ML110410298
| ML110410298 | |
| Person / Time | |
|---|---|
| Site: | Catawba  |
| Issue date: | 02/02/2011 |
| From: | NRC/RGN-II |
| To: | |
| References | |
| Download: ML110410298 (131) | |
Text
2010 Catawba Nuclear Station
- NRC Initial License Examination 1.
With Unit 2 operating at 100% power, the following sequence of events occurs:
1000:
A Train SSPS is placed in TEST.
1005:
BYA is LOCALLY closed in preparation for testing RTA.
1010:
A Train SSPS is placed in NORMAL.
1015:
RTA is opened LOCALLY.
1020:
B Train SSPS is placed in TEST.
1025:
BYB is closed LOCALLY.
Which ONE of the following describes the time at which a reactor trip will FIRST occur?
A.
1005 B.
1015 C.
1020 D.
1025 Page lof 100
2010 Catawba Nuclear Station
- NRC Initial License Examination 2.
Given the following Unit 1 conditions:
Initial The crew is recovering from a faulted Steam Generator outside containment, upstream of the MSIVs.
SI has been terminated.
The crew is maintaining the plant in a stable condition per ES-1.1 (SI Termination).
PZR pressure is 2285 psig and stable.
PZR level is 60% and stable.
Current NC pressure starts to decrease rapidly due to a significant leak on one of the PZR safety valves.
PRT pressure is 15 psig.
Based on the event in progress, which parameter will require the manual reinitiation of Safety Injection, and what will the PZR Safety valve tailpipe temperature be indicating?
A.
Low PZR level; 300°F tailpipe temperature.
B.
Low NC subcooling; 300°F tailpipe temperature.
C.
Low PZR level; 250°F tailpipe temperature.
D.
Low NC subcooling; 250° F tailpipe temperature.
Page 2 of 100
2010 Catawba Nuclear Station - NRC Initial License Examination 3.
Given the following Unit 1 conditions:
The unit has experienced a Reactor Trip and Safety Injection due to a Small-Break LOCA.
The crew has just completed the actions of E-0 (Reactor Trip or Safety Injection).
NV pump flow to the NC system Cold Legs is 390 gpm.
NC system pressure is 1300 psig and stable.
SG pressures are 1092 psig and stable.
NC system subcooling on the ICCM is 22°F and stable.
Which ONE of the following describes plant conditions upon transition to E-1 (Loss of Reactor or Secondary Coolant)?
NC Pumps Runninci?
SGs Needed for Heat Removal?
A.
YES YES B.
YES NO C.
NO YES D.
NO NO Page 3 of 100
2010 Catawba Nuclear Station
- NRC Initial License Examination 4.
Given the following Unit 1 conditions:
The Unit was initially at 100% power.
A double-ended break of Loop 1A Cold Leg occurred.
1 ETB has de-energized due to an overcurrent actuation.
E-1 (Loss of Reactor or Secondary Coolant) was implemented.
The crew has performed the appropriate steps of ES-i.4 (Transfer to Hot Leg Recirculation).
For the above conditions, and per ES-I.4; which I MC-1 1 control panel indication is used to determine if hot leg recirculation has been established; AND is hot leg recirculation flow sufficient?
Control Panel Indication Hot Leg Recirc Flow Sufficient?
A.
ND flow to Hot Legs B and C YES B.
ND flow to Hot Legs B and C NO C.
NI Pump discharge flow YES D.
NI Pump discharge flow NO Page 4 of 100
2010 Catawba Nuclear Station
- NRC Initial License Examination 5.
With Unit I reactor at 11% power, the following NC Pump 1A temperatures are noted:
Time:
1102 1105 1108 Stator Winding 305°F 307°F 312°F Motor Bearing 178°F 182°F 186°F Pump Bearing 216°F 227°F 230°F Which ONE of the following describes the time that an NC Pump 1A parameter FIRST exceeds a value which requires manual action; AND what is/are the manual action(s) required?
A.
1105
- Trip the reactor and then trip NCP 1A.
B.
C.
1108
- Trip the reactor and then trip NCP IA.
D.
1108-Trip NCP 1A ONLY.
Page 5 of 100
2010 Catawba Nuclear Station - NRC Initial License Examination 6.
Given the following Unit 1 conditions:
At 0320, a Loss of All AC Power occurred.
At 0350, ONE operator was dispatched to the Standby Shutdown Facility and placed the Standby Makeup Pump in service.
The Standby Makeup Pump has now been operating for 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 is correct regarding the above conditions?
A.
There should have been TWO operators dispatched, due to Independent Verification requirements.
B.
The Spent Fuel Pool level has decreased by approximately 5 inches.
C.
Damage to the NC Pump seals is expected to have occurred, due to previous loss of seal injection.
D.
Damage to the Standby Makeup Pump has occurred due to excessive operation.
Page 6 of 100
2010 Catawba Nuclear Station
- NRC Initial License Examination 7.
Given the following Unit 1 conditions:
Initial The Unit is in Mode 4.
NC temperature is 220°F and slowly decreasing.
1A ND Pump is running in RHR mode.
lB NC Pump is in service.
Subsequently IRAD-3, All, IEMF1 HI RAD 522 FF-57, AUX BLDG 522 Trip 2 alarm is received.
1 RAD-1, B13, 1 EMF41, AUX BLDG VENT HI RAD Trip 2 alarm is received.
Which ONE of the following describes; (1) the required immediate action(s),
AND (2) why is/are the action(s) taken for this event?
A.
(1)
Secure 1A ND Pump ONLY.
(2)
Reduce the rate of leakage.
B.
(1)
Secure 1A ND Pump AND 1 B NC Pump.
(2)
Reduce the rate of leakage.
C.
(1)
Secure IA ND Pump ONLY.
(2)
Minimize heat input to the system.
D.
(1)
Secure IA ND Pump AND lB NC Pump.
(2)
Minimize heat input to the system.
Page 7 of 100
2010 Catawba Nuclear Station
- NRC Initial License Examination 8.
Given the following Unit I conditions:
The Unit is at 100% power.
BOTH CF pumps tripped.
SSPS failed to process a reactor trip.
Which ONE of the following describes the status of the listed indications? (Assume no operator action.)
lAD-I, BI1, AMSAC Turbine Trip annunciator Turbine Control Panel Turbine Trip Light A.
Alarming LIT B.
Alarming OFF C.
NOT alarming LIT D.
NOT alarming OFF Page 8 of 100
2010 Catawba Nuclear Station
- NRC Initial License Examination 9.
Given the following Unit 1 conditions:
The Unit was at 100% power when a tube rupture on 1 B SIG occurred.
The crew is implementing E-3, (Steam Generator Tube Rupture).
Both motor driven CA Pumps are running.
CA system valve control has NOT been reset.
The UST (Upper Surge Tank) level begins decreasing rapidly.
The following alarms annunciate:
1 AD-5, Eli, CA PUMPS TRAIN A LOSS OF NORM SUCT I AD-5, E12, CA PUMPS TRAIN B LOSS OF NORM SUCT Which ONE of the following completes the statement below?
The MINIMUM level to be maintained in the INTACT SIGs is and 5 seconds after the above annunciators are received, the motor driven CA pumps will A.
11%; start taking suction from RN B.
29%; start taking suction from RN C.
11%; trip on low suction pressure D.
29%; trip on low suction pressure Page 9 of 100
2010 Catawba Nuclear Station
- NRC Initial License Examination 10.
Given the following Unit 1 conditions:
The Unit was initially at 100% power.
A Steam Line break occurs on the Main Steam line from 10 S/G, outside of containment.
Pressurizer pressure is stabilizing at approximately 1400 psig.
Which ONE of the following identifies expected valve positions which are ALL correct?
A.
1NV-15B, (Letdown Containment Isolation)
OPEN 1VI-77B (Vito Containment Isolation)
OPEN B.
INV-15B, (Letdown Containment isolation)
CLOSED lVl-77B (Vito Containment Isolation)
CLOSED C.
1NV-15B, (Letdown Containment isolation)
OPEN 1VI-77B (Vito Containment Isolation)
CLOSED D.
INV-15B, (Letdown Containment Isolation)
CLOSED IVI-77B (Vito Containment Isolation)
OPEN Page 10 of 100
2010 Catawba Nuclear Station - NRC Initial License Examination 11.
Given the following Unit I conditions:
Following a reactor trip the plant experienced a total loss of feedwater.
The condition caused the crew to initiate NC bleed and feed.
Subsequently CAPT #1 (Turbine Driven AFW) was restored to service and the crew is ready to establish feedwater flow to the selected steam generator.
Other plant conditions include:
Selected S/G Wide Range level is 4%.
NC loop hot leg temperature at 558°F.
Core exit thermocouple temperatures are STABLE.
Which ONE of the following completes the statement below?
Per FR-H.1, (Loss of Secondary Heat Sink), feedwater flow will be re-established to the selected S/G at (1 )_ to prevent excessive thermal stresses on the _(2).
A.
(1) less than 100 gpm until WR level is >12%
(2) reactor vessel nozzles B.
(1) less than 100 gpm until WR level is >12%
(2)
SG tubes C.
(1) a rate which results in CETs decreasing (2) reactor vessel nozzles D.
(1) a rate which results in CETs decreasing (2)
SG tubes Page 11 of 100
2010 Catawba Nuclear Station - NRC Initial License Examination 12.
Given the following conditions:
VI compressor E is tagged out.
A Loss of Offsite Power occurs.
The operating crew is performing actions in AP/01A155001022, (Loss of Instrument Air).
VI pressure is 60 psig and decreasing.
Which ONE of the following describes:
(1) what is the primary motive force for the S/G PORVs, considering the current conditions; AND (2) if the operators will have ensured that the Unit air-operated valves are in their failure mode positions?
A.
(1) N2 (2) YES B.
(1) N2 (2) NO C.
(1)Vl (2) NO D.
(1)VI (2) YES Page 12 of 100
2010 Catawba Nuclear Station
- NRC Initial License Examination 13.
Considering the guidance provided in AP/0/A15500/020, (Loss of Nuclear Service Water System):
Which ONE of the following describes:
(1) the RN flow requirement for continuous operation, AND (2) what is a correct sequence of actions for restoring the required flow?
A.
(1)
Total RN system flow less than 25,800 gpm.
(2)
Secure unneeded RN pump(s).
Initiate manual RN strainer backwash on unaffected RN pump(s).
B.
(1)
Each RN pump flow greater than 8600 gpm.
(2)
Establish flow through NS Heat Exchangers.
Secure unneeded RN pump(s).
C.
(1)
Total RN system flow less than 25,800 gpm.
(2)
Initiate manual RN strainer backwash on unaffected RN pump(s).
Secure unneeded RN pump(s).
D.
(1)
Each RN pump flow greater than 8600 gpm.
(2)
Secure unneeded RN pump(s).
Establish flow through NS Heat Exchangers.
Page 13 of 100
2010 Catawba Nuclear Station - NRC Initial License Examination 14.
Unit 1 is operating at 100% power when an air leak develops in the Turbine Building. The Turbine Building NEC reports that the leak is immediately upstream of the air operator for 1CF-6, (CF Pump 1A Recirc. Control).
If the NEC isolates air to stop the leak, 1CF-6 will (1)
, causing main feedwater pump final speed to be (2)
(Assume no actions taken by the Main Control Room Operator during the event.)
(1)
(2)
A.
OPEN HIGHER B.
OPEN LOWER C.
CLOSE HIGHER D.
CLOSE LOWER Page 14 of 100
2010 Catawba Nuclear Station
- NRC Initial License Examination 15.
Given the following Unit 1 conditions:
The Unit is operating at 100% power.
Main Generator hydrogen pressure is at 60 psig.
A grid disturbance then occurs.
The operator observes the following indications in the Control Room:
Turbine Load:
1225 MWs MegaVARS:
- 500 MVARs (a negative value)
Generator Voltage:
20.8 Kilovolts lAD-i, C/4 GEN EXCITATION LIMITER ACTIVE is in alarm.
lAD-I, D15 GEN VOLT REG COMMON TROUBLE is in alarm.
GEN VOLT REG EMERGENCY MANUAL MODE indicating light is LIT.
Which ONE of the following describes how the operator will operate the Turbine Generator controls, per AP/1/A15500/037, (Generator Voltage and Electric Grid Disturbances) to address these conditions?
Reference Provided A.
Maintain turbine load at 1225 MWs.
Manually lower excitation current to change MVARs by a value of 250 MVARs.
B.
Maintain turbine load at 1225 MWs.
Manually raise excitation current to change MVARs by a value of 250 MVARs.
C.
Decrease turbine load to 1180 MWs.
Operate the VOLTAGE ADJUST AUTO setpoint to a higher value.
D.
Decrease turbine load to 1180 MWs.
Operate the VOLTAGE ADJUST AUTO setpoint to a lower value.
Page 15 of 100
2010 Catawba Nuclear Station
- NRC Initial License Examination 16.
Given the following Unit 2 conditions:
The crew has entered ECA-1.2, (LOCA Outside Containment) from E-0, (Reactor Trip or Safety Injection) based upon high radiation levels in the Auxiliary Building.
After closing 2N1-173A (ND Hdr 2A to Cold Legs C&D), the following conditions exist:
NC pressure is approximately 1500 psig and slowly increasing.
FWST level is decreasing at a slightly higher rate.
Which ONE of the following completes the statements below?
The LOCA is _(1)_. The actions for mitigating any adverse parameter trend are _(2).
A.
(1) isolated; (2) stop the 2A ND Pump and close 2FW-27A (ND Pump 2A Suct from FWST).
B.
(1) isolated; (2) Leave the 2A ND Pump running and aligned through 2ND-32A (Train 2A Hot Leg Inj Isol) and 2ND-65B (Train 2B Hot Leg lnj Isol).
C.
(1) NOT isolated; (2) Re-open 2N1-173A, and close 2N1-178B (ND Hdr 2B to Cold Legs A&B).
D.
(1) NOT isolated; (2) Leave 2N1-173A closed and close 2Nl-178B.
Page 16 of 100
2010 Catawba Nuclear Station
- NRC Initial License Examination 17.
Given the following Unit 1 conditions:
The Unit was initially at 100% power.
An accident has occurred.
NC System pressure is 2335 psig.
Only ONE Charging (NV) pump is available.
NO Safety Injection (NI) pumps are available.
The criteria for initiating NC bleed and feed have just been met, per FR-H. 1, (Response to Loss of Secondary Heat Sink).
It has been two hours since the reactor tripped.
Which ONE of the following describes:
(1) if NC heat removal will be adequate; AND (2) what motive force will be used for PORV operation?
A.
(1)
Adequate, since at least one NV OR NI pump wW provide adequate heat removal.
(2)
Nitrogen is aligned ONLY to I NC-32B and 1 NC-34A.
B.
(1)
NOT adequate, since at least one NV AND one NI pump is required for adequate heat removal.
(2)
Instrument Air is used to operate All PZR PORVs.
C.
(1)
Adequate, since at least one NV OR NI pump will provide adequate heat removal.
(2)
Instrument Air is used to operate All PZR PORVs.
D.
(1)
NOT adequate, since at least one NV AND one NI pump is required for adequate heat removal.
(2)
Nitrogen is aligned ONLY to I NC-32B and I NC-34A.
Page 17 of 100
2010 Catawba Nuclear Station
- NRC Initial License Examination 18.
Which ONE of the following is a complete description of the reasons for the reactor operator performing a controlled depressurization of the Steam Generators (SGs) during the performance of ECA-1.1, (Loss of Emergency Coolant Recirculation)?
List of Reasons 1.
Minimize NC dilution potential in case of a subsequent tube rupture.
2.
To establish conditions for injection of the Cold Leg accumulators.
3.
Minimize reactor coolant flow from the LOCA.
4.
To establish conditions for ND system operation.
A.
I and 2 ONLY.
B.
3and4ONLY.
C.
I,2, and 3 ONLY D.
2,3, and 4 ONLY Page 18 of 100
2010 Catawba Nuclear Station
- NRC Initial License Examination 19.
Given the following Unit I conditions:
- The Unit is at 80% power with all systems aligned normal.
- Control rod Bank D is at 180 steps.
- Rod control is in AUTOMATIC.
- Turbine Impulse pressure values are:
Channel I:
693 psig Channel II:
693 psig Channel Ill:
0 psig Which ONE of the following completes the statement below?
For the above conditions, control bank D rods will be continuously _(1)_.
If the rods continue to move after controls are shifted to MANUAL the operator will initiate a reactor trip, and then confirm that the reactor is tripped by observing the _(2)_.
(1)
(2)
A.
Inserting Reactor Trip Breakers GREEN lights LIT, and the Bypass Breaker indicating lights DARK.
B.
Inserting GREEN lights LIT on both the Reactor Trip Breakers and Bypass Breakers.
C.
Withdrawing Reactor Trip Breakers GREEN lights LIT, and the Bypass Breaker indicating lights DARK.
D.
Withdrawing GREEN lights LIT on both the Reactor Trip Breakers and Bypass Breakers.
Page 19 of 100
2010 Catawba Nuclear Station
- NRC Initial License Examination 20.
Given the following Unit I conditions:
The Unit is at 100% power.
Channel I Pressurizer Level has been removed from service for calibration.
Channel II Pressurizer Level begins to malfunction and is drifting LOW.
Which ONE of the following describes expected alarms and how the operators will respond to these conditions?
A.
IAD-2, F19 DCS ALTERNATE ACTION, AND 1AD-6, C/9 PZR HI LEVEL DEV CONTROL Take manual control of 1 NV-294 and reduce charging flow.
B.
1AD-2, F/9 DCS ALTERNATE ACTION, AND IAD-6, C/9 PZR HI LEVEL DEV CONTROL Verify Pressurizer level is automatically controlling at 60%.
C.
IAD-2, F19 DCS ALTERNATE ACTION ONLY Take manual control of 1 NV-294 and reduce charging flow.
D.
IAD-2, F19 DCS ALTERNATE ACTION ONLY Verify Pressurizer level is automatically controlling at 60%.
Page 20 of 100
2010 Catawba Nuclear Station - NRC Initial License Examination 21.
Given the following conditions:
The refueling crew is lowering an irradiated fuel assembly next to a new fuel assembly in the core for Unit 1.
The assembly inadvertently drops completely into the core and a notable amount of bubbles are observed.
The RO notes that source range count rates increased by 0.4 decades and are stabilizing.
Which ONE of the following describes:
(1) the expected alarm; AND (2) how this alarm will impact evacuation of containment?
A.
(1)
IAD-2, D13 & D14 SIR HI FLUX LEVEL AT SHUTDOWN (2)
Causes an automatic actuation of the Containment Evacuation alarm.
B.
(1)
IAD-2, D13 & D14 SIR HI FLUX LEVEL AT SHUTDOWN (2)
Requires Control Room crew to manually actuate the Containment Evacuation alarm.
C.
(1) 1RAD-3, D12 1EMF-17 REACTOR BLDG REFUEL BRIDGE (2)
Causes an automatic actuation of the Containment Evacuation alarm.
D.
(1) 1RAD-3, D12 1EMF-17 REACTOR BLDG REFUEL BRIDGE (2)
Requires Control Room crew to manually actuate the Containment Evacuation alarm.
Page 21 of 100
2010 Catawba Nuclear Station
- NRC Initial License Examination 22.
During a fuel handling accident with VP in operation, which ONE of the following describes required operation of VP and containment closure, and what is the reason per AP/1/A/5500/025 (Damaged Spent Fuel)?
A.
Establish containment closure FIRST to prevent an unmonitored release to the environment.
B.
Establish containment closure FIRST to prevent an unfiltered release to the environment.
C.
Shutdown VP FIRST to prevent drawing a vacuum in containment.
D.
Shutdown VP FIRST, and then establish containment closure to terminate any release as soon as possible.
Page 22 of 100
2010 Catawba Nuclear Station
- NRC Initial License Examination 23.
Given the following Unit 1 conditions:
The Unit is at 100% power.
A fire is occurring on one of the turbine bearings.
Considering the classification of this fire, which ONE of the following is the correct sequence of how the deluge system is actuated, AND when the alarms are received in the control room?
Events (listed in random order):
1.
lAD-i, TURBINE OIL FIRE TURB TRIP, alarms.
2.
OAC alarm is received for high temperature at fire detector.
3.
The deluge valve automatically opens.
4.
Operator activates the breakglass station at IMC9 to open the deluge valve.
A.
1,2,3 B.
2,3,1 C.
2,4,1 D.
1,2,4 Page 23 of 100
2010 Catawba Nuclear Station
- NRC Initial License Examination 24.
Given the following conditions:
The Control Room has been evacuated, due to toxic gases, per AP/1/A/5500/017.
Control has been transferred to the Auxiliary Shutdown Panel (ASP).
A cooldown to COLD SHUTDOWN is in progress.
Pressurizer level is at 15% and slowly decreasing.
SG WR levels are approximately 50% and decreasing.
VCT level is 20% and slowly decreasing.
Which ONE of the following describes a local action which will be performed by the NEO for mitigating the above conditions?
A.
De-energize Pressurizer heaters.
B.
Align NV pump suction from VCT to FWST.
C.
Initiate emergency boration.
D.
Throttle CA flow to the SGs.
Page 24 of 100
2010 Catawba Nuclear Station - NRC Initial License Examination 25.
During implementation of AP/1/A/5500/018, (High Activity in Reactor Coolant), where does 1EMF-48, (NC Sample Line Reactor Coolant) draw its sample from, and which event(s) can it be used to diagnose?
Sample Drawn From Diagnose Which Event A.
Cold Leg failed fuel OR crud burst B.
Cold Leg failed fuel ONLY C.
Hot Leg failed fuel ONLY D.
Hot Leg failed fuel OR crud burst Page 25 of 100
2010 Catawba Nuclear Station
- NRC Initial License Examination 26.
Given the following Unit 1 conditions:
A Main Steam Line Break has occurred inside containment.
Containment pressure is 2.9 psig.
The crew has entered FR-P. 1, (Response to Imminent Pressurized Thermal Shock Condition).
NO NCPs are operating.
An NC System pressure reduction is in progress using a PZR PORV.
NC System subcooling is at 3OF.
RVLIS LR level indicates 70%.
Which ONE of the following describes the required action, per FR-P. 1?
A.
Continue with the depressurization of the NC System.
B.
Dump steam from an intact S/G to increase subcooling.
C.
Start NCP B to promote thermal mixing of NC System and ECCS fluids.
D.
Close the PORV to stop NC System depressurization.
Page 26 of 100
2010 Catawba Nuclear Station
- NRC Initial License Examination 27.
Given the following Unit 1 conditions:
A reactor trip occurred.
The crew was performing ES-0. 1, (Reactor Trip Response) when a loss of off-site power occurred.
The crew then transitions to ES-0.2, (Natural Circulation Cooldown) to cool down the plant.
Which ONE of the following identifies (1) potential subsequent conditions which involve entry into ES-0.3, (Natural Circulation Cooldown with Steam Void In Vessel).
AND (2) why the Reactor Vessel UR level is maintained at greater than 73% during the performance of ES-0.3?
A.
(1) If NCS subcooling drops to less than 20°F.
(2) To prevent loss of natural circulation due to steam voiding in the SIG tubes.
B.
(1) If NCS subcooling drops to less than 20°F.
(2) To ensure pressurizer level is adequate to accommodate void collapse.
C.
(1) If plant conditions require a cooldown rate faster than allowed by ES-0.2.
(2) To prevent loss of natural circulation due to steam voiding in the S/G tubes.
D.
(1) If plant conditions require a cooldown rate faster than allowed by ES-0.2.
(2) To ensure pressurizer level is adequate to accommodate void collapse.
Page 27 of 100
2010 Catawba Nuclear Station - NRC Initial License Examination 28.
Given the following Unit I conditions:
The Unit is in Mode 3 with 1A, lB and 1C NC Pumps in service.
One oil lift pump for ID NCP is in service.
The operator presses the ON pushbutton for 1 D NCP.
1 D NCP did NOT start.
Which ONE of the following describes a condition(s), which prevented 1 D NCP from starting?
A.
Annunciator 1AD-6 C14 (NCP D MTR Upper BRG KC Outlet HI/LO Flow) LIT.
B.
Annunciator 1AD-6 F/4 (NCP D Upper/Lower Oil Reservoir Lo Level) LIT.
C.
Bearing oil lift pressure at 400 psig.
D.
Bearing oil lift pressure at 550 psig AND Annunciator 1AD-6 F14 (NCP D Upper/Lower Oil Reservoir Lo Level) LIT.
Page 28 of 100
2010 Catawba Nuclear Station - NRC Initial License Examination 29.
Given the following Unit I conditions:
The Unit is operating at 40% power.
NCP IC trips on overcurrent.
Assuming no operator action, which ONE of the following describes the effect on the Departure from Nucleate Boiling Ratio (DNBR) AND reactor thermal power?
A.
DNBR will INCREASE.
Reactor power decreases and stabilizes at a new lower thermal power.
B.
DNBR will DECREASE.
Reactor power decreases and stabilizes at a new lower thermal power.
C.
DNBR will INCREASE.
Reactor power initially decreases and then returns to 40% thermal power.
D.
DNBR will DECREASE.
Reactor power initially decreases and then returns to 40% thermal power.
Page 29 of 100
2010 Catawba Nuclear Station
- NRC Initial License Examination 30.
Given the following Unit 1 conditions:
The Unit is at 45% power during a power escalation.
The control power for 1 NV-309, (Seal Water Injection Flow) control valve is lost.
Which ONE of the following describes the effect on NC Pump seal injection flow, and whether LCO 3.5.5, (Seal Injection Flow) entry is required?
Seal Iniection Flow LCO 3.5.5 Entry Conditions A.
Decreases NO B.
Decreases YES C.
Increases NO D.
Increases YES Page 30 of 100
2010 Catawba Nuclear Station - NRC Initial License Examination 31.
Which ONE of the following describes:
(I) one of the criteria given in ES-i.3, (Transfer to Cold Leg Recirculation) for initiation of ND auxiliary containment spray; AND (2) why that criterion is important?
A.
(1) Greater than 50 minutes after the reactor trip.
(2) To meet decay heat removal requirements.
B.
(1) NO greater than 50 minutes after the reactor trip.
(2) To ensure containment pressure will remain below design pressure.
C.
(1) Both ND pumps must be operating.
(2) To ensure containment pressure will remain below design pressure.
D.
(1) Both ND pumps must be operating.
(2) To meet decay heat removal requirements.
Page 31 of 100
2010 Catawba Nuclear Station
- NRC Initial License Examination 32.
Given the following Unit 1 conditions:
The Unit is in Mode 6.
Refueling cavity is filled to 23 feet.
Core reload is in progress.
NC temperature is 145°F.
1A residual heat removal (ND) train is in operation.
A leak has been reported on the 1A ND pump motor cooler. To repair the leak, cooling flow to the motor cooler must be isolated. Maintenance estimates it will take 70 minutes to complete repairs.
What is the reason for having one ND loop in operation in this condition and how does this affect the ability to continue core reload?
A.
Provides an indication of reactor coolant temperature.
Core reload must be stopped.
B.
Ensures that a core Keff of less than or equal to 0.95 is maintained during fuel handling operations.
Core reload must be stopped.
C.
Provide an indication of reactor coolant temperature.
Core reload may continue provided no operations are permitted that would dilute the refueling cavity boron concentration.
D.
Ensures that a core Keff of less than or equal to 0.95 is maintained during fuel handling operations.
Core reload may continue provided no operations are permitted that would dilute the refueling cavity boron concentration.
Page 32 of 100
2010 Catawba Nuclear Station
- NRC Initial License Examination 33.
Given the following Unit 1 conditions:
The Unit was initially at 100% power when a small break LOCA occurred.
A loss of all offsite power has occurred.
1 B DG panel trouble is in alarm due to lube oil pressure at 20 psig.
The crew transitions to ES-i.2, (Post LOCA Cooldown and Depressurization).
An operator has been dispatched to perform ES-i.2, Enclosure 4, (Power Alignment for CLA Valves).
Based on the above conditions:
(1) which accumulator isolation valves can be energized; AND (2) what action should be taken once the valves are energized?
A.
ALL.
Close ALL and leave them energized.
B.
ALL.
Close ALL and then de-energize them.
C.
ONLY TWO.
Close these TWO and vent ALL accumulators.
D.
ONLY TWO.
Close these TWO and vent the OTHER two accumulators.
Page 33 of 100
2010 Catawba Nuclear Station
- NRC Initial License Examination 34.
Given the following Unit 1 conditions:
The Unit has completed NC fill and vent activities following a forced outage.
A nitrogen blanket was placed on the pressurizer during the outage.
NC level is 85% in preparation for drawing a bubble in the pressurizer.
Which ONE of the following describes; (1)
What is the limit on pressurizer heat up rate per Selected Licensee Commitment 16.5-4 (Pressurizer)?
AND (2)
What indication verifies nitrogen venting is complete while drawing a bubble in the pressurizer, per OP/i /A16 100/001, (Unit Startup)?
A.
1.
100°F per hour 2.
PRT temperature equalizes with PZR steam space temperature.
B.
1.
100°Fperhour 2.
PRT level increases without a corresponding PRT pressure increase.
C.
1.
200°Fperhour 2.
PRT temperature equalizes with PZR steam space temperature.
D.
1.
200°F per hour 2.
PRT level increases without a corresponding PRT pressure increase.
Page 34 of 100
2010 Catawba Nuclear Station
- NRC Initial License Examination 35.
With Unit I at 100% power, the values of the four (4) channels of Pressurizer pressure control are as follows:
Channel I
2229 psig Channel II 2237 psig Channel III 2227 psig Channel IV 2232 psig Channel II then experiences a loss of power.
Which ONE of the following completes the following statement?
PRIOR to the Channel II failure, SELECTED Pressurizer pressure value was (1)
AFTER the Channel II failure, SELECTED Pressurizer pressure value is (2)
A.
(1) 2232 psig (2) 2229 psig B.
(1) 2232 psig (2) 2232 psig C.
(1) 2229 psig (2) 2229 psig D.
(1) 2237 psig (2) 2232 psig Page 35 of 100
2010 Catawba Nuclear Station - NRC Initial License Examination 36.
Per Technical Specifications, which ONE of the following reactor trips is required to be operable to provide protection against DNB when the plant is in Mode 2?
A.
Pressurizer low pressure B.
OPDT C.
Low NC loop flow D.
OTDT Page 36 of 100
2010 Catawba Nuclear Station
- NRC Initial License Examination 37.
Given the following Unit 2 conditions:
The Unit is conducting a startup at 6% power.
Intermediate range channel N-35 begins to operate erratically.
Rod motion was stopped.
The N-35 channel LEVEL TRIP switch is in BYPASS.
Which ONE of the following describes which N-35 fuses, if any, can be removed without resulting in a reactor trip?
A.
Control power fuses ONLY.
B.
Instrument power fuses ONLY.
C.
None of the fuses can be removed.
D.
Either the instrument power or control power fuses, but not both at the same time.
Page 37 of 100
2010 Catawba Nuclear Station
- NRC Initial License Examination 38.
Which ONE of the following describes how a loss of Vital I & C power panelboards 1 ERPA and 1 ERPD affects SSPS Train A and Train B?
Train A Logic Bay Train B Logic Bay A.
Energized Energized B.
Energized De-energized C.
De-energized De-energized D.
De-energized Energized Page 38 of 100
2010 Catawba Nuclear Station
- NRC Initial License Examination 39.
Given the following conditions:
The Unit is operating at 100% power.
Pressurizer Pressure Channel Ill has been placed in the required configuration for testing.
Which ONE of the following lists the resulting logic for a Safety Injection initiation on Pressurizer pressure based on the conditions above?
A.
1/3 B.
2/3 C.
1/2 D.
2/2 Page 39 of 100
2010 Catawba Nuclear Station - NRC Initial License Examination 40.
Given the following Unit 1 conditions:
The Unit is at 100% power.
The following Lower Containment Ventilation Units (LCVU) are operating in low speed:
LCVU1A LCVUIB LCVUID Upper Containment Ventilation Units (UCVU) IC is the only operating UCVU.
Containment pressure is at 0.2 psig.
LCVU 1D then jp on overcurrent.
Which ONE of the following describes:
(1) the next required action for providing additional containment cooling, per OP/1/A16450/001, (Containment Ventilation Systems), in addressing the above conditions; AND (2) if this action is NOT successful, AND no further operator actions are taken, which LCO entry, or alarm condition, will FIRST be reached for containment pressure?
A.
(1)
Start LCVU 1 C in low speed, AND start ALL UCVUs.
(2)
LCO 3.6.4, Containment Pressure B.
(1)
Start LCVU 1C in low speed, AND start ALL UCVUs.
(2)
Annunciator LIT which instructs operators to initiate high speed for LCVUs.
C.
(1)
Start LCVU 1 C in low speed ONLY.
(2)
LCO 3.6.4, Containment Pressure D.
(1)
Start LCVU 1C in low speed ONLY.
(2)
Annunciator LIT which instructs operators to initiate high speed for LCVUs.
Page 40 of 100
2010 Catawba Nuclear Station - NRC Initial License Examination 41.
With Unit 1 in MODE 4, which ONE of the following describes a condition associated with the Ice Condenser system which requires an LCO entry with a Required Action Completion Time of within ONE hour?
A.
I AD-13, B18, ICE BED HI TEMP SWITCHES alarms, due to one ice bed at 30°F and slowly increasing.
B.
1 AD-13, C18, GLYCOL EXPANSION TANK HI/LO LVL alarms, due to a low level at 30%.
C. Accumulation of ice in ice bed flow channels resulting in 20% blockage.
D. An ice condenser intermediate deck door temporarily blocked from opening.
Page 41 of 100
2010 Catawba Nuclear Station - NRC Initial License Examination 42.
Which ONE of the following conditions, or annunciator alarms, relating to operation of the Ice Condenser system will automatically close I NF-228A (NE Supply Containment Isolation Valve)?
A.
lAD-i 3, E/7, FLOOR COOLING SYS GLYCOL LO FLOW B.
All Unit I Glycol Chiller Compressors trip.
C.
lAD-i 3, D/8, GLYCOL EXPANSION TNK LO-LO LVL D.
All Unit 1 NE Glycol pumps trip.
Page 42 of 100
2010 Catawba Nuclear Station
- NRC Initial License Examination 43.
Given the following Unit 1 conditions:
Following an accident, containment pressure is 8 psig, and slowly decreasing.
1A and lB containment spray (NS) pumps are running.
The crew has begun the actions of ECA-1.1, (Loss of Emergency Coolant Recirculation).
Refueling Water Storage Tank (FWST) level is 10%, and slowly decreasing.
Efforts to OPEN lNl-185A (ND Pump IA Cont Sump Suct) and lNl-184B (ND Pump lB Cont Sump Suct) have NOT been successful.
Based on the above conditions, which ONE of the following describes:
(1)
How are the NS pumps operated at this point per ECA-1.1; AND (2)
What is the purpose for operating the NS pumps in this manner?
A.
(1)
Shutdown both NS pumps.
(2)
To prevent pump cavitation.
B.
(1)
Secure one NS pump. Continue operating one NS pump aligned to the FWST.
(2)
To conserve FWST inventory.
C.
(1)
Shutdown both NS pumps.
(2)
To conserve FWST inventory.
D.
(1)
Secure one NS pump. Continue operating one NS pump aligned to the FWST.
(2)
To prevent pump cavitation.
Page 43 of 100
2010 Catawba Nuclear Station
- NRC Initial License Examination 44.
Given the following Unit I conditions:
With the Unit initially at 100% power, a steam line rupture occurs inside containment.
The operating crew notes the following trend for containment pressure:
Data Pressure Data Pressure Point Point 1
0.5 psig 8
0.8 psig 2
1.5 psig 9
0.4 psig 3
3.5psig 10 0.3psig 4
3.4 psig 11 0.5 psig 5
3.0 psig 12 0.8 psig 6
2.1 psig 13 1.1 psig 7
1.4 psig 14 1.3 psig The operators have NOT taken any actions regarding ESF signals.
(Note: The term Data Point denotes a time interval.)
Which ONE of the following describes the status of the Containment Spray (NS) Pumps and Spray Valves, based on automatic operation for the three Data Points listed below?
Data Point #2 Data Point #8 Data Point #13 A.
Pumps ON ON ON Valves OPEN OPEN OPEN B.
Pumps ON OFF OFF Valves OPEN CLOSED CLOSED C.
Pumps OFF ON ON Valves CLOSED OPEN OPEN D.
Pumps OFF OFF OFF Valves CLOSED CLOSED CLOSED Page 44 of 100
2010 Catawba Nuclear Station
- NRC Initial License Examination 45.
Given the following Unit 1 conditions:
Unit 1 is at 100% power.
1A SIG has developed a 200 GPD tube leak.
o 1 EMF-33 (CSAE Discharge) Trip 2 light is LIT.
The following annunciator is alarming:
IRAD-1, B/i, 1EMF-33 CSAE EXHAUST HI RAD The reactor operator consults the Annunciator Response Procedure (ARP) and proceeds to Panel IMC13.
At Panel 1 MCi 3, the operator observes the as-found pictured control as seen in the photo below:
UN.HT 1
CSAE EXH For the above conditions, and prior to any operator action, which ONE of the following describes whether the as-found configuration of the control switch and indication in the photo above is CORRECT and why or why not?
A.
CORRECT configuration. Swapover of the CSAE exhaust to Aux. Bldg. (VA) filtered exhaust will ONLY occur when the operator takes the switch to the AUTO position.
B.
INCORRECT configuration. The VA light should be LIT due to automatic swapover caused by I EMF-33 reaching its Trip 2 setpoint.
C.
CORRECT configuration. Swapover of the CSAE exhaust to Aux. Bldg. (VA) filtered exhaust will ONLY occur when the operator OPENS 1ABF-D-i 1 or IABF-D-4.
D.
INCORRECT configuration. The switch should have already been in AUTO, to enable CSAE exhaust automatic swapover to Aux. Bldg. (VA) filtered exhaust immediately upon I EMF-33 reaching its Trip 2 setpoint.
Page 45 of 100
2010 Catawba Nuclear Station
- NRC Initial License Examination 46.
Given the following Unit 1 conditions:
The Unit is at 100% power.
IEMF-71 (SIG A Leakage) is INOPERABLE.
I EMF-74 (SIG D Leakage) Trip 2 is LIT.
1 EMF-33 (CSAE Offgas) Trip 2 is LIT.
The crew is currently taking actions in AP/1/A15500/010, (Reactor Coolant Leak).
For the noted times, the following parameters are also observed:
0330 0332 0334 0336 IA SIG Level (NR) %
65 67 68 66 IBSIGLeveI(NR)%
65 65 65 65 lCSIGLevel(NR)%
64 64 64 64 1DSIGLeveI(NR)%
65 65 65 65 LID Flow gpm 75 45 0
0 Charging Flowgpm 128 129 130 131 PZR Level %
55 51 47 43 Which ONE of the following describes:
(1) the cause of the above conditions; AND (2) what actions are taken to mitigate the event?
A.
(1) 1A SIG tube leakage (2)
Perform a rapid shutdown using AP/1/A15500/009, (Rapid Downpower) per API1O (Reactor Coolant Leak).
B.
(1) 1A SIG tube leakage (2)
Manually trip the reactor and initiate Safety Injection per AP/1/A155001010, (Reactor Coolant Leak) and go to E-0, (Reactor Trip and Safety Injection).
C.
(1)
ID S/G tube leakage (2)
Perform a rapid shutdown using AP/11A15500/009, (Rapid Downpower) per AP/lO (Reactor Coolant Leak).
D.
(1)
ID S/G tube leakage (2)
Manually trip the reactor and initiate Safety Injection per AP/1 0, (Reactor Coolant Leak) and go to E-0, (Reactor Trip and Safety Injection).
Page 46 of 100
2010 Catawba Nuclear Station - NRC Initial License Examination 47.
Given the following Unit 1 conditions:
The crew is shutting down the Unit for a refueling outage, per 0P111A161001002, (Controlling Procedure for Unit Shutdown).
The Unit is at 11% power.
CF flow is 14% for all S/Gs.
CA nozzle swap permissive lights are LIT for all S/Gs.
The operator depresses CA NOZZLE pushbutton.
Which ONE of the following completes the statement below?
When the first CF containment isolation valve has CLOSED, the speed of the operating CF pump will by 150 rpm.
A.
be manually DECREASED B.
automatically DECREASE C.
be manually INCREASED D.
automatically INCREASE Page 47 of 100
2010 Catawba Nuclear Station
- NRC Initial License Examination 48.
Given the following Unit 1 conditions:
The operators are conducting a cooldown in Mode 3.
Taveis5l0°F.
NC system pressure is 1900 psig.
1A CF Pump is maintaining S/G levels.
lB CF Pump is shutdown, per the OP.
ECCS Train A/B PZR PRESS has been blocked.
ECCS Train A/B STM PRESS has been blocked.
Both CA train AUTO-START-DEFEAT buttons have been depressed while engineering performs a special test.
Which ONE of the following describes events, which will, in conjunction with the above plant conditions, cause the motor driven CA pumps to start automatically?
A.
1A CF Pump trips, and causes all S/G levels to decrease below the Lo-Lo level trip setpoint.
B.
NC loop B spray valve, I NC-29, fails open, and results in NC pressure decreasing to 1500 psig.
C.
A steamline rupture in the doghouse causes 1C S/G pressure to decrease to 700 psig at a rate of 200 psig/sec.
D.
A feedline rupture results in ID S/G level decreasing to less than the Lo-Lo level setpoint and containment pressure increasing to 2 psig.
Page 48 of 100
2010 Catawba Nuclear Station
- NRC Initial License Examination 49.
Given the following Unit 1 conditions:
1A RN Pump is in service.
All other essential equipment is in operation on Unit 1, powered from B Train.
OPI1/A163501002, (Diesel Generator) is in progress with the 1A Diesel running in parallel to the grid when the following sequence of events occurs:
Load is reduced on the diesel to 200KW in anticipation of opening the Emergency Breaker.
Prior to opening the Emergency Breaker, the Normal Feeder Breaker from IATC spuriously OPENS.
Which ONE of the following completes the statement below?
The Blackout Sequencer Actuated Train A status light on Sl-14 and the A.
illuminates; 1A DG load increases B.
remains dark; 1A DG load increases C.
illuminates; 1A DG Emergency Breaker trips open, then re-closes 8.5 seconds later D.
remains dark; 1A DG Emergency Breaker trips open, then re-closes 8.5 seconds later Page 49 of 100
2010 Catawba Nuclear Station
- NRC Initial License Examination 50.
Given the following Unit 2 conditions:
The Unit is in Mode 4 during a normal cooldown for refueling.
Train 2A of ND is in service.
ND flow is stable at 3400 gpm.
NC return temperature is 154°F and decreasing.
Which ONE of the following describes:
(1) the effect on NC cooldown rate of losing power to Vital DC Bus 2EPA; AND (2) what action the operator will take to restore the desired cooldown rate, per API2/A155001029, (Loss of Vital or Aux Control Power)?
A.
(1) Rate INCREASES.
(2) Manually adjust setpoint of ND HX 2A OUTLET CTRL (for 2ND-26).
B.
(1) Rate DECREASES.
(2) Place PWR DISCON FOR 2Nl-173A switch to THROT and manually control 2Nl-173A.
C.
(1) Rate INCREASES.
(2) Place PWR DISCON FOR 2Nl-173A switch to THROT and manually control 2N1-173A.
D.
(1) Rate DECREASES.
(2) Manually adjust setpoint of ND HX 2A OUTLET CTRL (for 2ND-26).
Page 50 of 100
2010 Catawba Nuclear Station
- NRC Initial License Examination 51.
Given the following condition:
Inverter 1 KXIA experienced a total loss of output voltage.
Which ONE of the following describes:
(1) an indication used to aid in determining that the backup power supply has aligned to I KXPA; AND (2)
Once 1 KXIA has returned to normal operating parameters, how will I KXPA supply be swapped back to 1 KXIA?
A.
(1) 1KMAA In Sync light is LIT.
(2) Automatically after 60 seconds.
B.
(1) 1KMAA In Sync light is LIT.
(2) Manually.
C.
(1) 1KXAA Alternate Source Supplying Load light is LIT.
(2) Automatically after 60 seconds.
D.
(1) 1KXAA Alternate Source Supplying Load light is LIT.
(2) Manually.
Page 51 of 100
2010 Catawba Nuclear Station
- NRC Initial License Examination 52.
Given the following conditions:
DIG 1A has been started and fully loaded for a Periodic Test.
The control room operators note the following alarm annunciates:
1AD-12, D/1, DIESEL GEN HXAOUTLETFLOW-LO Lube oil temperature has risen rapidly and is now 207°F.
Jacket water outlet temperature is 185°F and rising.
DIG IA continues to run and is fully loaded.
Per the Annunciator Response Procedure, which ONE of the following actions will be taken FIRST to address the above conditions?
A.
Start an additional RN pump to increase DIG heat exchanger cooling water flow.
B.
Dispatch an operator to ensure that I RN-232A (DIG 1A Hx Inlet Isol) is FULLY open.
C.
Reduce loading on DIG 1A.
D.
Manually trip the D/G 1A.
Page 52 of 100
2010 Catawba Nuclear Station
- NRC Initial License Examination 53.
Consider the following Mode 1 plant conditions involving the following radiation monitors:
1 EMF-38, Containment Radiation Monitor.
I EMF-46A, KC (Component Cooling) System Train A Radiation Monitor.
IEMF-71, Main Steam Line Radiation Monitor.
Which ONE of the following identifies conditions where both Condition 1 and 2 will result in an entry into a Technical Specification LCO or a Selected Licensee Commitment (SLC)?
Condition I Condition 2 (IEMF-46A is NOT functional, and KC Train is in service.)
A.
B Plant at 50% power and 1 EMF-71 is NOT functional.
B.
A Plant at 50% power and 1 EMF-71 is NOT functional.
C.
B 1 EMF-38 sample lined up to INCORE ROOM and LOWER ONLY.
D.
A 1 EMF-38 sample lined up to INCORE ROOM and LOWER ONLY.
Page 53 of 100
2010 Catawba Nuclear Station - NRC Initial License Examination 54 Given the following Unit 1 conditions:
The Unit is at 100% power normal operations.
ONE train of KC is in service.
KC HX lB OTLT MODE control switch is configured as indicated in the photo below.
Which ONE of the following describes:
(1) the expected position of the KG HX IA OTLT MODE control switch; AND (2) if 1 2OVAC Auxiliary Power Panelboard 1 KPA becomes de-energized, KC HX IA Outlet controlling mode will be KC HX IA OTLT MODE Switch position A.
KCTEMP B.
iv1 FLOW r_.
U.
vhH ELiJV KC HX lB OTLT MODE 1 RN-351 Kc HX 18 OTLT T/V Sa RESET
/1 KC HX IA Outlet control mode after toss of IKPA MINi FLOW iviN FLOW LJ Er I
2010 Catawba Nuclear Station
- NRC Initial License Examination 55.
Given the following Unit 1 conditions:
The Unit was initially at 100% power.
A LOCA has occurred.
Containment pressure is 8 psig and decreasing.
Safety Injection signal has NOT been reset.
Phase A and Phase B have NOT been reset.
Which ONE of the following describes the response if the operator pushes the Phase A and the Phase B RESET buttons on control panel 1 MC-1 1 for the above conditions?
A.
- 1) Phase A will reset.
- 2) Phase B will NOT reset.
B.
- 1) Phase A will reset.
- 2) Phase B will reset.
C.
- 1) Phase A will NOT reset.
- 2) Phase B will NOT reset.
D.
- 1) Phase A will NOT reset.
- 2) Phase B will reset.
Page 55 of 100
2010 Catawba Nuclear Station
- NRC Initial License Examination 56.
Given the following Unit I conditions:
Plant conditions have been stable for the past 15 minutes.
Loop 1A Tave is 572.0 °F.
Loop 1 B Tave is 570.0 °F.
Loop 1C Tave is 570.0 °F.
Loop 1 D Tave is 568.0 °F.
Tref is 568.0° F.
The CRD Bank Select Switch is in MANUAL.
If the CRD Bank Select Switch is placed in AUTO, the control rods will initially A.
NOT step.
B.
Step IN at 8 steps/minute.
C.
Step IN at 40 steps/minute.
D.
Step IN at 54 steps/minute.
Page 56 of 100
2010 Catawba Nuclear Station
- NRC Initial License Examination 57.
Given the following Unit 1 conditions:
The Unit is at 100% power.
Channel Ill Pressurizer Level has been removed from service for calibration.
Channel I Pressurizer Level begins to malfunction and is drifting HIGH.
1AD-2, F/9 DCS ALTERNATE ACTION has alarmed.
Indicated Pressurizer Level Control Channel values are as follows:
Channel I
60%
Channel II 55%
Channel III 0%
A leak on the NC system develops.
The crew has determined the leakrate to be approximately 130 gpm.
Which ONE of the following describes the effect of these conditions as Pressurizer level continues to lower?
A.
Pressurizer heaters automatically trip.
Letdown automatically isolates.
B.
Pressurizer heaters automatically trip.
Operators will have to manually isolate letdown.
C.
Operators will have to manually trip all Pressurizer heaters.
Operators will have to manually isolate letdown.
D.
Operators will have to manually trip all Pressurizer heaters.
Letdown automatically isolates.
Page 57 of 100
2010 Catawba Nuclear Station
- NRC nitiaI License Examination 58.
Which ONE of the following completes the statement below to describe how a severe axial flux imbalance that is outside of the normal limits, as defined in the ROD Book section 3.9 (OAC Manual Input Data), affects automatic and/or manual rod withdrawal at 100% power?
Axial Flux Distribution (AFD) inputs to OPDT and OTDT cause both setpoints to which could actuate a C3 or C4 rod stop which will prevent A.
decrease; automatic AND manual rod withdrawal B.
increase; automatic AND manual rod withdrawal C.
decrease; automatic rod withdrawal ONLY D.
increase; automatic rod withdrawal ONLY Page 58 of 100
2010 Catawba Nuclear Station
- NRC Initial License Examination 59.
A Reactor Trip and Safety Injection have occurred on Unit 1 due to a Loss of Coolant Accident (LOCA).
The following conditions exist:
Containment pressure is 10 psig and slowly decreasing.
All NC pumps have been secured.
NC system subcooling is -50°F.
CETs indicate 750°F and increasing.
Reactor Vessel Lower Range Level (RVLIS) is currently 34% and slowly decreasing.
Which ONE of the following completes the statement below?
The status of Core Cooling is currently A.
ORANGE, and will remain ORANGE even if RVLIS increased by 10%.
B.
ORANGE, but will be YELLOW [1 CETs decreased by 100°F.
C.
RED, and will remain RED even if CETs decreased by 100°F.
D.
RED, but will be ORANGE H RVLIS stabilized at 10% higher.
Page 59 of 100
2010 Catawba Nuclear Station
- NRC Initial License Examination 60.
Unit 1 was operating at 100% power when the following sequence of events occurred:
A Loss of Offsite Power occurred.
The reactor tripped.
When 1A DIG attempted to load 1 ETA, 87G (Generator Differential) relay actuated due to a fault on the bus.
Just after the reactor trip, a LOCA inside containment developed.
Containment pressure has risen to 3.1 psig and is slowly increasing.
Which ONE of the following describes the status of the VE (Annulus Ventilation) fans?
A.
ONLY 1 B VE fan is running.
B.
1A AND I B VE fans are running.
C.
ONLY 1 B VE fan will start after a 9 minute time delay.
D.
1A AND I B VE fans will start after a 9 minute time delay.
Page 60 of 100
2010 Catawba Nuclear Station
- NRC Initial License Examination 61.
Given the following Unit 1 conditions:
The Unit is in Mode 5.
Containment Purge is in service.
Which ONE of the following identifies alarm conditions that will automatically isolate Containment Purge?
A.
I EMF-1 1, Reactor Bldg. Incore Inst Rm Monitor.
1 EMF-36, Unit Vent Gas Hi Rad Monitor.
B.
I EMF-36, Unit Vent Gas Hi Rad Monitor.
1 EMF-38, Containment Particulate Radiation Monitor.
C.
1 EMF-38, Containment Particulate Radiation Monitor.
1 EMF-39, Containment Gas Hi Rad Monitor.
D.
1 EMF-1 1, Reactor Bldg. Incore Inst Rm Monitor.
I EMF-39, Containment Gas Hi Rad Monitor.
Page 61 of 100
2010 Catawba Nuclear Station
- NRC Initial License Examination 62 Given the following Unit 1 conditions:
The reactor is stable at approximately 10-8 amps.
A Steam Generator safety valve fails open.
NO Reactor Trip or Safety Injection actuation occurs.
Which ONE of the following describes the difference in the plant response if this event were to occur at the beginning-of-life (BOL) as compared to the end-of-life (EOL)?
The NC system will stabilize at a lower temperature at A.
BOL than at EOL.
Reactor power will stabilize at a higher value at BOL than at EOL.
B.
BOL than at EOL.
Reactor power will stabilize at approximately the same value at BOL and EOL.
C.
EOL than at BOL.
Reactor power will stabilize at a higher value at BOL than at EOL.
D.
EOL than at BOL.
Reactor power will stabilize at approximately the same value at BOL and EOL.
Page 62 of 100
2010 Catawba Nuclear Station
- NRC Initial License Examination 63.
Given the following Unit 1 conditions:
The Unit is at 100% power.
Generator Breaker IA then spuriously opens.
The ONLY annunciator associated with the Turbine or Generator that alarms is:
lAD-li, F/i, GEN BKR B OVERCURRENT Which ONE of the following describes:
(1) another annunciator which SHOULD have alarmed; AND (2) a manual action required for mitigating these conditions?
A.
(1) lAD-i, F14, TURB RUNBACK INITIATED (2)
Manually initiate a turbine runback to 48%.
B.
(1) lAD-i, F14, TURB RUNBACK INITIATED (2)
If condition has not cleared within three minutes, manually trip Generator Breaker 1 B.
C.
(1) lAD-il, F17, GEN BKR B TROUBLE (2)
Manually initiate a turbine runback to 48%.
D.
(1) lAD-il, F/7, GEN BKR B TROUBLE (2)
If condition has not cleared within three minutes, manually trip Generator Breaker 1 B.
Page 63 of 100
2010 Catawba Nuclear Station
- NRC Initial License Examination 64.
Given the following conditions:
All liquid waste system controls are in normal configuration.
A large amount of highly radioactive water begins entering the CA pump room sump.
The room sump pump automatically starts and the sump level begins to decrease slowly.
The following annunciator is then received:
1RAD-2, B13, IEMF 52 CLEAN AREA FLOOR DRN HI RAD.
Due to the above conditions, the level in the Turbine Building Sump will A.
NOT increase, because the CA Pump Room Sump Pumps will trip on high radiation.
B.
INCREASE to the sump pumps automatic start setpoint, the sump pumps start, and then they all trip on high radiation.
C.
NOT increase, because the CA Pump Room Sump Pumps discharge is automatically rerouted to the ND/NS sump due to the high radiation.
D.
INCREASE, but the sump pumps will start ONLY if the Emergency High level is reached.
Page 64 of 100
2010 Catawba Nuclear Station
- NRC Initial License Examination 65.
An air leak has developed on the VI System.
As VI pressure decreases, the setpoint at which 1VI-500, VI Supply to VS closes is _(1)_,
and the setpoint at which the VS BASE air compressor starts is (2)_.
A.
(1) 85psig.
(2) 90 psig.
B.
(1) 80psig (2) 76 psig C.
(1) 85 psig (2) 80 psig.
D.
(1) 80 psig.
(2) 90 psig.
Page 65 of 100
2010 Catawba Nuclear Station
- NRC Initial License Examination 66.
Which ONE of the following describes a required duty of the Reactor Operator stationed in the Control Room during fuel handling operations, per NSD 414, (Fuel Handling)?
A.
Monitor Source Range instrumentation while a fuel assembly is being placed in the core.
B.
During actual core alterations, conduct PT11N45501001 C (Refueling Communications Test) once each hour.
C.
Monitor the refueling communications circuit to track correct placement of fuel assemblies.
D. Ensure the correct fuel assembly is being moved to the reactor building by monitoring SFP upender camera.
Page 66 of 100
2010 Catawba Nuclear Station
- NRC Initial License Examination 67.
Given the following conditions:
Unit 1 is at 50% power.
Unit 2 is in a refueling outage, with fuel moves in progress from the Reactor cavity to the Spent Fuel Pool.
For the above conditions, and per Selected Licensee Commitment 16.13-4, (Minimum Station Staffing Requirements), which ONE of the following describes the MINIMUM required number of ROs to be present at all times in the area marked NORMAL (total for both units);
AND the total MINIMUM required number of ROs on shift?
Reference Provided Total Minimum ROs in Area Marked NORMAL Total Minimum Required ROs on shift A.
2 3
B.
2 4
C.
3 3
D.
3 4
Page 67 of 100
2010 Catawba Nuclear Station - NRC Initial License Examination 68.
The DC Power On light for CAPT Elect Ops Status, ISA-145 on 1MC-1O is OUT.
Which ONE of the following describes the process for troubleshooting this condition?
A.
Follow the guidance in lAD-il, L/2 (BIO ALTERNATE CONT PWR LOSS) to dispatch an NEO to 1TB0X0398 (SB594, U-31) to troubleshoot the condition.
B.
Replace the light bulb.
If it is still not lit, this means FUI or FU2 is blown. Write a Work Request to have the fuse replaced.
C.
Follow the guidance in OMP 2-39, (Changing Control Indicating Lights).
It provides a specific troubleshooting sequence for this particular component.
D.
Use OP/01A163501014, (Operations Troubleshooting Guidelines), to determine the cause of this condition. Then write the appropriate Work Request for repair.
Page 68 of 100
2010 Catawba Nuclear Station
- NRC Initial License Examination 69.
During a pre-job brief it is identified that several steps in an operating procedure need to be performed in a different sequence from what is written in the procedure.
Which individual must approve this change, per NSD-704, (Technical Procedure Use and Adherence)?
A.
Operations Shift Manager (OSM)
B.
Shift Operations Manager (SCM)
C.
Work Control Center (WCC) Supervisor D.
Unit Supervisor (US)
Page 69 of 100
2010 Catawba Nuclear Station
- NRC Initial License Examination 70.
Given the following conditions:
A planned Liquid Waste Release of Waste Monitor Tank (WMT) A was initiated at 1110.
The following timeline of events then occurs:
1120- 1 RAD-1, C/5, EMF-49 LIQUID WASTE DISCH HI RAD alarms.
1130
- The release is manually re-initiated without re-sampling.
1145-1 RAD-1, C15, EMF-49 LIQUID WASTE DISCH HI RAD alarms.
1155
- The release is manually re-initiated without re-sampling.
1215-IRAD-1, C/5, EMF-49 LIQUID WASTE DISCH HI RAD alarms.
Based on the above conditions:
(1) can the release be manually re-initiated, without re-sampling, per OP/0/B/6500/1 13, (Operations Liquid Waste Release);
AND (2) which valve has been automatically closed to terminate the release?
A.
(1)
NO (2) 1WL-X28 B.
(1)
NO (2) 1WL-124 C.
(1)
YES (2) 1 WL-X28 D.
(1)
YES (2)
IWL-124 Page 70 of 100
2010 Catawba Nuclear Station - NRC Initial License Examination 71.
Prior to a refueling outage an operator has received 1500 mR so far for that year. During the refueling outage, the operators first assigned task results in receiving a dose of 230 mR.
The NEXT time the operator logs on to the EDC computer, what type of notification flag will the operator see next to their name; AND per NSD 507,( Radiation Protection), what will be required prior to RCA entry?
Type of Notification Flag Required Prior to RCA Entry A.
Alert Notify their supervisor.
B.
Alert Obtain a dose extension approval.
C.
Exclude Notify their supervisor.
D.
Exclude Obtain a dose extension approval.
Page 71 of 100
2010 Catawba Nuclear Station
- NRC Initial License Examination 72 Given the following conditions:
An operator has logged on to RWP #23, Entry for Routine Plant and Systems Operation (Operations), Task #4, Entry for Troubleshooting/Special Inspections.
The operator needs to reach across a Contaminated Area boundary to access a component located in Unit I ND Pump Room.
RP has determined that donning a single rubber glove will be adequate to prevent a personnel contamination.
The operator then proceeds as follows:
1.
Contacts RP and obtains verbal authorization to reach across the Contaminated Area boundary.
2.
Dons the rubber glove and accesses the component across the Contaminated Area boundary.
3.
Removes rubber glove and places it in the contaminated clothing container.
4.
Exits the ND Pump Room and immediately proceeds to SPA (Single Point of Access).
5.
Notifies RP of the results of completion of the task.
6.
Exits the RCA and reports to the Control Room.
Which ONE of the following is the correct evaluation of the above conditions, per NSD 507, (Radiation Protection)?
A.
All RWP requirements were met.
B.
RP was required to issue a revised RWP before authorizing this operation.
C.
RP was required to take additional smear samples of the component before the operation.
D.
The operator was required to frisk hands and feet upon exiting the ND Pump Room.
Page 72 of 100
2010 Catawba Nuclear Station
- NRC Initial License Examination 73.
Of the four (4) nuclear instruments listed in F-O, (Critical Safety Function Status Trees), for assessing the Subcriticality safety function, which ONE is a Post-Accident Monitoring (PAM) instrument required by LCO 3.3.3, PAM (Post-Accident Monitoring) Instrumentation?
A.
Source Range B.
Intermediate Range C.
Power Range D.
Wide Range Page 73 of 100
2010 Catawba Nuclear Station
- NRC Initial License Examination 74.
Given the following Unit I conditions:
The Unit is at 50% power.
Which ONE of the following indications will result in entry into an emergency procedure?
(Consider each separately.)
A.
EHC pressure indicates 1000 psig.
B.
1NCFT5O1O (NC Flow Loop A Ch. 2) indicates 85%.
C.
Stator inlet pressure indicates 45 psig.
D.
Two (2) SG IC NR levels indicate 84%.
Page 74 of 100
2010 Catawba Nuclear Station
- NRC Initial License Examination 75.
Given the following Unit I conditions:
The Unit is tripped from 100% power due to a loss of Component Cooling Water (KC).
The crew transitions to ES-0.1, (Reactor Trip Response), while continuing to perform the actions of AP/1/A15500/021, (Loss of Component Cooling).
Fifteen minutes after the trip, the following conditions exist:
SG pressures are all approximately 1005 psig and stable.
NC System pressure is 2235 psig and stable.
Thot is approximately 570°F in all loops and slowly decreasing.
Core Exit Thermocouples (CETs) indicate approximately 575°F and stable.
Tcold is approximately 547°F in all loops and stable.
Reference Provided Which ONE of the following describes the status of NC System heat removal for the current plant conditions?
A.
Natural circulation is NOT established. NC System subcooling is inadequate.
B.
Natural circulation is NOT established. CETs are not decreasing.
C.
Natural circulation is established. Heat removal is via the steam dumps.
D.
Natural circulation is established. Heat removal is via the SG PORVs.
Page 75 of 100
2010 Catawba Nuclear Station
- NRC Initial License Examination 76.
Given the following Unit 1 conditions:
Initial With the Unit initially at 100% power, a large break LOCA has occurred.
Containment sump level is currently 3.4 feet and increasing slowly.
ES-I.3, (Transfer to Cold Leg Recirculation) has been completed through Step 11 and proper recirculation flow has been verified.
Subsequent 1A and lB ND Pump amps have begun to oscillate.
For the above conditions, which ONE of the following describes; (1) the procedure that will be implemented, AND (2) the actions contained in that procedure which will mitigate the conditions?
A.
(1)
ECA-1.1, (Loss of Emergency Coolant Recirculation)
(2)
Secure the ND, NV, AND NI pumps, and initiate FWST makeup.
B.
(I)
ECA-I.3, (Containment Sump Blockage)
(2)
Secure the NV and NI pumps ONLY, and perform valve alignments to reduce ND flow.
C.
(I)
ECA-l.l, (Loss of Emergency Coolant Recirculation)
(2)
Secure the NV and NI pumps ONLY, and perform valve alignments to reduce ND flow.
D.
(1)
ECA-1.3, (Containment Sump Blockage)
(2)
Secure the ND, NV, AND NI pumps, and initiate FWST makeup.
Page 76 of 100
2010 Catawba Nuclear Station
- NRC Initial License Examination 77.
Given the following Unit 1 conditions:
Initial As part of a planned shutdown, the Unit has just entered Mode 5.
NC temperature is 190°F and slowly decreasing.
Train 1A of ND is in service.
Subsequent lAD-il, All 4KV ESS PWR TRAIN A TROUBLE, alarms.
Status lights for UV on IA Train 4KV LIT momentarily, AND are now DARK.
NC temperature is now 191°F and slowly increasing.
Which ONE of the following describes the abnormal procedure which will be used to restore ND cooling at the EARLIEST time, and what is the required action directed by that procedure?
A.
APII/N55001007, (Loss of Normal Power);
Restart 1A ND pump.
B.
AP/1/A15500/007, (Loss of Normal Power);
Start 1 B ND pump from the opposite essential train.
C.
AP/1/A15500/019, (Loss of Residual Heat Removal System);
Restart 1A ND pump.
D.
AP/1/Al5500/019, (Loss of Residual Heat Removal System);
Start 1 B ND pump from the opposite essential train.
Page 77 of 100
2010 Catawba Nuclear Station - NRC Initial License Examination 78.
Unit 2 is at 100% power when an NEC reports the breaker for 2KC-56A (KC To ND Hx 2A Sup Isol) looks damaged. The SPOC crew determines that the valve will NOT open.
What is the MINIMUM KC flow required through this valve when aligned for cold leg recirculation; AND for the situation above, what system is required to be declared INOPERABLE?
A.
5000 gpm 2ATrain of KC B.
5000 gpm 2A Train of ND C.
5700 gpm 2A Train of ND D.
5700 gpm 2ATrain of KC Page 78 of 100
2010 Catawba Nuclear Station - NRC Initial License Examination 79.
Given the following Unit 1 conditions:
Initial A loss of offsite power has occurred.
1A DIG did NOT start, due to an 86N relay actuation.
1 B DIG started and all blackout sequencer loads on I ETB have been energized.
Subsequent Three (3) hours later, the following parameters for IA SIG are noted:
CA flow to 1A SIG indicates 0 gpm.
IA SIG Narrow Range levels indicate:
Channel 1:
0% and stable Channel 2:
40% and stable Channel 3:
0% and stable Channel 4:
40% and stable For the above conditions, which ONE of the following describes the Required Action per T.S.
3.3.3 (PAM Instrumentation) for the CA flow; and if the Completion Time for this Required Action is NOT met, what is required?
A.
30 days; Submit a report to the NRC within 14 days.
B.
7 days; Submit a report to the NRC within 14 days.
C.
30 days; Be in Mode 4 in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
D.
7 days Be in Mode 4 in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
Page 79 of 100
2010 Catawba Nuclear Station
- NRC Initial License Examination 80.
Given the following conditions:
Unit 1 is at 100% power.
Unit2isat60%.
At 1059 the Shift Manager receives notification from TCC (Transmission Control Center) that a grid frequency disturbance is developing.
The operating crew begins trending generator frequency as follows:
Time Unit I Unit 2 11:00:00 60.0 Hz 59.9 Hz 1102:00 59.7 Hz 59.9 Hz 11:03:00 58.7 Hz 58.7 Hz 11:04:00 58.6 Hz 58.7 Hz 1:05:00 58.3 Hz 59.6 Hz 11:06:00 58.2 Hz 59.1 Hz 11:07:00 58.8 Hz 58.8 Hz 11:08:00 58.7 Hz 58.8 Hz 11:09:00 58.8 Hz 59.0 Hz 11:10:00 58.7 Hz 59.1 Hz For the above conditions:
1.
What guidance is contained in AP/1/A15500/037, (Generator Voltage and Electric Grid Disturbances) which the SRO will use to address the above conditions?
2.
What components will be operated as part of the mitigative actions, per AP/1/N5500/037?
A.
Separate Unit 1 ONLY from the grid.
Open associated Generator PCBs ONLY.
B.
Separate Unit 1 ONLY from the grid.
Open associated Unit tie PCBs ONLY.
C.
Separate BOTH Units from the grid.
Open associated Generator PCBs ONLY.
D.
Separate BOTH Units from the grid.
Open associated Unit tie PCB5 ONLY.
Page 80 of 100
2010 Catawba Nuclear Station
- NRC Initial License Examination 81.
Given the following Unit 1 conditions:
Initial The Unit was at 100% power.
A feedwater line break inside containment has occurred.
1A and 1 B CA Pumps did NOT auto start.
CAPT #1 tripped on mechanical overspeed.
The operating crew entered FR-H. 1, (Response to Loss of Secondary Heat Sink).
Bleed and feed of the NC System has been initiated.
Step 34 and Step 35 have been completed for closing CA flow control valves, and to continue attempts to establish secondary heat sink.
Subsequently 1A CA Pump has been repaired and has been started, per Step 7, Attempt to establish CA flow to at least one SIG, of FR-H.1.
What procedural guidance will be implemented for restoring CA flow to the 1A S/G; AND what operational concern is addressed by implementing that particular procedural guidance?
A.
OPEN CA flow control valve per Step 7.
Minimize the time that bleed and feed is used for vessel integrity concerns.
B.
OPEN CA flow control valve per Step 7.
Limit the vulnerability of thermal stress damage to only ONE S/G.
C.
OPEN CA flow control valve per Enclosure 8, SIG CA Flow Restoration.
Minimize the time that bleed and feed is used for vessel integrity concerns.
D.
OPEN CA flow control valve per Enclosure 8, S/G CA Flow Restoration.
Limit the vulnerability of thermal stress damage to only ONE S/G.
Page 81 of 100
2010 Catawba Nuclear Station
- NRC Initial License Examination 82.
Given the following Unit I conditions:
A Unit startup is in progress.
Power is currently at I X 10 E-11 amps.
After a high voltage setting adjustment, the indications are as follows:
SR Nis now indicate as follows:
N-31: 3X1OE2CPS N-32: 3X1OE3CPS Which ONE of the following describes; (1) the required action per LCO 3.3.1, Reactor Trip Instrumentation; AND (2) the basis for the required action, per Tech. Spec. Bases 3.3.1?
A.
(1)
Suspend operations involving positive reactivity additions.
(2)
To place the core in a more stable condition.
B.
(1)
Open Reactor Trip Breakers.
(2)
To place the core in a more stable condition.
C.
(1)
Suspend operations involving positive reactivity additions.
(2)
To preclude any power escalation.
D.
(1)
Open Reactor Trip Breakers.
(2)
To preclude any power escalation.
Page 82 of 100
2010 Catawba Nuclear Station
- NRC Initial License Examination 83.
Given the following conditions:
Initial Both Units are at 100% power.
A failure of an isolation valve in the Waste Gas (WG) System has initiated an unintentional release of radioactive gas into the Auxiliary Building.
The pressure in one (1) of the Waste Gas Decay Tanks is decreasing.
EMF-41, Aux. Bldg. Radiation Monitor, receives a Trip 2 signal.
I EMF-36, Unit Vent Radiation Monitor, receives a Trip 2 signal.
Subsequent The leak is then isolated using a manual valve.
Chemistry has calculated that 92,000 curies of noble gases were released before the leak was isolated.
Reference Provided 1.
The effect of manually isolating this leak ensures that radiation exposure to a MEMBER OF THE PUBLIC located at (on the referenced drawing) will be less than 0.5 rem, per SLC 16.11-19, Gas Storage Tanks.
2.
AFTER the leak is isolated, which procedure(s) are used which contain the detailed steps for any required ventilation system realignment(s)?
A.
(1)
Location 1 (2) 0P101N64501003, (Auxiliary Building Ventilation System) ONLY.
B.
(1)
Location 1 (2)
OP/0/A/6450/003, (Auxiliary Building Ventilation System) AND OP/I 1A164501004, (Fuel Pool Ventilation System)
C.
(1)
Location 2 (2)
OP/0/A16450/003, (Auxiliary Building Ventilation System) ONLY.
D.
(1)
Location 2 (2)
OP/0/N6450/003, (Auxiliary Building Ventilation System) AND OP/1/N6450/004, (Fuel Pool Ventilation System)
Page 83 of 100
2010 Catawba Nuclear Station
- NRC Initial License Examination 84.
Given the following Unit 1 conditions:
The Unit was at 50% power.
A LOCA occurred.
Other plant conditions are:
Pressurizer level is 0% and stable.
Containment pressure is 1.3 psig and increasing.
Lower Containment humidity is 70% and increasing.
Subcooling is (-) 10°F and decreasing.
Phase A RESET lights are DARK.
The RO reports the following indications on alarm panel 1 RAD-1:
1 EMF-39 Containment Gas Hi Rad is LIT.
1 EMF-38, 39 Containment Loss of Sample Flow is DARK.
Per, RP/01, (Classification of Emergency), the above conditions are consistent with which Initial Classification (IC);
AND will the release Is Occurring block be checked on Line 6 of the Emergency Notification Form?
Reference Provided A.
4.1.S.3 NO B.
4.1.A.1 NO C.
4.1.S.3 YES D.
4.1.A.1 YES Page 84 of 100
2010 Catawba Nuclear Station
- NRC Initial License Examination 85.
Given the following Unit 1 conditions:
The Unit was initially at 100% power.
The reactor tripped due to a spurious turbine trip.
The crew has performed and exited E-O, (Reactor Trip or Safety Injection).
A Main Steam Isolation signal was received.
lB SIG pressure is being maintained at approximately 1210 psig.
Note:
FR-H.2, (Response to Steam Generator Overpressure)
FR-H.4, (Response to Loss of Normal Steam Release Capabilities)
Based on the above conditions 1.
What is the MAXIMUM number of safety relief valves associated with 1 B S/G that did NOT function as designed?
2.
Which functional recovery procedure will be selected for mitigation?
A.
3 FR-H.2 B.
5 FR-H.2 C.
3 FR-H.4 D.
5 FR-H.4 Page 85 of 100
2010 Catawba Nuclear Station
- NRC Initial License Examination 86.
Given the following Unit 1 conditions:
The Unit is at 100% power.
The operators have determined that due to a controller malfunction, 1KC-132 (Letdn Hx OtIt Temp CtrI) percent (%) output is steadily INCREASING.
All other plant parameters appear normal.
Considering the impact of the above conditions, what procedure will be implemented which contains the correct action for mitigation?
A.
Per OP/1/B/6100/OIOH, 1AD-7, Ff3, LETDN HX OUTLET HI TEMP, take manual control of IKC-132.
B.
Per API1IA/5500/013, (Boron Dilution), bypass the mixed bed demineralizers.
C.
Per OP/1/B/61 00/01 OH, 1AD-7, F/3, LETDN HX OUTLET HI TEMP, bypass the mixed bed demineralizers.
D.
Per AP/1/N5500/013, (Boron Dilution), take manual control of 1KC-132.
Page 86 of 100
2010 Catawba Nuclear Station
- NRC Initial License Examination 87.
Regarding the requirements of Table 3.3.2-1, Engineered Safety Feature Actuation System Instrumentation:
Which ONE of the following bistables is recuired to be placed in BYPASS if it is declared INOPERABLE?
A.
Containment Pressure
- High B.
Containment Pressure
- High High C.
SG Water Level Low Low D.
Nuclear Service Water Suction Transfer - Low Pit Level Page 87 of 100
2010 Catawba Nuclear Station
- NRC Initial License Examination 88.
Given the following Unit I conditions:
At 1000 The Unit was at 100% power.
A Zone A and Zone B lockout occurs.
The CRS has implemented AP111A155001007, (Loss of Normal Power).
At 1005 Zone A and Zone B lockouts have been cleared and RESET.
At 1010 The CRS has directed that preparations for performance of Case Ill actions be initiated.
The crew performs Step 13 of AP/1/A15500/007, and notes that the YV Operable light is NOT lit.
Which ONE of the following describes at time 1010; (1) the status of the YV Isolated and the RN Operable lights, AND (2) what action is required which will maintain containment cooling for these conditions, per AP/1 IA/55001007?
A.
(1)
LIT.
(2)
Ensure at least two (2) RN pumps operating.
B.
(1)
LIT.
(2)
Return YV to normal operation.
C.
(1)
DARK.
(2)
Ensure at least two (2) RN pumps operating.
D.
(1)
DARK.
(2)
Return YV to normal operation.
Page 88 of 100
2010 Catawba Nuclear Station
- NRC Initial License Examination 89.
Given the following Unit 1 conditions:
The Unit is operating at 100% power.
A loss of all off-site power occurs.
The reactor tripped.
The undervoltage status lights for I ETA and 1 ETB are LIT.
IA DIG failed to automatically start.
I B D/G automatically started but the output breaker has NOT closed.
The crew enters E-0, (Reactor Trip or Safety Injection) and begins performing the Immediate Actions.
Based on the conditions above, the SRO is directed to GOTO a different procedure (1) completing afl E-0 Immediate Actions, and will implement a contingency procedure which contains direction to FIRST (2)
A.
(1) priorto (2) initiate automatic load sequencing for IETB by manually initiating SI.
B.
(1) prior to (2) attempt a manual start of IA D/G from the Control Room.
C.
(1) immediately after (2) attempt a manual start of 1A D/G from the Control Room.
D.
(1) immediately after (2) initiate automatic load sequencing for 1ETB by manually initiating SI.
Page 89 of 100
2010 Catawba Nuclear Station
- NRC Initial License Examination 90.
Given the following Unit I conditions:
The Unit is at 100% power.
Fuel shuffles are in progress in the Spent Fuel Pool.
The gas sample pump for 1 EMF-42, Fuel Bldg. Ventilation Monitor, loses power.
Which ONE of the following describes:
(1) per Tech. Specs., whether fuel moves may continue (YES or NO);
AND (2) which procedure will the SRO use for directing an operator on checking the status of the individual power supply breaker for the gas sample pump?
A.
(1)
YES (2)
OP/1/B/6100/O1OY, (Annunciator Response for 1RAD-2), E/3, IEMF 42 FUEL BLDG VENT LOSS OF FLOW B.
(1)
NO (2)
OP/1/B/6100/O1OY, (Annunciator Response for 1RAD-2), E/3, 1EMF 42 FUEL BLDG VENT LOSS OF FLOW C.
(1)
YES (2)
OP/1/A/6350/001, (Normal Power Checklist)
D.
(1)
NO (2)
OP/11N6350/001, (Normal Power Checklist)
Page 90 of 100
2010 Catawba Nuclear Station
- NRC Initial License Examination 91.
With Unit I in Mode 2, how is the requirement to seal closed Containment Purge Exhaust Valve 1VP-IOA administratively satisfied, and what is the Technical Specification basis for the requirement?
A.
Manually close the operating air isolation valve.
A detailed analysis has not been performed to prove it will close during a LOCA in time to ensure offsite dose remains within limits.
B.
Use of a keyswitch which removes power from the solenoid valve for the operating air.
A detailed analysis has not been performed to prove it will close during a LOCA in time to ensure offsite dose remains within limits.
C.
Manually close the operating air isolation valve.
Ensures the measured leakage rate for purge system valves is maintained at 0.01 La when pressurized to Pa, since the valves have a history of leaking after each operation.
D.
Use of a keyswitch which removes power from the solenoid valve for the operating air.
Ensures the measured leakage rate for purge system valves is maintained at < 0.01 La when pressurized to Pa, since the valves have a history of leaking after each operation.
Page 91 of 100
2010 Catawba Nuclear Station - NRC Initial License Examination 92.
Given the following conditions:
0815 A planned release of the contents of Waste Monitor Tank (WMT) B was initiated per OP/01B1650011 13, (Operations Liquid Waste Release).
0820 1 EMF-49, Liquid Waste Discharge Hi Rad, alarms in Trip 2 condition.
IEMF-49 is indicating 2.1E÷05 cpm and has been validated.
IWL-124 (Waste Monitor Tank Pumps Discharge) is OPEN.
1WL-187 (Waste Monitor Tank B Pump Disch to Radiation Monitor) is OPEN.
0830 The Unit Supervisor dispatches SPOC to attempt manual actions to terminate the release.
0915 SPOC reports that the release can be manually terminated but it will take an additional 10-12 minutes to complete the action.
Reference Provided Which ONE of the following describes; (1) the manual action(s) the SRO is required to direct SPOC to perform, as directed by the alarm response procedure; AND (2) what is the classification of this event?
A.
(1) closure of IWL-124 ONLY.
(2)
Unusual Event B.
(1) closure of IWL-124 AND 1WL-187.
(2)
Alert C.
(1) closure of 1WL-124 ONLY.
(2)
Alert D.
(1) closure of 1WL-124 AND IWL-187.
(2)
Unusual Event Page 92 of 100
2010 Catawba Nuclear Station
- NRC Initial License Examination 93.
Given the following Unit 1 conditions:
The Unit is in Mode 2.
A Containment Air Release (VQ) is in progress.
Annunciator 1 RAD-1, A!2, 1 EMF-39 CONTAINMENT GAS HI RAD, goes into alarm (Trip
- 2) due to a valid condition.
Containment Ventilation Isolation Train A and B RESET lights are LIT.
Which ONE of the following describes if the release has automatically terminated (YES or NO);
AND what procedure will the SRO FIRST refer to for mitigating the consequences of these conditions?
A.
YES Go to AP/1/A155001010, (Reactor Coolant Leak), for steps on mitigating the leak in containment.
B.NO Go to 0P111A164501017, (Containment Air Release and Addition System), for steps on manual termination of the release.
C.
YES Go to OP/1/A164501017, (Containment Air Release and Addition System), for determining whether a new Gaseous Waste Release permit is required prior to reinitiating the release.
D.NO Go to AP/1/A15500/O1O, (Reactor Coolant Leak), for steps on mitigating the leak in containment.
Page 93 of 100
2010 Catawba Nuclear Station
- NRC Initial License Examination 94.
A Security Event NOT involving a HOSTILE ACTION is in progress.
Which ONE of the following completes the statement below?
To determine IF a PA announcement to the site is needed, per the appropriate procedure, the Shift Manager is recwired to conduct a discussion with (1)
- and use (2) as guidance to determine what specific information to include in the PA announcement.
(1)
(2)
A.
Security ONLY B.
Security ONLY C.
Security AND Station Manager D.
Security AND Station Manager RPIO2, (Notification of Unusual Event)
NSD 217, (Nuclear Security Program)
RPIO2, (Notification of Unusual Event)
NSD 217, (Nuclear Security Program)
Page 94 of 100
2010 Catawba Nuclear Station
- NRC Initial License Examination 95.
Given the following initial conditions:
Unit 1 experienced a complete loss of the switchyard.
The crew was performing steps in ES-0.2, (Natural Circulation Cooldown).
Station management recommended a rapid cooldown due to secondary inventory concerns.
The crew transitioned to ES-0.3, (Natural Circulation Cooldown) with Steam Void in the Vessel.
Subsequent conditions:
Pressurizer level is 92% and increasing.
Reactor vessel Upper Range (UR) level is 70% and decreasing.
The STA notes a YELLOW path on NC INVENTORY and confers with the OSM regarding whether to transition to FR-l.3, (Response to Voids in Reactor Vessel).
Which ONE of the following identifies:
(1) the action the SRO is required to direct the crew to perform in order to control void growth without interrupting natural circulation; AND (2) the required procedure which directs the action?
A.
(1)
Open reactor vessel head vents.
(2)
FR-l.3 B.
(1)
Open reactor vessel head vents.
(2)
ES-0.3 C.
(1)
Energize pressurizer heaters.
(2)
FR-l.3 D.
(1)
Energize pressurizer heaters.
(2)
ES-0.3 Page 95 of 100
2010 Catawba Nuclear Station
- NRC Initial License Examination 96.
A planned troubleshooting activity will involve the following two items:
ITEM #1.
Placing a jumper on 1A DG Room normal vent fan.
ITEM #2.
Removing power from DG 1A starting air solenoid.
Which ONE of the following identifies the individual who is required to sign for approval of this troubleshooting activity, and why, in accordance with OPIO/A16350/014, (Troubleshooting Guidelines)?
A.
Superintendent of Operations, because of ITEM #1.
B.
Superintendent of Operations, because of ITEM #2.
C.
Engineering Supervisor, because of ITEM #1.
D.
Engineering Supervisor, because of ITEM #2.
Page 96 of 100
2010 Catawba Nuclear Station
- NRC Initial License Examination 97.
Given the following Unit 1 conditions:
Following a refueling outage, the Unit is in Mode 2.
Containment integrity was initially established.
Subsequently, it was determined that the status of the Personnel Air Locks (PAL) is as follows:
Upper Airlock Inner Door Operable Upper Airlock Outer Door Operable Lower Airlock Inner Door Inoperable Lower Airlock Outer Door Operable Repairs required are on the barrel (airlock side of the inner door).
Which ONE of the following identifies; (1) whether the Lower Airlock Outer Door is allowed to be OPENED to make the repair, AND (2) which document provides the justification to support the decision?
A.
(1)
YES (2)
Bases for Tech. Spec. 3.6.2, (Containment Air Locks)
B.
(1)
YES (2)
Site Directive 3.1.2, (Access to Reactor Building and Areas Having High Pressure Steam Relief Devices)
C.
(1)
NO (2)
Bases for Tech. Spec. 3.6.2, (Containment Air Locks)
D.
(1)
NO (2)
Site Directive 3.1.2, (Access to Reactor Building and Areas Having High Pressure Steam Relief Devices)
Page 97 of 100
2010 Catawba Nuclear Station
- NRC Initial License Examination 98.
Given the following conditions:
A LOCA has occurred and a Site Area Emergency has been declared.
The EOF, TSC, and OSC have been activated.
It is necessary to enter 1A NI Pump Room to prevent core damage.
Projected dose rate in the pump room is 1.16E+5 mr/hr.
Duration of the exposure will be 3 minutes.
Which ONE of the following describes the requirements for approving this exposure, as specified in RP-1 8, (Emergency Worker Dose Extension)?
A.
Approval may be obtained from EITHER the Emergency Coordinator the EOF Director.
B.
Requires approval from BOTH the Emergency Coordinator AND the EOF Director.
C.
Approval may be obtained from EITHER the RP Manager OR EOF Director.
D.
Requires approval from BOTH the RP Manager AND EOF Director.
Page 98 of 100
2010 Catawba Nuclear Station
- NRC Initial License Examination 99.
With Unit 2 at 100% power, which ONE of the following identifies:
(1) the event which requires the EARLIEST notification to an offsite agency; AND (2) which agency(ies) is/are notified of this event using the Selective Signaling Telephone?
A.
(1)
Twodropped rods (2)
(1)
Twodropped rods (2) the States and Counties C.
(1) reactor coolant leak exceeds capacity of one charging pump with letdown secured.
(2)
(1) reactor coolant leak exceeds capacity of one charging pump with letdown secured.
(2) the States and Counties Page 99 of 100
2010 Catawba Nuclear Station
- NRC Initial License Examination 100.
Given the following conditions:
A General Emergency has been declared.
The TSC, OSC, and EOF have NOT been activated.
Which ONE of the following completes both statements, in accordance with RPIOO5, (General Emergency)?
(1)
The OSM/Emergency Coordinators responsibility of making Protective Action Recommendations be delegated.
(2)
Turnover of command and control to the TSC or EOF the OSM/Emergency Coordinator of classification, notification and Protective Action Recommendation (PAR) responsibilities.
A.
CAN relieves B.
CAN does NOT relieve C.
CANNOT does NOT relieve D.
CANNOT relieves Page 100 of 100
PROVIDED REFERENCES
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Operations Management Procedure 1-9 Page 1 of 1.1 Area Definition LEGEND:
V//I EMERGENCY NORMAL LXø1 LIMITED
REACTOR TRIP RESPONSE I PAGE NO.
I CNS I
24of30 I
EP/1/AI5000IES-O.1
- Page 1 Of I Revision 32 I
Natural Circulation Monitoring Parameters NOTE: Allow NC temperature to stabilize after transfer of steam dumps to pressure mode.
2.
IF Natural Circulation flow is not established, THEN increase dumping steam to establish Natural Circulation flow.
SIC PRESSURE (PSIG) 1200 1100 1000 900 800 700 600 500 400 300 200 100 0
200 250 300 350 400 450 500 50 NC TEMPERATURE (°E) 600
.1 RP/O/AJ5000Iool Fission Product Barrier Matrix Page 1 of 5 1.
Use EALs to determine Fission Product Barrier status (Intact, Potential Loss, or Loss). Add points for all barriers. Classify according to the table below.
I Note 1: An event (or multiple events) could occur which results in the conclusion that exceeding the Loss or Potential Loss thresholds is IMMINENT (i.e., within 1-3 hours). In this IMMINENT LOSS situation, use judgement and classify as if the thresholds are exceeded.
I Note 2: When determining Fission Product Barrier status, the Fuel Clad Barrier should be considered to be lost or potentially lost if the conditions for the Fuel Clad Barrier loss or potential loss EALs were met previously validated and sustained, even if the conditions do not currently exist.
I Note 3: Critical Safety Function (CSF) indications are not meant to include transient alarm conditions which may appear during the start-up of engineered safeguards equipment. A CSF condition is satisfied when the alarmed state is valid and sustained. The STA should be consulted to affirm that a CSF has been validated prior to the CSF being used as a basis to classify an emergency.
I Example: If ECA-O.O, Loss of All AC Power Procedure, is implemented with an appropriate CSF alarm condition valid and sustained, the CSF should be used as the basis to classify an emergency prior to any function restoration procedure being implemented within the confines of ECA-O.O.
IC Unusual Event IC Alert IC Site Area Emergency IC General Emergency 4.1.U. 1 Potential Loss of 4.1 A. I Loss OR Potential Loss 4.1.S. I Loss OR Potential Loss 4.1.0.1 Loss of All Three Barriers Containment of of Both Nuclear Coolant System Nuclear Coolant System AND Fuel Clad 4.1.U.2 Loss of Containment 4.1.A.2 Loss OR Potential Loss 4.1.S.2 Loss 4.1.G.2 Loss of Any Two Barriers of Fuel Clad AND AND Potential Loss Potential Loss of the Third Combinations of Both Nuclear Coolant System AND Fuel_Clad 4.1.A.3 Potential Loss of 4.1.S.3 Loss of Containment Containment AND AND Loss OR Potential Loss Loss Potential Loss of Any Other Barrier of Any Other Barrier
.1 RP!O/A!s000lool Fission Product Barrier Matrix Page 2 of 5 NOTE:
If a barrier is affected, it has a single point value based on a potential loss or a loss. Not Applicable is included in the table as a place holder only, and has no point value assigned.
Barrier Points (1-5)
Potential Loss (X)
Loss (X)
Total Points Classification Containment 1
3 1
3 Unusual Event NCS 4
5 46 Alert Fuel Clad 4
5 7
10 Site Area Emergency Total Points 11
- 13 General Emergency 1.
Compare plant conditions against the Fission Barrier Matrix on pages 3 through 5 of 5.
2.
Determine the potential loss or loss status for each barrier (Containment, NCS and Fuel Clad) based on the EAL symptom description.
3.
For each barrier, write the highest single point value applicable for the barrier in the Points column and mark the appropriate loss column.
4.
Add the points in the Points column and record the sum as Total Points.
5.
Determine the classification level based on the number of Total Points.
6.
In the table on page 1 of 5, under one of the four classification columns, select the event number (e.g. 4.1.A. 1 for Loss of Nuclear Coolant System) that best fits the loss of barrier descriptions.
7.
Using the number (e.g. 4.1.A. 1), select the preprinted notification form Qj a blank notification form and complete the required information for Emergency Coordinator approval and transmittal.
.1 JP!OIAJsoooIoo1 Fission Product Barrier Matrix Page 3 of 5 4.1.C CONTAINMENT BARRIER 4.LN NCS BARRIER 4.1.F FUEL CLAD BARRIER POTENTIALLOSS-LOSS-POTENTIALLOSS-LOSS-POTENTIAL LOSS-LOSS-(1 Point)
(3 Points)
(4 Points)
(5 Points)
(4 Points)
(5 Points)
- 1. Critical Safety Function Status
- 1. Critical Safety Function Status
- 1. Critical Safety Function Status Containment-RED Not applicable NCS Integrity-Red Not applicable Core Cooling-Core Cooling-Red Orange Core cooling-RED Heat Sink-Red Path is indicated Heat Sink-Red for >15 minutes
- 2. Containment Conditions
- 2. NCS Leak Rate
- 2. Primary Coolant Activity Level Containment Rapid unexplained Unisolable leak GREATER THAN Not applicable Coolant Activity Pressure> 15 PSIG decrease in exceeding the available makeup GREATER THAN containment capacity of one capacity as 300 llCilcc Dose H2 concentration>
pressure following charging pump in indicated by a loss Equivalent Iodine 9%
initial increase the normal ofNCS subcooling.
(DEl) 1-131 charging mode Containment Containment with letdown pressure greater than pressure or sump isolated.
3 psig with less than level response not one full train of NS consistent with and a VX-CARF LOCA conditions.
operating.
NOTE: Refer to Emergency Plan, Sect. D, 4.l.C.2, last paragraph for inability to maintain normal annulus pressure.
CONTINUED CONTINUED CONTINUED
.1 1u10/A!s000/ool Fission Product Barrier Matrix Page 4 of 5 4.l.C CONTAINMENT BARRIER 4.1.N NCS BARRIER 4.l.F FUEL CLAD BARRIER POTENTIAL LOSS LOSS POTENTIAL LOSS LOSS -
POTENTIAL LOSS LOSS -
(1 Point)
(3 Points)
(4 Points)
(5 Points)
(4 Points)
(5 Points)
- 3. Containment Isolation Valves Status After
- 3. SG Tube Rupture
- 3. Containment Radiation Monitoring Containment Isolation Actuation Not applicable Containment Primary-to-Indication that a Not applicable Containment isolation is Secondary leak SG is Ruptured and radiation monitor incomplete and a rate exceeds the has a Non-Isolable 53 A or 53 B direct release path capacity of one secondary line fault Reading at time from containment charging pump in since Shutdown.
exists to the the normal Indication that a environment charging mode SG is ruptured and 0-0.5 hrs> 99 RIhr with letdown a prolonged release 0.5-2 hrs> 43 RIhr
- isolated, of contaminated 2-4 hrs> 31 RJhr secondary coolant 4-8 hrs > 22 R!hr is occurring from
>8 hrs> 13 R!hr the affected SG to the environment
- 4. SG Secondary Side Release With Primary-to-
- 4. Containment Radiation Monitoring
- 4. Emergency Coordinator/EOF Director Secondary Leakage Judgement Not applicable Release of Not applicable Not applicable Any condition, including inability to monitor secondary side to the barrier that in the opinion of the the environment Emergency Coordinator/EOF Director with primary to indicates LOSS or POTENTIAL LOSS of the secondary leakage fuel clad barrier.
GREATER THAN Tech Spec allowable END CONTINUED CONTINUED
.1 Fission Product Barrier Matrix RPIOIAi5000Ioo 1 Page 5 of 5 4.LC CONTAINMENT BARRIER POTENTIAL LOSS
(1 Point)
- 5. Significant Radioactive Inventory In Containment Containment Rad.
Monitor EMF53A or 53B Reading at time since shutdown:
0
- 0.5 hr >390 RJhr 0.5 -2 hr > 170 RJhr 2-4 hr
>l25RJhr 4-8 hr
> 9ORfhr
> 8 hr
> 53 R!hr 4.1.N NCS BARRIER POTENTIAL LOSS (4 Points)
- 5. Emergency Coordinator/EOF Director Judgement Any condition, including inability to monitor the barrier that in the opinion of the Emergency Coordinator /EOF Director indicates LOSS or POTENTIAL LOSS of the NCS barrier.
- 6. Emergency Coordinator /EOF Director Judgment Any condition, including inability to monitor the barrier that in the opinion of the Emergency Coordinator/EOF Director indicates LOSS or POTENTIAL LOSS of the containment barrier.
LOSS -
(3 Points)
LOSS (5 Points)
Not applicable 4.1.F FUEL CLAD BARRIER POTENTIAL LOSS LOSS -
(4 Points)
(5 Points)
END END
.2 RPIO/A!5000/oo 1 UNUSUAL EVENT System Malfunctions ALERT Page 1 of2 SITE AREA EMERGENCY GENERAL EMERGENCY 4.2.U.1 Inability to Reach Required Shutdown Within Technical Specification Limits.
OPERATING MODE:
1,2,3,4 4.2.U.2 Unplanned Loss of Most or All Safety System Annunciation or Indication in the Control Room for Greater Than 15 Minutes.
4.2.A.1 Unplanned Loss of Most or All 4.2.S.1 Safety System Annunciation or Indication in Control Room With Either (1) a Significant Transient in Progress, or (2)
Compensatory Non-Alarming Indicators Unavailable.
4.2.A.1-1 The following conditions exist:
Unplanned loss of most (>50%)
annunciators associated with safety systems for greater than 15 minutes.
Inability to Monitor a Significant Transient in Progress.
exist:
Loss of most (>50%)
Annunciators associated with safety systems.
AND Unplanned loss of most (>50%)
annunciators associated with safety systems for greater than 15 minutes.
In the opinion of the Operations Shift Manager/Emergency Coordinator/EOF Director, the loss of the annunciators or indicators requires additional personnel (beyond normal shift compliment) to safely operate the unit.
AND In the opinion of the Operations Shift Manager/Emergency Coordinator/EOF Director, the loss of the annunciators or indicators requires additional personnel (beyond normal shift compliment) to safely operate the unit.
AND EITHER of the following:
A significant plant transient is in progress Loss of the OAC.
END A significant plant transient is in progress.
AND Loss of the OAC.
AND Inability to provide manual monitoring of any of the following Critical Safety Functions:
subcriticality core cooling heat sink containment.
END 4.2.U.1-1 Plant is not brought to required operating mode within Technical Specifications LCO Action Statement OPERATING MODE:
1, 2, 3, 4 Time.
OPERATING MODE:
1,2,3,4 4.2.S.1-1 The following conditions OPERATING MODE:
1,2, 3, 4 4.2.U.2-1 The following conditions exist:
AND CONTINUED
.2
/O/5oooIoo1 System Malfunctions Page 2 of 2 UNUSUAL EVENT ALERT SITE AREA EMERGENCY GENERAL EMERGENCY 4.2.U.3 Fuel Clad Degradation.
OPERATING MODE:
1,2,3*
4.2.U.3-1 Dose Equivalent 1-13 1 greater than the Technical Specifications allowable limit. (*Mode 3 with TAV >5000 F) 4.2.U.4 Reactor Coolant System (NCS)
Leakage.
OPERATING MODE:
1,2,3,4 4.2.U.4-1 Unidentified leakage 10 gpm.
4.2.U.4-2 Pressure boundary leakage 10 gpm.
4.2.U.4-3 Identified leakage ? 25 gpm 4.2.U.5 Unplanned Loss of All Onsite or Offsite Communications.
OPERATING MODE:
ALL 4.2.U.5-1 Loss of all onsite communications capability (internal phone system, PA system, onsite radio system) affecting the ability to perform routine operations.
4.2.U.5-2 Loss of all offsite communications capability (Selective Signaling, NRC ETS lines, offsite radio system, commercial phone system) affecting the ability to communicate with offsite authorities.
END
.3 1u/O/A/5000/oo 1 UNUSUAL EVENT Abnormal Rad Levels/Radiological Effluent ALERT SITE AREA EMERGENCY Page 1 of5 GENERAL EMERGENCY OPERATING MODE:
ALL 4.3.U.1-1 A valid Trip 2 alarm on radiation monitor EMF-49L or EMF-57 for>
60 minutes or will likely continue for
> 60 minutes which indicates that the release may have exceeded the initiating condition and indicates the need to assess the release with procedure HP/OfB/1 009/014.
4.3.U.1-2 A valid indication on radiation monitor EMF-36L of 3.OOE+04 cpm for> 60 minutes or will likely continue for 60 minutes, which indicates that the release may have exceeded the initiating condition and indicates the need to assess the release with procedure SH/0/B/2005/00 I.
Any Unplanned Release of Gaseous or Liquid Radioactivity to the Environment that Exceeds 200 Times the SLC limits for 15 Minutes or Longer.
4.3.A.1-1 A valid indication on radiation monitor EMF-49L or EMF-57 of l.2E+05 cpmfor 15 minutes or will likely continue for l5 minutes, which indicates that the release may have exceeded the initiating condition and indicates the need to assess the release with procedure HP/0/B/1 009/014.
4.3.S.1 Boundary Dose Resulting from an Actual or Imminent Release of Radioactivity Exceeds 100 mrem TEDE or 500 mrem CDE Adult Thyroid for the Actual or Projected Duration of the Release.
OPERATING MODE:
ALL 4.3.S.1-1 A valid indication on radiation monitor EMF 36L of 2.7E+06 cpm sustained for> 15 minutes.
4.3.S.1-2 Dose assessment team calculations indicate dose consequences greater than 100 rnrem TEDE or 500 rnrem CDE Adult Thyroid at the site boundary.
(Continued) 4.3.G.1 Boundary Dose Resulting from an Actual or Imminent Release of Radioactivity that Exceeds 1000 mrem TEDE or 5000 mrem CDE Adult Thyroid for the Actual or Projected Duration of the Release.
OPERATING MODE:
ALL 4.3.G.1-1 A valid indication on radiation monitor EMF 36H of 8.3E+03 cpm sustained for> 15 minutes.
4.3.G.1-2 Dose assessment team calculations indicate dose consequences greater than 1000 mrem TEDE or 5000 mrem CDE Adult Thyroid at the site boundary.
4.3.U.1 Any Unplanned Release of Gaseous 4.3.A.l or Liquid Radioactivity to the Environment that Exceeds Two Times the SLC Limits for 60 Minutes or Longer.
OPERATING MODE:
ALL (Continued)
(Continued)
(Continued)
.3 RP/O/AJ5000/oo 1 UNUSUAL EVENT Abnormal Rad Levels/Radiological Effluent ALERT SITE AREA EMERGENCY Page 2 of 5 GENERAL EMERGENCY 4.3.U.1-3 Gaseous effluent being released exceeds two times SLC 16.11-6 for>
60 minutes as determined by RP procedure.
4.3.U.1-4 Liquid effluent being released exceeds two times SLC 16.11-1 for>
60 minutes as determined by RP procedure.
Note:
If the monitor reading is sustained for the time period indicated in the EAL j the required assessments (procedure calculations) cannot be completed within this time period, declaration must be made based on the valid radiation monitor reading.
(Continued) 4.3.A.1-2 A valid indication on radiation monitor EMF-36L of 5.4E+05 cpm for 15 minutes or will likely continue for 15 minutes, which indicates that the release may have exceeded the initiating condition and indicates the need to assess the release with procedure SHJOJBI2005/00l.
4.3.A.1-3 Gaseous effluent being released exceeds 200 times the level ofSLC 16.11-6 for> 15 minutes as determined by RP procedure.
4.3.A.1-4 Liquid effluent being released exceeds 200 times the level of SLC 16.11-1 for> 15 minutes as determined by RP procedure.
4.3.S.1-3 Analysis of field survey results or field survey samples indicates dose consequences greater than 100 mrem TEDE or 500 mrem CDE Adult Thyroid at the site boundary.
Note 1:
These EMF readings are calculated based on average annual meteorology, site boundary dose rate, and design unit vent flow rate.
Calculations by the dose assessment team use actual meteorology, release duration, and unit vent flow rate. Therefore, these EMF readings should not be used if dose assessment team calculations are available.
4.3.G.1-3 Analysis of field survey results or field survey samples indicates dose consequences greater than 1000 mrem TEDE or 5000 mrem CDE Adult Thyroid at the site boundary.
Note 1:
These EMF readings are calculated based on average annual meteorology, site boundary dose rate, and design unit vent flow rate.
Calculations by the dose assessment team use actual meteorology, release duration, and unit vent flow rate. Therefore, these EMF readings should not be used if dose assessment team calculations are available.
Note:
If the monitor reading is sustained for the time period indicated in the EAL AND the required assessments (procedure calculations) cannot be completed within this time period, declaration must be made based on the valid radiation monitor reading.
(Continued)
Note 2:
If dose assessment team calculations cannot be completed in 15 minutes, then valid monitor reading should be used for emergency classification.
END Note 2:
If dose assessment team calculations cannot be completed in 15 minutes, then valid monitor reading should be used for emergency classification.
END
.3 UNUSUAL EVENT Abnormal Rad Levels/Radiological Effluent ALERT SITE AREA EMERGENCY PP/O/AI5ooo/oo 1 Page 3 of 5 GENERAL EMERGENCY 4.3.U.2 Unexpected Increase in Plant Radiation or Airborne Concentration.
OPERATING MODE:
ALL 4.3.U.2-1 Indication of uncontrolled water level decrease o greater than 6 inches in the reactor refueling cavity with all irradiated fuel assemblies remaining covered by water.
4.3.U.2-2 Uncontrolled water level decrease of greater than 6 inches in the spent fuel pool and fuel transfer canal with all irradiated fuel assemblies remaining covered by water.
4.3.U.2-3 Unplanned valid area EMF reading exceeds levels as shown in Enclosure 4.10.
END 4.3.A.2 Major Damage to Irradiated Fuel or Loss of Water Level that Has or Will Result in the Uncovering of Irradiated Fuel Outside the Reactor Vessel.
Does not apply to spent fuel in dry cask storage. Refer to EPLAN Section D basis document OPERATING MODE:
ALL 4.3.A.2-l An unplanned valid trip II alarm on any of the following radiation monitors:
Spent Fuel Building Refueling Bridge 1EMF-15 2EMF-4 Spent Fuel Pool Ventilation 1EMF-42 2EMF-42 Reactor Building Refueling Bridge (applies to Mode 6 and No Mode Only) 1EMF-17 2EMF-2 Containment Noble Gas Monitor (Applies to Mode 6 and No Mode Only)
IEMF-39 2EMF-39 (Continued)
.3 IP/O/A/5ooo/oo1 Abnormal Rad Levels/Radiological Effluent Page 4 of 5 UNUSUAL EVENT ALERT SITE AREA EMERGENCY GENERAL EMERGENCY 4.3.A.2-2 Plant personnel report that water level drop in reactor refueling cavity, spent fuel pooi, or fuel transfer canal has or will exceed makeup capacity such that any irradiated fuel will become uncovered.
4.3.A.2-3 NC system wide range level <95% after initiation of NC system make-up.
AND Any irradiated fuel assembly not capable of being lowered into spent fuel pool or reactor vessel.
4.3.A.2-4 Spent Fuel Pool or Fuel Transfer Canal level decrease of>2 feet after initiation of makeup.
AND Any irradiated fuel assembly not capable of being fully lowered into the spent fuel pool racks or transfer canal fuel transfer system basket.
(Continued)
.3
/O/5ooo/ooi Abnormal Rad Levels/Radiological Effluent Page 5 of 5 UNUSUAL EVENT ALERT SITE AREA EMERGENCY GENERAL EMERGENCY 4.3.A.3 Release of Radioactive Material or Increases in Radiation Levels Within the Facility That Impedes Operation of Systems Required to Maintain Safe Operations or to Establish or Maintain Cold Shutdown.
OPERATING MODE:
ALL 4.3.A.3-1 Valid reading on 1EMF-12 greater than 15 mrem/hr in the Control Room.
4.3.A.3-2 Valid indication of radiation levels greater than 15 mrem/hr in the Central Alarm Station (CAS) or Secondary Alarm Station (SAS).
4.3.A.3-3 Valid radiation monitor reading exceeds the levels shown in Enclosure 4.10.
END
.4 1u/O/AI5000/oo 1 Loss of Shutdown Functions Page 1 of3 UNUSUAL EVENT ALERT SITE AREA EMERGENCY GENERAL EMERGENCY 4.4.A.1 Failure of Reactor Protection System Instrumentation to Complete or Initiate an Automatic Reactor Trip Once a Reactor Protection System Setpoint Has Been Exceeded and Manual Trip Was Successful.
4.4.S.1 Failure of Reactor Protection System Instrumentation to Complete or Initiate an Automatic Reactor Trip Once a Reactor Protection System Setpoint Has Been Exceeded and Manual Trip Was NOT Successful.
4.4.G.1 Failure of the Reactor Protection System to Complete an Automatic Trip and Manual Trip Was NOT Successful and There is Indication of an Extreme Challenge to the Ability to Cool the Core.
OPERATING MODE:
1 OPERATING MODE:
1,2,3 4.4.A.1-1 The following conditions exist:
Valid reactor trip signal received or required and automatic reactor trip was not successful.
AND Manual reactor trip from the control room is successful and reactor power is less than 5%
and decreasing.
(Continued)
OPERATING MODE:
1 4.4.S.1-1 The following conditions exist:
Valid reactor trip signal received or required and automatic reactor trip was not successful.
AND Manual reactor trip from the control room was not successful in reducing reactor power to less than 5% and decreasing.
4.4.G.1-1 The following conditions exist:
Valid reactor trip signal received or required and automatic reactor trip was not successful.
AND Manual reactor trip from the control room was not successful in reducing reactor power to less than 5% and decreasing.
AND (Continued)
EITHER of the following conditions exist:
Core Cooling CSF-RED Heat Sink CSF-RED.
END END
.4 RP/O/AI5000/ool Loss of Shutdown Functions Page 2 of 3 UNUSUAL EVENT ALERT SITE AREA EMERGENCY GENERAL EMERGENCY 4.4.A.2 Inability to Maintain Plant 4.4.S.2 Complete Loss of Function in Cold Shutdown.
Needed to Achieve or Maintain hot Shutdown.
OPERATING MODE:
5,6 OPERATING MODE:
1,2,3,4 4.4.A.2-1 Total loss of ND and/or RN and/or KC.
4.4.S.2-1 Subcriticality CSF-RED.
AND 4.4.S.2-2 Heat Sink CSF-RED.
One of the following:
4.4.S.3 Loss of Water Level in the Reactor Vessel That Has or a
Inability to maintain Will Uncover Fuel in the reactor coolant temperature Reactor Vessel.
below 200°F OPERATING MODE:
5,6 Uncontrolled reactor 4.4.S.3-1 Failure of heat sink causes loss coolant temperature rise to of cold shutdown conditions.
>180°F.
AND Lower range Reactor Vessel Level Indication System (RVLIS) decreasing after initiation ofNC system makeup.
4.4.S.3-2 Failure of heat sink causes loss of cold shutdown conditions.
AND Reactor Coolant (NC) system mid or wide range level less than 11% and decreasing after initiation of NC system makeup.
(Continued)
.4 R.P/O/Ai5000/ool Loss of Shutdown Functions Page 3 of 3 UNUSUAL EVENT ALERT SITE AREA EMERGENCY GENERAL EMERGENCY 4.4.S.3-3 Failure of heat sink causes loss of cold shutdown conditions.
AND Either train ultrasonic level indication less than 7.25% and decreasing after initiation of NC system makeup.
END
.5 RP/OIA!5000/oo 1 UNUSUAL EVENT ALERT Loss of Power SITE AREA EMERGENCY Page 1 of2 GENERAL EMERGENCY 4.5.U.1 Loss of All Offsite Power to Essential Busses for Greater Than 15 Minutes.
OPERATING MODE:
1,2,3,4 4.5.U.1-1 The following conditions exist:
Loss of offsite power to essential buses ETA and ETB for greater than 15 minutes.
AND Both emergency diesel generators are supplying power to their respective essential busses.
(Continued) 4.5.A.1 Loss of All Offsite Power and Loss of All Onsite AC Power to Essential Busses During Cold Shutdown Or Refueling Mode.
OPERATING MODE:
5,6, No Mode 4.5.A.1-1 Loss of all offsite and onsite AC power as indicated by:
Loss ofpower on essential buses ETA and ETB.
AND Failure to restore power to at least one essential bus within 15 minutes.
(Continued) 4.5.S.1 Loss of All Offsite Power and Loss of All Onsite AC Power to Essential Busses.
OPERATING MODE:
1,2,3,4 4.5.5.1-1 Loss of all offsite and onsite AC power as indicated by:
Loss ofpower on essential buses ETA and ETB.
AND Failure to restore power to at least one essential bus within 15 minutes.
(Continued) 4.5.G.1 Prolonged Loss of All (Offsite and Onsite) AC Power.
OPERATING MODE:
1,2,3,4 4.5.G.1-1 Prolonged loss of all offsite and onsite AC power as indicated by:
Loss of power on essential buses ETA and ETB for greater than 15 minutes.
AND Standby Shutdown Facility (S SF) fails to supply NC pump seal injection OR CA supply to Steam Generators.
AND (Continued)
.5 UNUSUAL EVENT ALERT Loss of Power SITE AREA EMERGENCY IO/A/sooo/oo 1 Page 2 of 2 GENERAL EMERGENCY OPERATING MODE:
5,6, No Mode 4.5.U.1-2 The following conditions exist:
Loss of offsite power to essential buses ETA and ETB for greater than 15 minutes.
AND One emergency diesel generator is supplying power to its respective essential bus.
4.5.U.2 Unplanned Loss of Required DC Power During Cold Shutdown or Refueling Mode for Greater than 15 Minutes.
OPERATING MODE:
5,6 4.5.U.2-1 The following conditions Unplanned loss of both unit related busses: EBA and EBD both <112 VDC, and EBB and EBC both <109 VDC.
AND Failure to restore power to at least one required DC bus within 15 minutes from the time of loss.
END 4.5.A.2 AC power to essential busses reduced to a single power source for greater than 15 minutes such that an additional single failure could result in station blackout.
OPERATING MODE:
1,2,3,4 4.5.A.2-1 The following condition exists:
AC power capability has been degraded to one essential bus powered from a single power source for> 15 mm. due to the loss of all but one of:
4.5.S.2 Loss of All Vital DC Power.
OPERATING MODE:
1,2,3,4 4.5.S.2-1 The following conditions exist:
Unplanned loss ofboth unit related busses: EBA arid EBD both <112 VDC, and EBB and EBC both <109 VDC.
Failure to restore power to at least one required DC bus within 15 minutes from the time of loss.
At least one of the following conditions exist:
Restoration of at least one essential bus within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is QI likely Indication of continuing degradation of core cooling based on Fission Product Barrier monitoring.
END AND exist:
SATB ATD DIG B SATA ATC DIG A END END
.6 UNUSUAL EVENT Fire/Explosion and Security Events ALERT SITE AREA EMERGENCY Iu/O/AJ5000/oo 1 Page 1 of3 GENERAL EMERGENCY 4.6.U.1 Fire Within Protected Area Boundary ia:
Extinguished Within 15 Minutes of Detection OR Explosion Within the Protected Area Boundary.
OPERATING MODE:
ALL 4.6.U.1-1 Fire in any of the following areas I extinguished within 15 minutes of control room notification or verification of a control room fire alarm.
Reactor Building Auxiliary Building Diesel Generator Rooms Control Room RN Pumphouse SSF CAS SAS Doghouses FWST Turbine Building Service Building Monitor Tank Building ISFSI Unit 1/2 Transformer Yard Areas 4.6.A.1 Fire or Explosion Affecting the Operability of Plant Safety Systems Required to Establish or Maintain Safe Shutdown.
OPERATING MODE:
1, 2,3, 4, 5, 6 4.6.A.1-1 The following conditions exist:
(Non-security events)
Fire or explosion in any of the following areas:
Reactor Building Auxiliary Building Diesel Generator Rooms Control Room RN Pumphouse SSF CAS SAS FWST Doghouses (Applies in Mode 1,2, 3,4 only).
AND One of the following:
Affected safety system parameter indications show degraded performance 4.6.S.1 HOSTILE ACTION within the PROTECTED AREA OPERATING MODE:
ALL 4.6.S.1-1 A HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA as reported by the CNS Security Shift Supervision END 4.6.G.1 HOSTILE ACTION Resulting in Loss of Physical Control of the Facility.
OPERATING MODE:
ALL 4.6.G.1-l A HOSTILE ACTION has occurred such that plant personnel are unable to operate equipment required to maintain safety functions.
4.6.G.1-2 A HOSTILE ACTION has caused failure of Spent Fuel Cooling Systems and IMMINENT fuel damage is likely for a freshly off-loaded reactor core in pool.
END (Continued)
(Continued)
UNUSUAL EVENT.6 Fire/Explosion and Security Events ALERT SITE AREA EMERGENCY i.P/O/A/5ooo/oo 1 Page 2 of 3 GENERAL EMERGENCY 4.6.U.l-2 Report by plant personnel of an unanticipated explosion within protected area boundary resulting in visible damage to permanent structure or equipment or a loaded cask in the ISFSI.
4.6.U.2 Confirmed SECURITY CONDITION or Threat Which Indicates a Potential Degradation in the Level of Safety of the Plant.
OPERATING MODE:
All 4.6.U.2-1 A SECURITY CONDITION that does NOT involve a HOSTILE ACTION as reported by the CNS Security Shift Supervision.
4.6.U.2-2 A credible site-specific security threat notification.
4.6.U.2-3 A validated notification from NRC providing information of an aircraft threat.
END Plant personnel report visible damage to permanent structures or equipment within the specified area required to establish or maintain safe shutdown within the specifications.
Note:
Only one train of a system needs to be affected or damaged in order to satisfy this condition.
4.6.A.2 Fire or Explosion Affecting the Operability of Plant Safety Systems Required to Establish or Maintain Safe Shutdown.
OPERATING MODE:
No Mode 4.6.A.2-1 The following conditions exist:
(Non-security events)
Fire or explosion in any of the following areas:
Spent Fuel Pool Auxiliary Building.
RN Pump house AND One of the following:
Spent Fuel Pool level andlor temperature show degraded performance (Continued)
.6 iu/O/A/5000/ool Fire/Explosion and Security Events Page 3 of 3 UNUSUAL EVENT ALERT SITE AREA EMERGENCY GENERAL EMERGENCY Plant personnel report visible damage to permanent structures or equipment supporting spent fuel pool cooling.
4.6.A.3 HOSTILE ACTION Within the OWNER CONTROLLED AREA or Airborne Attack Threat.
OPERATING MODE:
ALL 4.6.A.3-1 A HOSTILE ACTION is occurring or has occurred within the OWNER CONTROLLED AREA as reported by the CNS Security Shift Supervision 4.6.A.3-2 A validated notification from NRC of airliner attack threat less than 30 minutes away.
END
UNUSUAL EVENT.7 Natural Disasters, Hazards, And Other Conditions Affecting Plant Safety ALERT SITE AREA EMERGENCY RIvO!Ai5000/oo 1 Page 1 of4 GENERAL EMERGENCY 4.7.U.l Natural and Destructive Phenomena Affecting the Protected Area.
OPERATING MODE:
ALL 4.7.U.1-1 Tremor felt and valid alarm on the Syscom Seismic Monitoring System (OAC C 1D2252).
4.7.U.1-2 Report by plant personnel of tornado striking within protected area boundary/ISFSI.
4.7.U.1-3 Vehicle crash into plant structures or systems within protected area boundary/ISFSI.
4.7.U.1-4 Report of turbine failure resulting in casing penetration or damage to turbine or generator seals.
4.7.U.1-5 Independent Spent Fuel Cask tipped over or dropped greater than 24 inches.
4.7.U.1-6 Uncontrolled flooding in the ISFSI area.
4.7.U.1-7 Tornado generated missiles(s) impacting the ISFSI.
4.7.A.1 Natural and Destructive Phenomena Affecting the Plant Vital Area.
OPERATING MODE:
ALL 4.7.A.1-1 Valid OBE Exceeded Alarm on 1AD-4,B/8 4.7.A.1-2 Tornado or high winds:
Tornado striking plant structures within the vital area:
Reactor Building Auxiliary Building FWST Diesel Generator Rooms Control Room RN Pumphouse SSF Doghouses CAS SAS sustained winds 74 mph for
> 15 minutes.
4.7.S.1 Control Room Evacuation Has Been Initiated and Plant Control Cannot Be Established.
OPERATING MODE:
ALL 4.7.S.1-1 The following conditions exist:
Control Room evacuation has been initiated per AP!1(2)/A15500!O 17 AND Control of the plant cannot be established from the ASP or the SSF within 15 minutes.
4.7.S.2 Other Conditions Existing Which in the Judgement of the Emergency Coordinator/EOF Director Warrant Declaration of Site Area Emergency.
OPERATING MODE:
ALL 4.7.S.2-1 Other conditions exist which in the Judgement of the Emergency Coordinator/EOF Director indicate actual or likely major failures of plant functions needed for protection of the public.
END 4.7.G.1 Other Conditions Existing Which in the Judgement of the Emergency Coordinator/EOF Director Warrant Declaration of General Emergency.
OPERATING MODE:
ALL 4.7.G.1-1 Other conditions exist which in the Judgement of the Emergency Coordinator/EOF Director indicate:
(1) actual or imminent substantial core degradation with potential for loss of containment OR (2) potential for uncontrolled radionuclide releases. These releases can reasonably be expected to exceed Environmental Protection Agency Protective Action Guideline levels outside the site boundary.
OR (Continued)
END (Continued)
.7 RPIOIA!5000/oo 1 Natural Disasters, Hazards, And Other Conditions Affecting Plant Safety Page 2 of 4 4.7.U.2 Release of Toxic or Flammable Gases Deemed Detrimental to Safe Operation of the Plant.
OPERATING MODE:
ALL 4.7.U.2-1 Report or detection of toxic or flammable gases thatic enter within the site boundary in amounts that can affect safe operation of the plant.
4.7.U.2-2 Report by Local, County or State Officials for potential evacuation of site personnel based on offsite event.
4.7.U.3 Other Conditions Existing Which in the Judgement of the Emergency Coordinator/EOF Director Warrant Declaration of an Unusual Event.
OPERATING MODE:
ALL 4.7.A.1-3 Visible structural damage caused by ei:
Vehicle crashes OR Turbine failure generated (Continued)
- missiles, OR Other catastrophic events on any of the following plant structures:
Reactor Building Auxiliary Building FWST Diesel Generator Rooms ControlRoom RN Pump House SSF Doghouses CAS SAS 4.7.U.3-l Other conditions exist which in the judgement of the Emergency CoordinatorfEOF Director indicate a potential degradation of the level of safety of the plant.
UNUSUAL EVENT ALERT SITE AREA EMERGENCY GENERAL EMERGENCY END
.7 RP/O/A/5000lool Natural Disasters, Hazards, And Other Conditions Affecting Plant Safety Page 3 of 4 UNUSUAL EVENT ALERT SITE AREA EMERGENCY GENERAL EMERGENCY 4.7.A.2 Release of Toxic or Flammable Gases Within a Facility Structure Which Jeopardizes Operation of Systems Required to Maintain Safe Operations or to Establish or Maintain Cold Shutdown.
OPERATING MODE:
ALL 4.7.A.2-l Report or detection of toxic gases within a Facility Structure in concentrations that will be life threatening to plant personnel.
4.7.A.2-2 Report or detection of flammable gases within a Facility Structure in concentra tions that will affect the safe operation of the plant.
Structures for the above EALs:
Reactor Building Auxiliary Building Diesel Generator Rooms Control Room RN Pumphouse SSF CAS SAS (Continued)
.7 RP/O/AI5000Iool Natural Disasters, Hazards, And Other Conditions Affecting Plant Safety Page 4 of 4 UNUSUAL EVENT ALERT SITE AREA EMERGENCY GENERAL EMERGENCY 4.7.A.3 Control Room Evacuation Has Been Initiated.
OPERATING MODE:
ALL 4.7.A.3-1 Control Room evacuation has been initiated per AP!1(2)/A!5500/O1 7.
4.7.A.4 Other Conditions Existing Which in the Judgement of the Emergency Coordinator/EOF Director Warrant Declaration of an Alert.
OPERATING MODE:
ALL 4.7.A.4-1 Other conditions exist which in the Judgement of the Emergency Coordinator/BOF Director indicate that plant safety systems may be degraded and that increased monitoring of plant functions is warranted.
END
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5, III 11111 11111 11111 11111 II
December 2010 CNS
- NRC Initial License Examination Written Exam KEY 1.
C 50.C 99.D 2.
D 51.
C 100.D 3.
A 52.D 4.
C 53.B 5.
A 54.D 6.
C 55.B 7.
B 56.B 8.
A 57.C 9.
A 58.A 10.D 59.D 11.B 60.A 12.8 61.C 13.D 62.B 14.A 63.A 15.B 64.C 16.A 65.D 17.A 66.A 18.D 67.A 19.C 68.C 20.C 69.A 21.C 70.B 22.A 71.A 23.C 72.D 24.8 73.D 25.D 74.D 26.A 75.C 27.C 76.BeD 28.C 77.A 29.
D 78.
B 30.A 79.A 31.A 80.D 32.A 81.D 33.D 82.C 34.
B
- 83. Bj co
- 35. A 84.
C 36.D 85.C 37.
B 38.A 87.B 39.A 88.A 40.C 89.B
- 90. A 42.C 91.B 43.C 92.C 44.C 93.D 45.A 94.A 46.B 95.D 47.D 96.B 48.D 97.A 49.B 98.A Page lof 1