalid and; (i) Occurred while the system was properly removed from service; or (ii) Occurred after the safety function had been already completed.
As indicated in 10 CFR 50.73(a)(1), in the case of an invalid actuation reported under 10 CFR 50.73(a)(2)(iv)(A) other than actuation of the RPS when the reactor is critical, the licensee may, at its option, provide a telephone notification to the NRC Operations Center within 60 days after discovery of the event instead of submitting a written LER.
REVISION NO.: PROCEDURE TITLE: PAGE:
30 REGULATORY REPORTING 19 of 108 PROCEDURE NO.:
LI-AA-102-1001 NUCLEAR FLEET ADMINISTRATIVE ATTACHMENT 1 REPORTABLE EVENTS (Page 4 of 8) 8 Hour Report § 50.72(b)(3)(iv)(B) The systems to § 50.73(a)(2)(iv)(B) The systems to which the which the requirements of paragraph (b)(3)(iv)(A) of requirements of paragraph (a)(2)(iv)(A) of this this section apply are: section apply are:
(1) Reactor protection system (RPS) including: (1) Reactor protection system (RPS) including:
reactor scram and reactor trip. 5 reactor scram or reactor trip.
(2) General containment isolation signals affecting (2) General containment isolation signals affecting containment isolation valves in more than one containment isolation valves in more than one system or multiple main steam isolation valves system or multiple main steam isolation valves (MSIVs). (MSIVs).
(3) Emergency core cooling systems (ECCS) for (3) Emergency core cooling systems (ECCS) for pressurized water reactors (PWRs) including: high- pressurized water reactors (PWRs) including: high-head, intermediate-head, and low-head injection head, intermediate-head, and low-head injection systems and the low pressure injection function of systems and the low pressure injection function of residual (decay) heat removal systems. residual (decay) heat removal systems.
(4) ECCS for boiling water reactors (BWRs) including: (4) ECCS for boiling water reactors (BWRs) high-pressure and low-pressure core spray systems; including: high-pressure and low-pressure core high-pressure coolant injection system; low pressure spray systems; high-pressure coolant injection injection function of the residual heat removal system; low pressure injection function of the system. residual heat removal system.
(5) BWR reactor core isolation cooling system; isolation (5) BWR reactor core isolation cooling system; condenser system; and feedwater coolant injection isolation condenser system; and feedwater system. coolant injection system.
(6) PWR auxiliary or emergency feedwater (6) PWR auxiliary or emergency feedwater system. system.
(7) Containment heat removal and (7) Containment heat removal and depressurization depressurization systems, including systems, including containment spray and fan containment spray and fan cooler systems cooler systems.
(8) Emergency ac electrical power systems, including: (8) Emergency ac electrical power systems, emergency diesel generators (EDGs); hydroelectric including: emergency diesel generators (EDGs);
hydroelectric facilities used in lieu of EDGs at facilities used in lieu of EDGs at the Oconee Station; the Oconee Station; and BWR dedicated and BWR dedicated Division 3 EDGs.
Division 3 EDGs.
5 Actuation of the RPS when the reactor is critical is (9) Emergency service water systems that do not reportable under § 50.72(b)(2)(iv)(B) normally run and that serve as ultimate heat sinks.
REVISION NO.: PROCEDURE TITLE: PAGE:
30 REGULATORY REPORTING 20 of 108 PROCEDURE NO.:
LI-AA-102-1001 NUCLEAR FLEET ADMINISTRATIVE ATTACHMENT 1 REPORTABLE EVENTS (Page 5 of 8)
Event or Condition that Could Have Prevented Fulfillment of a Safety Function (See NUREG-1022 Section 3.2.7) 8 Hour Report § 50.72(b)(3)(v) Any event or 60 Day LER § 50.73(a)(2)(v) Any event or condition condition that at the time of discovery could have that could have prevented the fulfillment of the safety prevented the fulfillment of the safety function of function of structures or systems that are needed to:
structures or systems that are needed to:
(A) Shut down the reactor and maintain it in a (A) Shut down the reactor and maintain it in a safe safe shutdown condition; shutdown condition; (B) Remove residual heat; (B) Remove residual heat; (C) Control the release of radioactive material; or (C) Control the release of radioactive material; or (D) Mitigate the consequences of an accident.
(D) Mitigate the consequences of an accident.
8 Hour Report § 50.72(b)(3)(vi) Events covered in § 50.73(a)(2)(vi) Events covered in paragraph (b)(3)(v) of this section may include one or paragraph (a)(2)(v) of this section may include one or more procedural errors, equipment failures, and/or more procedural errors, equipment failures, and/or discovery of design, analysis, fabrication, construction, discovery of design, analysis, fabrication, and/or procedural inadequacies. However, individual construction, and/or procedural inadequacies.
component failures need not be reported pursuant to However, individual component failures need not be paragraph (b)(3)(v) of this section if redundant reported pursuant to paragraph (a)(2)(v) of this equipment in the same system was operable and section if redundant equipment in the same system available to perform the required safety function. was operable and available to perform the required safety function.
Common Cause Inoperability of Independent Trains or Channels (See NUREG-1022 Section 3.2.8) 60 Day LER § 50.73(a)(2)(vii) Any event where a single cause or condition caused at least one independent train or channel to become inoperable in multiple systems or two independent trains or channels to become inoperable in a single system designed to:
(A) Shut down the reactor and maintain it in a safe shutdown condition; (B) Remove residual heat; (C) Control the release of radioactive material; or (D) Mitigate the consequences of an accident.
REVISION NO.: PROCEDURE TITLE: PAGE:
30 REGULATORY REPORTING 21 of 108 PROCEDURE NO.:
LI-AA-102-1001 NUCLEAR FLEET ADMINISTRATIVE ATTACHMENT 1 REPORTABLE EVENTS (Page 6 of 8)
Radioactive Release (See NUREG-1022 Section 3.2.9) 60 Day LER § 50.73(a)(2)(viii)(A) Any airborne radioactive release that, when averaged over a time period of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, resulted in airborne radionuclide concentrations in an unrestricted area that exceeded 20 times the applicable concentration limits specified in appendix B to part 20, table 2, column 1.
60 Day LER § 50.73(a)(2)(viii)(B) Any liquid effluent release that, when averaged over a time period of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, exceeds 20 times the applicable concentrations specified in appendix B to part 20, table 2, column 2, at the point of entry into the receiving waters (i.e.,
unrestricted area) for all radionuclides except tritium and dissolved noble gases.
Internal Threat or Hampering (See NUREG-1022 Section 3.2.10) 60 Day LER § 50.73(a)(2)(x) Any event that posed an actual threat to the safety of the nuclear power plant or significantly hampered site personnel in the performance of duties necessary for the safe operation of the nuclear power plant including fires, toxic gas releases, or radioactive releases.
Transport of a Contaminated Person Offsite (See NUREG-1022 Section 3.2.11) 8 Hour Report § 50.72(b)(3)(xii) Any event requiring the transport of a radioactively contaminated person to an offsite medical facility for treatment.
REVISION NO.: PROCEDURE TITLE: PAGE:
30 REGULATORY REPORTING 22 of 108 PROCEDURE NO.:
LI-AA-102-1001 NUCLEAR FLEET ADMINISTRATIVE ATTACHMENT 1 REPORTABLE EVENTS (Page 7 of 8)
News Release or Notification of Other Government Agency (See NUREG-1022 Section 3.2.12) 4 Hour Report § 50.72(b)(2)(xi) Any event or situation, related to the health and safety of the public or onsite personnel, or protection of the environment, for which a news release is planned or notification to other government agencies has been or will be made.
Such an event may include an onsite fatality or inadvertent release of radioactively contaminated materials.
Loss of Emergency Preparedness Capabilities (See NUREG-1022 Section 3.2.13) 8 Hour Report § 50.72(b)(3)(xiii) Any event that results in a major loss of emergency assessment capability, offsite response capability, or offsite communications capability (e.g., significant portion of control room indication, emergency notification system, or offsite notification system).
REVISION NO.: PROCEDURE TITLE: PAGE:
30 REGULATORY REPORTING 23 of 108 PROCEDURE NO.:
LI-AA-102-1001 NUCLEAR FLEET ADMINISTRATIVE ATTACHMENT 1 REPORTABLE EVENTS (Page 8 of 8)
Single Cause that Could Have Prevented Fulfillment of the Safety Functions of Trains or Channels in Different Systems (See NUREG-1022 Section 3.2.14) 60 Day LER § 50.73(a)(2)(ix)(A) Any event or condition that as a result of a single cause could have prevented the fulfillment of a safety function for two or more trains or channels in different systems that are needed to:
(1) Shut down the reactor and maintain it in a safe shutdown condition; (2) Remove residual heat; (3) Control the release of radioactive material; or (4) Mitigate the consequences of an accident.
§ 50.73(a)(2)(ix)(B) Events covered in paragraph (ix)(A) of this section may include cases of procedural error, equipment failure, and/or discovery of a design, analysis, fabrication, construction, and/or procedural inadequacy. However, licensees are not required to report an event pursuant to paragraph (ix)(A) of this section if the event results from:
(1) A shared dependency among trains or channels that is a natural or expected consequence of the approved plant design; or (2) Normal and expected wear or degradation.
REACTIVITY CONTROL SYSTEMS 3/4.1.2 BORATION SYSTEMS FLOW PATH - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.1.2.1 As a minimum, one of the following boron injection flow paths shall be OPERABLE and capable of being powered from an OPERABLE emergency power source:
- a. A flow path from the boric acid storage tanks via a boric acid transfer pump and a charging pump to the Reactor Coolant System if the boric acid storage tank in Specification 3.1.2.4a. is OPERABLE, or
- b. The flow path from the refueling water storage tank via a charging pump to the Reactor Coolant System if the refueling water storage tank in Specification 3.1.2.4b. is OPERABLE.
APPLICABILITY: MODES 5 and 6.
ACTION:
With none of the above flow paths OPERABLE or capable of being powered from an OPERABLE emergency power source, suspend all operations involving CORE ALTERATIONS or positive reactivity changes.
SURVEILLANCE REQUIREMENTS 4.1.2.1 At least one of the above required flow paths shall be demonstrated OPERABLE:
- a. In accordance with the Surveillance Frequency Control Program by verifying that the temperature of the rooms containing flow path components is greater than or equal to 62F when a flow path from the boric acid tanks is used, and
- b. In accordance with the Surveillance Frequency Control Program by verifying that each valve (manual, power-operated, or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position.
TURKEY POINT - UNITS 3 & 4 3/4 1-7 AMENDMENT NOS. 263 AND 258
REACTIVITY CONTROL SYSTEMS FLOW PATHS - OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.2 The following boron injection flow paths shall be OPERABLE:
- a. The source path from a boric acid storage tank via a boric acid transfer pump to the charging pump suction*, and
- b. At least one of the two source paths from the refueling water storage tank to the charging pump suction; and,
- c. The flow path from the charging pump discharge to the Reactor Coolant System via the regenerative heat exchanger.
APPLICABILITY: MODES 1, 2, 3, and 4.
ACTION:
- a. With no boration source path from a boric acid storage tank OPERABLE,
- 1. Demonstrate the OPERABILITY of the second source path from the refueling water storage tank to the charging pump suction by verifying the flow path valve alignment; and
- 2. Restore the boration source path from a boric acid storage tank to OPERABLE status within 70 hours8.101852e-4 days <br />0.0194 hours <br />1.157407e-4 weeks <br />2.6635e-5 months <br /> or be in at least HOT STANDBY and borated to a boron concentration equivalent to at least the required SHUTDOWN MARGIN at COLD SHUTDOWN at 200F within the next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />; restore the boration source path from a boric acid storage tank to OPERABLE status within the next 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
- b. With only one boration source path OPERABLE or the regenerative heat exchanger flow path to the RCS inoperable, restore the required flow paths to OPERABLE status within 70 hours8.101852e-4 days <br />0.0194 hours <br />1.157407e-4 weeks <br />2.6635e-5 months <br /> or be in at least HOT STANDBY and borated to a boron concentration equivalent to at least the required SHUTDOWN MARGIN at COLD SHUTDOWN at 200F within the next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />; restore at least two boration source paths to OPERABLE status within the next 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
- c. With the boration source path from a boric acid storage tank and the charging pump discharge path via the regenerative heat exchanger inoperable, within one hour initiate boration to a boron concentration equivalent to the required SHUTDOWN MARGIN at COLD SHUTDOWN at 200F and go to COLD SHUTDOWN as soon as possible within the limitations of the boration and pressurizer level control functions of the CVCS.
- The flow required in Specification 3.1.2.2.a above shall be isolated from the other unit from the boric acid transfer pump discharge to the charging pump suction.
TURKEY POINT - UNITS 3 & 4 3/4 1-8 AMENDMENT NOS. 260 AND 255
REACTIVITY CONTROL SYSTEMS SURVEI LLANCE REQUIREMENTS 4.1.2.2 The above required flow paths shall be demonstrated OPERABLE:
- a. In accordance with the Surveillance Frequency Control Program by verifying that the temperature of the rooms containing flow path components is greater than or equal to 62F when a flow path from the boric acid tanks is used;
- b. In accordance with the Surveillance Frequency Control Program by verifying that each valve (manual, power-operated, or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position;
- c. In accordance with the Surveillance Frequency Control Program by verifying that the flow path required by Specification 3.1.2.2a. and c. delivers at least 16 gpm to the RCS.
TURKEY POINT - UNITS 3 & 4 3/4 1-9 AMENDMENT NOS. 263 AND 258
EMERGENCY CORE COOLING SYSTEMS 3/4.5.2 ECCS SUBSYSTEMS - Tavg GREATER THAN OR EQUAL TO 350oF LIMITING CONDITION FOR OPERATION 3.5.2 The following Emergency Core Cooling System (ECCS) equipment and flow paths shall be OPERABLE:
- a. Four Safety Injection (SI) pumps, each capable of being powered from its associated OPERABLE diesel generator#, with discharge flow paths aligned to the RCS cold legs,*
- b. Two RHR heat exchangers,
- d. A flow path capable of taking suction from the refueling water storage tank as defined in Specification 3.5.4, and
- e. Two flow paths capable of taking suction from the containment sump.
APPLICABILITY: MODES 1, 2, and 3**
ACTION:
- a. With one of the following components inoperable:
- 1. RHR heat exchanger,
- 3. RHR parallel injection flow path, or
- 4. SI parallel injection flow path Restore the inoperable component to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or in accordance with the Risk Informed Completion Time Program, or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
- b. Deleted
- c. With one of the four required Safety Injection pumps or its associated discharge flow path inoperable and the opposite unit in MODE 1, 2, or 3, restore the pump or flow path to OPERABLE status within 30 days or be in at least HOT STANDBY within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
- Only three Safety Injection (SI) pumps (two associated with the unit and one from the opposite unit), each capable of being powered from its associated OPERABLE diesel generator#, with discharge flow paths aligned to the RCS cold leg are required if the opposite unit is in MODE 4, 5, 6 or defueled.
- The provisions of Specification 4.0.4 are not applicable for entry into MODE 3 for the Safety Injection flow paths isolated pursuant to Specification 3.4.9.3 provided that the Safety Injection flow paths are restored to OPERABLE status prior to Tavg exceeding 380oF. Safety Injection flow paths may be isolated when Tavg is less than 380oF.
- Inoperability of the required diesel generators does not constitute inoperability of the associated Safety Injection pumps.
TURKEY POINT - UNITS 3 & 4 3/4 5-3 AMENDMENT NOS. 289 AND 283
EMERGENCY CORE COOLING SYSTEMS 3/4.5.2 ECCS SUBSYSTEMS - Tavg GREATER THAN OR EQUAL TO 350oF LIMITING CONDITION FOR OPERATION
- d. With two of the four required Safety Injection pumps or their associated discharge flow paths inoperable and the opposite unit in MODE 1, 2, or 3, restore one of the two inoperable pumps or flow paths to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or in accordance with the Risk Informed Completion Time Program, or be in at least HOT STANDBY within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. This ACTION applies to both units simultaneously.
- e. With one of the three required Safety Injection pumps or its associated discharge flow path inoperable and the opposite unit in MODE 4, 5, 6, or defueled, restore the pump or flow path to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or in accordance with the Risk Informed Completion Time Program, or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
- f. With a required Safety Injection pump OPERABLE but not capable of being powered from its associated diesel generator, restore the capability within 14 days or in accordance with the Risk Informed Completion Time Program, or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
- g. With an ECCS subsystem inoperable due to an RHR pump or its associated discharge flow path being inoperable, restore the inoperable RHR pump or its associated discharge flow path to OPERABLE status within 7 days or in accordance with the Risk Informed Completion Time Program, or be in as least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
- h. With the suction flow path from the refueling water storage tank inoperable, restore the suction flow path to OPERABLE status within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
TURKEY POINT - UNITS 3 & 4 3/4 5-4 AMENDMENT NOS. 289 AND 283
CONTAINMENT SYSTEMS 3/4.6.4 CONTAINMENT ISOLATION VALVES LIMITING CONDITION FOR OPERATION 3.6.4 Each containment isolation valve shall be OPERABLE with isolation times less than or equal to required isolation times.
APPLICABILITY: MODES 1, 2, 3, and 4.
NOTES: 1. Enter applicable ACTIONS for systems made inoperable by containment isolation valves.
- 2. Enter the ACTION of LCO 3.6.1.2, Containment Leakage, when isolation valve leakage results in exceeding the overall containment leakage rate acceptance criteria.
ACTION:
With one or more isolation valves inoperable, maintain at least one isolation valve OPERABLE in each affected penetration that is open and either:
- a. Restore the inoperable valve(s) to OPERABLE status within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or in accordance with the Risk Informed Completion Time Program, or
- b. Isolate each affected penetration within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or in accordance with the Risk Informed Completion Time Program, by use of at least one deactivated automatic containment isolation valve secured in the isolation position, or
- c. Isolate each affected penetration within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or in accordance with the Risk Informed Completion Time Program, by use of at least one closed manual valve or blind flange, or
- d. Be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
SURVEILLANCE REQUIREMENTS 4.6.4.1 The isolation valves shall be demonstrated OPERABLE prior to returning the valve to service after maintenance, repair or replacement work is performed on the valve or its associated actuator, control or power circuit by performance of a cycling test, and verification of isolation time.
TURKEY POINT - UNITS 3 & 4 3/4 6-16 AMENDMENT NOS. 287 AND 281
CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) 4.6.4.2 Each isolation valve shall be demonstrated OPERABLE during the COLD SHUTDOWN or REFUELING MODE in accordance with the Surveillance Frequency Control Program by:
- a. Verifying that on a Phase A Isolation test signal, each Phase A isolation valve actuates to its isolation position;
- b. Verifying that on a Phase B Isolation test signal, each Phase B isolation valve actuates to its isolation position; and
- c. Verifying that on a Containment Ventilation Isolation test signal, each purge, exhaust and instrument air bleed valve actuates to its isolation position.
4.6.4.3 The isolation time of each power-operated or automatic valve shall be determined to be within its limit when tested in accordance with the INSERVICE TESTING PROGRAM.
3/4.6.5 DELETED 3/4.6.6 DELETED TURKEY POINT - UNITS 3 & 4 3/4 6-17 AMENDMENT NOS. 274 AND 269