ML022690308
| ML022690308 | |
| Person / Time | |
|---|---|
| Site: | Harris  |
| Issue date: | 03/22/2002 |
| From: | Ernstes M Division of Reactor Safety II |
| To: | Scarola J Carolina Power & Light Co |
| References | |
| 50-400/02301 50-400/02301 | |
| Download: ML022690308 (143) | |
See also: IR 05000400/2002301
Text
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Submittal
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Reactor Operator Written Exam
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SHEARON HARRIS
EXAM 2002-301
50-4oo
AUGUST 26 - 29, 2002
Harris Nuclear Plant
Common Written Questions
QUESTION: 1 Given the following conditions: "* The Main Turbine is operating at 1800 rpm in preparation for synchronizing to the
grid.
"* Reactor power is being maintained at approximately 12% using the Condenser Steam
Dumps.
- Condenser Vacuum Pump 'A' is under clearance.
- Condenser Vacuum Pump 'B' trips.
Assuming NO operator actions, condenser vacuum degrades until ...
a. the turbine and the reactor trip, and condenser steam dump operation is blocked
b. the turbine trips, and condenser steam dump operation is blocked, but the reactor
remains critical
c. condenser steam dump operation is blocked, but vacuum stabilizes above the
turbine trip setpoint
d. the turbine and reactor trip, but vacuum stabilizes above the steam dumps
interlock setpoint
ANSWER:
a. the turbine and the reactor trip, and condenser steam dump operation is blocked
Harris Nuclear Plant
Common Written Questions
QUESTION: 2 Given the following conditions: A reactor trip and safety injection have occurred. Steam Generator parameters have decreased to the following values:
SG LEVEL PRESSURE
A 32% 870 psig
B 12% 420 psig
C 34% 830 psig
NO operator actions have been taken. Which of the following components is mispositioned?
a. 1FCV-2051B, MDAFW FCV to B SG, CLOSED b. 1FCV-2051C, MDAFW FCV to C SG, OPEN c. 1MS-70, MS B SG to AFW Turbine, CLOSED d. LMS-72, MS C SG to AFW Turbine, OPEN
ANSWER:
d. 1MS-72, MS C SG to AFW Turbine, OPEN
Harris Nuclear Plant
Common Written Questions
QUESTION: 3 If a Containment Ventilation Isolation (CVI) signal occurred, which of the following Containment Ventilation fans would NOT trip directly from the CVI signal, but would trip as a result of being interlocked with other fans?
a. Normal Purge Supply fans (AH-82 A & B) b. Pre-Entry Purge Makeup fans (AH-81 A & B) c. Airborne Radioactivity Removal fans (S-1A & B) d. CNMT Pre-entry Purge Exhaust fans (E-5 A & B) ANSWER: b. Pre-Entry Purge Makeup fans (AH-81 A & B)
Harris Nuclear Plant
Common Written Questions
QUESTION: 4 Hydrogen concentration in the Waste Gas System, downstream of the catalytic recombiners, is limited to 4% to ...
a. maintain levels below flammability limits. b. ensure proper operation of the recombiner. c. limit the volume of waste gas generated. d. minimize the radioactive content of the waste gas decay tanks. ANSWER: a. maintain levels below flammability limits.
Harris Nuclear Plant
Common Written Questions
QUESTION: 5 Given the following conditions:
- A large break LOCA has occurred.
- Containment pressure peaked at 15 psig and has decreased to 6 psig.
- Actions are being taken to place the plant in cold leg recirculation in accordance with
EPP-010, "Transfer to Cold Leg Recirculation."
- Two (2) CSIPs, two (2) RHR Pumps, and two (2) Containment Spray Pumps are
running.
- The crew has just completed alignment of Safety Injection for recirculation and is in
the process of verifying Containment Spray alignment when the Reactor Operator notes Containment Sump level is 25%.
Which of the following actions should be taken?
a. * Stop both trains of Containment Spray
* Maintain both trains of RHR Pumps and CSIPs operating
b. * Stop both trains of Containment Spray
0 Stop one (1) train of RHR Pumps and CSIPs
c. * Stop one (1) train of Containment Spray
* Stop one (1) train of RHR Pumps and CSIPs
d. * Stop both trains of Containment Spray
* Stop both trains of RHR Pumps and CSIPs
ANSWER:
d. * Stop both trains of Containment Spray
* Stop both trains of RHR Pumps and CSIPs
Harris Nuclear Plant
Common Written Questions
QUESTION: 6 Given the following conditions:
- The plant is operating at 93% power.
- Condensate Pump 1B trips on motor overcurrent.
- Condensate Booster Pump 1B trips as a result of the trip of Condensate Pump lB.
Which of the following describes the effect of these events on the Main Feed Pumps AND the required operator action?
a. * Main Feed Pumps 1A and 1B remain running
* Trip the reactor and go to PATH-I
b. * Main Feed Pumps 1A and 1B remain running
* Verify a turbine runback occurs
c. e Main Feed Pump lB trips
* Trip the reactor and go to PATH-1
d. * Main Feed Pump 1B trips
0 Verify a turbine runback occurs
ANSWER:
c. * Main Feed Pump 1B trips
* Trip the reactor and go to PATH-1
Harris Nuclear Plant
Common Written Questions
QUESTION: 7 Given the following conditions:
- The plant is solid in Mode 5 with one (1) RCP in operation.
- RHR Pump A-SA is providing letdown flow with PK-145.l, LTDN PRESSURE
1CS-38, in MAN.
- CSIP A-SA is providing RCS makeup and seal injection.
If instrument air is lost to 1CS-38 (PCV-145), the operator should ...
a. trip CSIP A-SA. b. trip RHR Pump A-SA. c. maintain letdown flow using HC-142.1, RHR Letdown 1CS-28. d. open one PRZ PORV.
ANSWER:
a. trip CSIP A-SA.
Harris Nuclear Plant
Common Written Questions
QUESTION: 8 Given the following conditions: "* An I&C technician reports that both of the Control Room Normal Outside Air Intake
Isolation radiation monitors have failed detectors.
"* It will take somewhere between four (4) and eight (8) hours to replace the detectors. Which of the following states the action which must be taken within one (1) hour, in accordance with Technical Specification 3.3.3.1 ?
a. Establish operation of the Control Room Emergency Filtration System in the
Recirculation Mode of Operation
b. Initiate the preplanned alternate method of radiation monitoring
c. Return the monitors to service, or be in Hot Standby within the next six (6) hours
d. Perform a surveillance test on the Control Room Emergency Filtration System, or
be in Hot Standby within the next six (6) hours
ANSWER:
a. Establish operation of the Control Room Emergency Filtration System in the
Recirculation Mode of Operation
Harris Nuclear Plant
Common Written Questions
QUESTION: 9 Given the following conditions:
- A reactor trip occurred from 75% power approximately 2 hours ago.
- The operating crew is attempting to close the Reactor Trip Breakers.
- All controls and switches are in their normal alignment for plant conditions.
Assuming all other conditions are met for closing the Reactor Trip Breakers, which of the following sets of conditions would physically allow the breakers to close when the REACTOR TRIP BREAKERS TRAINS A&B switch is taken to the CLOSE position?
a. * SG 'A' level is 18%
* IR channel N-36 is failed high
b. * SG 'A' level is 18%
* RCP 'A' is secured
c. * IR channel N-36 is failed high
* PRZ pressure is 1920 psig
d. * PRZ pressure is 1920 psig
* RCP 'A' is secured
ANSWER:
d. * PRZ pressure is 1920 psig
* RCP 'A' is secured
Harris Nuclear Plant
Common Written Questions
QUESTION: 10 The plant is in Mode 1. Prior to entering the Personnel Air Lock, the opposite door must be closed. How is the opposite door checked closed and what would be the consequences of attempting to enter with the opposite door open?
a. * Each door contains a visual indication (red/green light) of the opposite door's
position
* Technical Specifications would be violated
b. o The equalizing valve will NOT open if the opposite door is open
* Technical Specifications would be violated
c. * Each door contains a visual indication (red/green light) of the opposite door's
position
* An interlock will prevent entry if the opposite door is open
d. o The equalizing valve will NOT open if the opposite door is open
& An interlock will prevent entry if the opposite door is open
ANSWER:
c. e Each door contains a visual indication (red/green light) of the opposite door's
position
* An interlock will prevent entry if the opposite door is open
Harris Nuclear Plant
Common Written Questions
QUESTION: 11 Given the following conditions:
- The Containment Spray Actuation System (CSAS) is input by the following four (4)
Containment Pressure transmitters: "* Channel I, PT-950A "* Channel II, PT-951A "* Channel III, PT-952A "* Channel IV, PT-953A
- PT-950A is in TEST for surveillance testing purposes.
- PT-952A is failed low.
- A large break LOCA occurs and actual Containment Pressure reaches 21 psig.
Which of the following describes the response of the Containment Spray system?
a. NEITHER train of Containment Spray will automatically actuate b. ONLY Train 'A' of Containment Spray will automatically actuate c. ONLY Train 'B' of Containment Spray will automatically actuate d. BOTH trains of Containment Spray will automatically actuate ANSWER: d. BOTH trains of Containment Spray will automatically actuate
Harris Nuclear Plant
Common Written Questions
QUESTION: 12 Given the following conditions: "* Several Fuel Handling Building (FHB) area radiation monitors on both trains have
reached the high alarm setpoint.
"* AOP-005, "Radiation Monitoring System," has directed the operator to verify that the
FHB ventilation has shifted to the emergency exhaust lineup.
- Both FHB Emergency Exhaust Fans, E-12 and E-13, are RUNNING.
- FHB Emergency Exhaust Fan Inlets, 1FV-2 SA and 1FV-4 SB, are OPEN
Which of the following additional alignments is expected?
a. * FHB Operating Floor Supply Fans (AH-56, AH-57, AH-58, AH-59)
SECURED
0 FHB Normal Exhaust Isolation Dampers (FL-D4, FL-D5, FL-D21, FL-D22)
OPEN
b. * FHB Operating Floor Supply Fans (AH-56, AH-57, AH-58, AH-59)
RUNNING
* FHB Normal Exhaust Isolation Dampers (FL-D4, FL-D5, FL-D21, FL-D22)
OPEN
c. * FHB Operating Floor Supply Fans (AH-56, AH-57, AH-58, AH-59)
RUNNING
* FHB Normal Exhaust Isolation Dampers (FL-D4, FL-D5, FL-D21, FL-D22)
SHUT
AH-58, AH-59)
d. 0 FHB Operating Floor Supply Fans (AH-56, AH-57,
SECURED
* FHB Normal Exhaust Isolation Dampers (FL-D4, FL-D5, FL-D21, FL-D22)
SHUT
ANSWER:
d. * FHB Operating Floor Supply Fans (AH-56, AH-57, AH-58, AH-59)
SECURED
* FHB Normal Exhaust Isolation Dampers (FL-D4, FL-D5, FL-D21, FL-D22)
SHUT
Harris Nuclear Plant
Common Written Questions
QUESTION: 13 Given the following conditions:
- A loss of offsite power has occurred with the plant at 100% power.
- The operating crew is performing the actions of EOP-EPP-001, "Loss of AC Power to
1A-SA and lB-SB Buses."
- A SGTR has been identified in SG 'C'.
- SGs 'A' and 'B' are being depressurized to 180 psig.
Which of the following describes the method used AND the bases for depressurizing SGs 'A' and 'B' to 180 psig?
a. * Method - Operate the SG PORVs 'A' and 'B' from the MCB
* Bases - Lower RCS pressure below rmptured SG pressure to backfill from SG
'C' to the RCS
b. * Method - Operate the SG PORVs 'A' and 'B' locally
* Bases - Lower RCS pressure below ruptured SG pressure to backfill from SG
'C' to the RCS
c. * Method - Operate the SG PORVs 'A' and 'B' from the MCB
* Bases - Minimize RCP seal damage and RCS inventory loss
d. * Method - Operate the SG PORVs 'A' and 'B' locally
* Bases - Minimize RCP seal damage and RCS inventory loss
ANSWER:
d. * Method - Operate the SG PORVs 'A' and 'B' locally
* Bases - Minimize RCP seal damage and RCS inventory loss
Harris Nuclear Plant
Common Written Questions
QUESTION: 14 Chemistry reports that the RCS Dose Equivalent Iodine (DEI- 131) activity has exceeded the limit and a shutdown is required.
0
The plant is to be placed in Hot Standby with T-avg less than 500 F to ...
a. enhance the ability of the mixed bed demineralizers to remove fission products in
the event of a small break LOCA.
b. minimize the deposition of fission products and activation products on the core
surfaces in the event of a large break LOCA.
c. prevent additional fuel cladding oxidation from occurring in the event of a large
break LOCA.
d. prevent the release of radioactivity to the environment in the event of a SGTR.
ANSWER:
d. prevent the release of radioactivity to the environment in the event of a SGTR.
Harris Nuclear Plant
Common Written Questions
QUESTION: 15 Given the following conditions:
- The plant is operating at 50% power.
- Bank 'D' Control Rods are at 140 steps.
- All control systems are in automatic and at program values.
- The Median Select AT Circuit output has failed high.
Which of the following will occur?
a. ALB-020-2-1, TURBINE AUTOMATIC LOADING STOP, alarms b. ALB-013-8-3, BANK LO-LO INSERTION LIMIT, alarms c. Bank 'D' Control Rods step inward d. Charging flow increases
ANSWER:
b. ALB-013-8-3, BANK LO-LO INSERTION LIMIT, alarms
Harris Nuclear Plant
Common Written Questions
QUESTION: 16 Which one of the following statements describes the reason why some selected 480-V MCC loads have two supply breakers in series?
a. The loads are safety-related, requiring redundant train protection
b. The loads are in Containment, requiring redundant overcurrent protection for the
penetration
c. The loads are safety-related, requiring redundant protection with different
sensitivity
d. The loads are capable of being operated from the ACP, requiring redundant
control functions
ANSWER:
b. The loads are in Containment, requiring redundant overcurrent protection for the
penetration
Harris Nuclear Plant
Common Written Questions
QUESTION: 17 Given the following conditions:
- RCS boron concentration is 1900 ppm.
- Boric Acid Tank concentration is 7100 ppm.
Which of the following RWMU Flow Controller potentiometer settings will result in the HIGHEST ACCEPTABLE total automatic Primary Makeup System flow rate for these conditions?
a. 5.63 b. 6.25 C. 6.88 d. 7.50
ANSWER:
c. 6.88
Harris Nuclear Plant
Common Written Questions
QUESTION: 18 Given the following conditions:
- The site has experienced a loss of offsite power.
- EDG 'A' has started and sequenced all loads.
A valve misalignment has isolated ESW cooling to EDG 'A'.
How long can the EDG operate at full load under these conditions with NO adverse effects?
a. One (1) minute
b. Five (5) minutes
0
c. Until Jacket Water Cooler Outlet temperature exceeds 185 F
0
d. Until Lube Oil Cooler Outlet temperature exceeds 185 F
ANSWER:
a. One (1) minute
Harris Nuclear Plant
Common Written Questions
QUESTION: 19 The plant is operating at 100% power with the following conditions:
Time Ambient Temp CT Basin Temp
1200 35 OF 64 OF
1600 20 OF 60 OF
2000 10 OF 58 OF
Which of the following describes the correct CT Deicing Gate Valve alignment for these conditions?
1600 2000 a. Full Open Full Open b. Full Open Half Open c. Half Open Full Open d. Half Open Half Open
ANSWER:
b. Full Open Half Open
Harris Nuclear Plant
Common Written Questions
QUESTION: 20 Given the following conditions:
- A fire has occurred in cable spread Room A - RAB 286 which requires a plant
shutdown.
- 'A' SG pressure is 1000 psig.
- 'A' SG wide range level is 78%.
- 'A' SG narrow range level is unavailable.
- AFW flow is being supplied to 'A' SG.
Which of the following actions should be taken?
a. Decrease AFW flow to lower 'A' SG wide range level to < 75% b. Decrease AFW flow to lower 'A' SG wide range level to < 57% c. Increase AFW flow to raise 'A' SG wide range level to> 57% d. Increase AFW flow to raise 'A' SG wide range level to > 75%
ANSWER:
a. Decrease AFW flow to lower 'A' SG wide range level to < 75%
Harris Nuclear Plant
Common Written Questions
QUESTION: 21 Given the following conditions:
- The plant is operating at 30% power.
- All control systems are in automatic.
- T-ref fails low.
Which of the following describes the response of the rod control system?
0
a. Rods initially step in at 8 steps per minute to reduce Tavg to 553 F
0
b. Rods initially step in at 8 steps per minute to reduce Tavg to 557 F
0
c. Rods initially step in at 72 steps per minute to reduce Tavg to 553 F
0
d. Rods initially step in at 72 steps per minute to reduce Tavg to 557 F
ANSWER:
0 d. Rods initially step in at 72 steps per minute to reduce Tavg to 557 F
Harris Nuclear Plant
Common Written Questions
QUESTION: 22 While establishing a bubble in the PRZ per GP-002, "Normal Plant Heatup From Cold Solid to Hot Subcritical MODE 5 to MODE 3," letdown pressure control valve 1CS-38 (PK- 145.1), Low Pressure Letdown Pressure Controller, opens progressively. Which of the following describes why PK-145.1 opens?
a. Thermal expansion of liquid in the pressurizer b. Change in CCW heat load c. Spray valves are shut while drawing a bubble d. Switchover of letdown to orifices from RHR-CVCS cross-connect
ANSWER:
a. Thermal expansion of liquid in the pressurizer
Harris Nuclear Plant
Common Written Questions
QUESTION: 23 Given the following conditions:
- A reactor trip has occurred from 28% power.
- RCS Tavg has stabilized at no-load conditions.
Which of the following describes the expected status of the Main Feed Regulating Valves and the Main FW Isolation Valves?
a. o Main Feed Regulating Valves OPEN
* Main FW Isolation Valves OPEN
b. 9 Main Feed Regulating Valves OPEN
* Main FW Isolation Valves CLOSED
c. o Main Feed Regulating Valves CLOSED
* Main FW Isolation Valves OPEN
d. o Main Feed Regulating Valves CLOSED
0 Main FW Isolation Valves CLOSED
ANSWER:
c. * Main Feed Regulating Valves CLOSED
* Main FW Isolation Valves OPEN
Harris Nuclear Plant
Common Written Questions
QUESTION: 24 Given the following conditions:
- The plant is being heated up with RCS temperature at 280'F.
- Containment pressure is indicating (-) 0.8 inches WG.
- 1CB-2 & CB-D51 SA, Vacuum Relief 1CB-2 & CB-D51 SA, is in AUTO.
- 1CB-6 & CB-D52 SB, Vacuum Relief 1CB-6 & CB-D52 SB, is in AUTO.
Assuming NO operator actions, which of the following will automatically occur?
a. lCB-2 & CB-D51 SA will open when Containment pressure decreases to (-) 1.0
inches WG; 1CB-6 & CB-D52 SB will open if Containment pressure continues to
decrease to (-) 2.25 inches WG
b. 1CB-6 & CB-D52 SB will open when Containment pressure decreases to (-) 1.0
inches WG; 1CB-2 & CB-D51 SA will open if Containment pressure continues to
decrease to (-) 2.25 inches WG
c. 1CB-2 & CB-D51 SA and 1CB-6 & CB-D52 SB will both open when
Containment pressure decreases to (-) 1.0 inches WG
d. 1CB-2 & CB-D51 SA and 1CB-6 & CB-D52 SB will both open when
Containment pressure decreases to (-) 2.25 inches WG
ANSWER:
d. 1CB-2 & CB-D51 SA and 1CB-6 & CB-D52 SB will both open when
Containment pressure decreases to (-) 2.25 inches WG
Harris Nuclear Plant
Common Written Questions
QUESTION: 25 A loss of 125 VDC bus DP-1B-SB has just occurred. Which of the following AFW Pumps, if any, are considered inoperable?
a. NO AFW pumps are inoperable b. ONLY MDAFW Pump 1B-SB is inoperable c. ONLY the TDAFW Pump is inoperable d. BOTH MDAFW Pump lB-SB and the TDAFW Pump are inoperable
ANSWER:
d. BOTH MDAFW Pump 1B-SB and the TDAFW Pump are inoperable
Harris Nuclear Plant
Common Written Questions
QUESTION: 26 Given the following conditions:
- The plant is being maintained at 1900 psig.
- RCS temperature is 500TF and stable.
- Excess letdown and normal letdown are both in service.
The following indications are noted:
"* Normal letdown is 67 gpm "* RCP 1A seal injection flow is 9 gpm "* RCP 1B seal injection flow is 7 gpm "* RCP 1C seal injection flow is 8 gpm "* RCP IA seal leakoff flow is 2.5 gpm "* RCP 1B seal leakoff flow is 2.0 gpm "* RCP IC seal leakoff flow is 2.5 gpm
In order to maintain pressurizer level constant, charging flow should be adjusted to indicate...
a. 36 gpm. b. 43 gpm. c. 50 gpm. d. 74 gpm. ANSWER: c. 50 gpm.
Harris Nuclear Plant
Common Written Questions
QUESTION: 27 Which of the following describes the start sequence of the Fire Pumps?
a. The Motor Driven Fire Pump will only start after a 30 second time delay if the
Diesel Driven Fire Pump has received a start signal and is not maintaining Ž 100
psig.
b. The Motor Driven Fire Pump will start at * 93 psig and the Diesel Driven Fire
Pump will start at *< 83 psig.
c. The Diesel Driven Fire Pump will start at * 93 psig and the Motor Driven Fire
Pump will start at * 83 psig.
d. The Diesel Driven Fire Pump will only start after a 30 second time delay if the
Motor Driven Fire Pump has received a start signal and is not maintaining Ž 100
psig.
ANSWER:
b. The Motor Driven Fire Pump will start at * 93 psig and the Diesel Driven Fire
Pump will start at * 83 psig.
Harris Nuclear Plant
Common Written Questions
QUESTION: 28 Given the following conditions:
- An operator is required to complete a valve lineup in an area where the radiation level
is 50 mrem/hour.
- The operator's current annual Total Effective Dose Equivalent (TEDE) is 1450 mrem.
- The operator has NOT received any dose outside CP&L facilities.
What is the MAXIMUM time that the operator may work in this area and not exceed CP&L's Annual Administrative Dose Limit?
a. One (1) hour b. Eleven (11) hours c. Fifty-one (51) hours d. Seventy-one (71) hours
ANSWER:
b. Eleven (11) hours
Harris Nuclear Plant
Common Written Questions
QUESTION: 29 Given the following:
- The unit is at 45% power.
- RCP 'B' trips.
- All SG level controllers are in AUTO.
- NO operator action is taken.
Which of the following describes the response of SG 'B' level?
a. Increases to approximately 70% and stabilizes without any significant decrease in
level during the transient
b. Decreases to approximately 30% and stabilizes without any significant increase in
level during the transient
c. Increases to approximately 70% and then decreases to approximately 30% before
stabilizing
d. Decreases to approximately 30% and then increases to approximately 70% before
stabilizing
ANSWER:
d. Decreases to approximately 30% and then increases to approximately 70% before
stabilizing
Harris Nuclear Plant
Common Written Questions
QUESTION: 30 Given the following conditions:
- Shortly following a loss of offsite power, the following indications are noted on Train
'A' Emergency Safeguards Sequencer (ESS) light box:
CNMT FAN CNMT FAN CNMT FAN CNMT FAN SW BSTR PUMP
HIGH AH-2B LOW AH-2A LOW AH-2B START A
HIGH AH-2A
OFF OFF OFF LIT
LIT
- Prior to AUTO ACT COMPLETE MAN LOAD PERMITTED (Load Block 9)
lighting, a steam break occurs inside Containment, causing a Safety Injection.
Following completion of the sequencer, which of the following indications would be expected on the Train 'A' ESS light box?
CNMT FAN CNMT FAN CNMT FAN CNMT FAN SW BSTR PUMP
HIGH AH-2B LOW AH-2A LOW AH-2B START A
HIGH AH-2A
a. LIT LIT OFF OFF LIT
LIT OFF OFF OFF
b. LIT
OFF LIT OFF LIT
C. OFF
OFF LIT OFF OFF
d. OFF
ANSWER:
c. OFF OFF LIT OFF LIT
Harris Nuclear Plant
Common Written Questions
QUESTION: 31 Given the following conditions:
- A reactor trip and safety injection occurred several minutes ago.
- A loss of offsite power has just occurred.
- Both 6.9 KV buses 1A-SA and lB-SB are being supplied by the diesel generators.
Which of the following components has NO power available?
a. Containment Fan Cooler AH-1 b. Containment Fan Coil Unit AH-37A c. Primary Shield Cooling Fan S-2A d. Reactor Support Cooling Fan S-4A
ANSWER:
b. Containment Fan Coil Unit AH-37A
Harris Nuclear Plant
Common Written Questions
QUESTION: 32 Given the following plant conditions:
- The plant is operating at 100% power.
- 1CS-7, 45 GPM Letdown Orifice A, and 1CS-8, 60 GPM Letdown Orifice B, are
closed.
- 1CS-9, 60 GPM Letdown Orifice C, is open.
- The Reactor Makeup System is setup properly and is in AUTO.
- VCT level transmitter, LT-1 12, fails high.
Assuming NO operator action, which of the following describes the plant response?
a. Charging Pump suction is eventually lost as VCT level decreases
b. 1CS-120 (LCV-1 15A), Letdown VCT/Hold Up Tank, aligns to the VCT and NO
automatic makeup will occur
c. 1CS-120 (LCV-1 15A), Letdown VCT/Hold Up Tank, aligns to the HUT and a
CONTINUOUS makeup to the VCT will occur
d. 1CS-120 (LCV-1 15A), Letdown VCT/Hold Up Tank, aligns to the HUT and
INTERMITTENT makeups at normal setpoints will occur
ANSWER:
d. 1CS-120 (LCV-115A), Letdown VCT/Hold Up Tank, aligns to the HUT and
INTERMITTENT makeups at normal setpoints will occur
Harris Nuclear Plant
Common Written Questions
QUESTION: 33 Given the following conditions:
- CCW Pump 'A' needs to be removed from service for motor replacement.
- CCW Pump 'C' is being aligned to replace CCW Pump 'A'.
Which of the following describes the reason for the Kirk Key interlock associated with CCW Pump 'C'?
a. Prevent aligning CCW Pumps 'A' and 'C' to 6.9 KV Bus 1A-SA simultaneously
b. Prevent aligning CCW Pump 'C' to 6.9 KV Buses 1A-SA and lB-SB
simultaneously
c. Ensure CCW Pump 'C' is racked into the proper 6.9 KV Bus
d. Ensure CCW Pump 'C' breaker does NOT auto close while being racked in
ANSWER:
a. Prevent aligning CCW Pumps 'A' and 'C' to 6.9 KV Bus 1A-SA simultaneously
Harris Nuclear Plant
Common Written Questions
QUESTION: 34 Given the following conditions:
- The plant is currently operating at 30% power.
- RCS boron concentration is 900 ppm.
- Core burnup is 300 EFPD.
- Control Bank 'D' rods are inadvertently withdrawn from 135 steps to 155 steps.
BEFORE RCS temperature increases in response to the rod withdrawal, reactor power will increase from 30% to approximately ...
a. 32%. b. 36%. c. 40%. d. 44%.
ANSWER:
b. 36%.
Harris Nuclear Plant
Common Written Questions
QUESTION: 35 Given the following conditions:
The plant is operating at 68% power.
S Control Bank D, Group 1, step counter indicates 187 steps.
- Control Bank D, Group 2, step counter indicates 187 steps.
- Control Bank D rod heights are as follows:
Group 1 Rod Steam
H2 186
B8 186
H14 192
P8 180
Group 2 Rod Steps
F6 186
F10 198
K10 186
K6 180
Which of the following describes the action, if any, that must be taken within one (1) hour for these conditions?
a. NO actions are required b. Realign rods F6 and K6 within 12 steps of each other c. Reduce power below 50% d. Determine the position of the rods using the movable incore detectors
ANSWER:
a. NO actions are required
Harris Nuclear Plant
Common Written Questions
QUESTION: 36 Given the following conditions:
- The AutoLog is NOT functioning.
- The Reactor Operator is maintaining a manual log.
The following log entries have been made:
- 0956 B-SB CSIP trip
- 1005 Started A-SA CSIP per AOP-018
- 1011 Established normal letdown
At 1030, the Reactor Operator realizes he forgot to make a 0957 entry that letdown had been isolated. Which of the following entries would be a proper entry in accordance with OMM-016, Operator Logs?
a. * 1030 Isolated normal letdown b. L.E. 1030 Isolated normal letdown c. *0957 Isolated normal letdown d. L.E. 0957 Isolated normal letdown
ANSWER:
d. L.E. 0957 Isolated normal letdown
Harris Nuclear Plant
Common Written Questions
QUESTION: 37 Given the following conditions: "* Following a large break LOCA, a transition has been made from EPP PATH-1 to
EPP-010, "Transfer to Cold Leg Recirculation."
"* The operator attempts to open 1RH-25, RHR A to Charging Pump Suction Valve, and
1RH-63, RHR B to Charging Pump Suction Valve.
- 1RH-25 opens, but 1RH-63 fails to open.
Which of the following describes a condition that prevents 1RH-63 from opening AND the actions that should be taken?
a. o 1CS-752, CSIP 'B' Alternate Miniflow, failed to close.
* Maintain RHR Train 'B' aligned for Cold Leg Injection until RWST level
decreases to 3% and then secure RHR Train 'B'.
b. * 1SI-301, CNMT Sump to RHR Pump 'B' Suction, failed to open.
* Maintain RHR Train 'B' aligned for Cold Leg Injection until RWST level
decreases to 3% and then secure RHR Train 'B'.
c. 0 1CS-752, CSIP 'B' Alternate Miniflow, failed to close.
* Close 1CS-753, CSIP 'B' Alternate Miniflow Isolation, and open 1RH-63,
RHR B to Charging Pump Suction Valve.
d. & 1SI-301, CNMT Sump to RHR Pump 'B' Suction, failed to open.
* Open 1SI-3 11, CNMT Sump to RHR Pump 'B' Suction, and open 1RH-63,
RHR B to Charging Pump Suction Valve.
ANSWER:
c. 0 1CS-752, CSIP 'B' Alternate Miniflow, failed to close.
* Close 1CS-753, CSIP 'B' Alternate Miniflow Isolation, and open 1RH-63,
RHR B to Charging Pump Suction Valve.
Harris Nuclear Plant
Common Written Questions
QUESTION: 38 Given the following conditions:
- The plant is operating at 100% power.
- Spent fuel is being moved in Spent Fuel Pool 'B'.
- The suction pipe from Spent Fuel Pool 'B' to the Spent Fuel Pool Cooling Pump
completely severs.
Level in the Spent Fuel Pool will decrease and stabilize at ...
a. 18 feet above the fuel assemblies. Makeup should be initiated using AOP-0 13,
"Fuel Handling Accident."
b. 18 feet above the fuel assemblies. Makeup should be initiated using OP- 116,
"Fuel Pool Cooling System."
c. 21 feet above the fuel assemblies. Makeup should be initiated using AOP-013,
"Fuel Handling Accident."
d. 21 feet above the fuel assemblies. Makeup should be initiated using OP- 116,
"Fuel Pool Cooling System."
ANSWER:
Makeup should be initiated using OP-1 16,
b. 18 feet above the fuel assemblies.
"Fuel Pool Cooling System."
Harris Nuclear Plant
Common Written Questions
QUESTION: 39 Given the following conditions:
- The plant has tripped from 100% power due to a trip of 'B'RCP.
W and C' RCPs are running. 'A'
Which of the following is the expected RVLIS Dynamic Head indication?
a. 36% b. 41% c. 63% d. 100%
ANSWER:
c. 63%
Harris Nuclear Plant
Common Written Questions
QUESTION: 40 Given the following conditions:
- The plant is operating at 40% power.
- AOP-005, "Radiation Monitoring System," has been entered.
- REM-1WC-3544, WPB CCW HX Inlet Monitor, is in HIGH alarm.
As a result of the high alarm, which of the following will automatically close?
a. ICC-252, RCP Thermal Barrier Flow Control Valve b. 3WC-4, WPB CCW Surge Tank Overflow Valve c. 1CC-304, CCW to Gross Failed Fuel Detector d. 3WC-7, WPB CCW Surge Tank Drain Valve
ANSWER:
b. 3WC-4, WPB CCW Surge Tank Overflow Valve
Harris Nuclear Plant
Common Written Questions
QUESTION: 41 The following post-SGTR cooldown procedures all cooldown and depressurize the RCS to RHR conditions:
- EPP-017, "Post SGTR Cooldown Using Backfill"
- EPP-018, "Post SGTR Cooldown Using Blowdown"
- EPP-019, "Post SGTR Cooldown Using Steam Dump"
Which of the following describe how the depressurization and cooldown in EPP-017 differs from that in EPP-018 and EPP-019?
a. * EPP-017 maintains RCS pressure above the ruptured SG pressure
* EPP-018 and EPP-019 maintain RCS pressure the same as the ruptured SG
pressure
b. e EPP-017 maintains RCS pressure below the ruptured SG pressure
* EPP-018 and EPP-019 maintain RCS pressure the same as the ruptured SG
pressure
c. e EPP-017 maintains RCS pressure the same as the ruptured SG pressure
* EPP-018 and EPP-019 maintain RCS pressure above the ruptured SG pressure
d. * EPP-017 maintains RCS pressure the same as the ruptured SG pressure
* EPP-018 and EPP-019 maintain RCS pressure below the ruptured SG pressure
ANSWER:
b. * EPP-017 maintains RCS pressure below the ruptured SG pressure
& EPP-018 and EPP-019 maintain RCS pressure the same as the ruptured SG
pressure
Harris Nuclear Plant
Common Written Questions
QUESTION: 42 Given the following conditions:
- A Control Bank 'D' rod has dropped into the core while operating at 100% power.
- The operating crew has reduced power to 74%.
- Three (3) hours later, they are attempting to withdraw the dropped rod.
In accordance with AOP-001, "Malfunction of Rod Control and Indication System," to maintain programmed Tavg while recovering the dropped rod ...
a. raise turbine load. b. reduce turbine load. c. borate the RCS. d. dilute the RCS.
ANSWER:
a. raise turbine load.
Harris Nuclear Plant
Common Written Questions
QUESTION: 43 The plant is in Mode 1. VCT pressure has decreased to 8 psig. Which of the following is the effect on the plant?
a. VCT water flashes to steam b. Insufficient cooling is available to the No. 2 RCP seals c. Insufficient seal injection is available to the RCPs d. CSIPs begin cavitating due to gas binding
ANSWER:
b. Insufficient cooling is available to the No. 2 RCP seals
Harris Nuclear Plant
Common Written Questions
QUESTION: 44 Given the following conditions: 0 A plant startup is being performed per GP-005, "Power Operation (MODE 2 to
MODE 1)."
- The SG PORVs controllers are set at 87%.
- The Steam Dump Controller has been incorrectly set at 89%.
While preparing to latch the Main Turbine, RCS temperature will be maintained at approximately ...
a. 553 0F. b. 557 0F. c. 562°F. d. 5640F.
ANSWER:
c. 5620 F.
Harris Nuclear Plant
Common Written Questions
QUESTION: 45 Given the following conditions: "* The plant is operating at 100% power when a high radiation condition occurs inside
containment.
"* RC-3561A, Containment Ventilation Isolation radiation monitor (Train A), goes into
high (RED) alarm.
"* RC-3561B, Containment Ventilation Isolation radiation monitor (Train B), is out-of
service for testing.
"* RC-3561C, Containment Ventilation Isolation radiation monitor (Train A), does
NOT respond to the high radiation condition.
"* RC-3561D, Containment Ventilation Isolation radiation monitor (Train B), goes into
high (RED) alarm.
Which train(s) of Containment Ventilation Isolation will actuate, if any?
a. Neither Train 'A' nor 'B' b. Train 'A' only c. Train 'B' only d. Both Train 'A' and 'B'
ANSWER:
d. Both Train 'A' and 'B'
Harris Nuclear Plant
Common Written Questions
QUESTION: 46 Given the following conditions:
- The unit is in Mode 4, performing a cooldown on RHR.
- Both trains of CCW are in service.
- NSW Pump 'A' is operating.
- NSW Pump 'B' is in standby.
- Both ESW Pumps are available, but are NOT running.
- NSW Pump 'A' experiences a sheared shaft.
Which of the following automatically occurs AND what is the effect on the plant cooldown?
a. * ESW aligns on a low flow signal to cool Train 'A' CCW ONLY
* Train 'B' RHR and CCW must be secured.
b. * ESW aligns on a low flow signal to cool BOTH trains of CCW.
* Neither train of RHR and CCW must be secured.
c. * ESW aligns on a low pressure signal to cool Train 'A' CCW ONLY.
* Train 'B' RHR and CCW must be secured.
d. * ESW aligns on a low pressure signal to cool BOTH trains of CCW.
* Neither train of RHR and CCW must be secured.
ANSWER:
d. * ESW aligns on a low pressure signal to cool BOTH trains of CCW.
* Neither train of RHR and CCW must be secured.
Harris Nuclear Plant
Common Written Questions
QUESTION: 47 Which of the following conditions would permit securing Containment Spray per EOP PATH-1 Guide?
a. * Actuation caused by a LOCA
0 Time since LOCA occurred is 6 hours
* Containment pressure is 9 psig
b. * Actuation caused by a LOCA
* Time since LOCA occurred is 3 hours
* Containment pressure is 5 psig
c. * Actuation caused by a Steam Line Break
* Time since Steam Line Break occurred is 3 hours
0 Containment pressure is 5 psig
d. * Actuation caused by a Steam Line Break
* Time since Steam Line Break occurred is 6 hours
* Containment pressure is 9 psig
ANSWER:
c. * Actuation caused by a Steam Line Break
"* Time since Steam Line Break occurred is 3 hours
"* Containment pressure is 5 psig
Harris Nuclear Plant
Common Written Questions
QUESTION: 48 Given the following conditions:
- The plant is in Mode 3 with Tavg at 557F.
- All systems are in their normal alignment.
- Safety Injection is manually actuated inadvertently.
Which of the following describes the impact on Instrument Air inside Containment?
a. * IA-819, Containment Instrument Air, closes
* SI and Phase A must BOTH be reset to allow opening IA-819
b. * IA-819, Containment Instrument Air, closes
* ONLY SI must be reset to allow opening IA-819
c. e IA-819, Containment Instrument Air, closes
* ONLY Phase A must be reset to allow opening IA-819
d. * IA-819, Containment Instrument Air, remains open
* NO actions are required to be taken to restore IA to Containment
ANSWER:
c. * IA-819, Containment Instrument Air, closes
* ONLY Phase A must be reset to allow opening IA-819
Harris Nuclear Plant
Common Written Questions
QUESTION: 49 Given the following conditions:
- The unit is operating at 100% power.
- A turbine trip occurs.
Assuming NO operator actions, which of the following describes the expected FINAL CONDITION of SG pressure and Turbine First Stage Impulse Pressure as compared to the 100% power conditions?
a. SG pressure INCREASES
Turbine First Stage Impulse Pressure INCREASES
b. e SG pressure INCREASES
* Turbine First Stage Impulse Pressure DECREASES
C. S SG pressure DECREASES
Turbine First Stage Impulse Pressure INCREASES
d. e SG pressure DECREASES
* Turbine First Stage Impulse Pressure DECREASES
ANSWER:
b. o SG pressure INCREASES
* Turbine First Stage Impulse Pressure DECREASES
Harris Nuclear Plant
Common Written Questions
QUESTION: 50 Given the following conditions:
- A reactor trip occurred due to a loss of offsite power.
- The plant is being cooled down on RHR per EPP-006, "Natural Circulation
Cooldown with Steam Void in Vessel with RVLIS."
- RCS cold leg temperatures are 190TF.
- Steam generator pressures are 50 psig.
- RVLIS upper range indicates greater than 100%.
- Three CRDM fans have been running during the entire cooldown.
Steam should be dumped from all SGs to ensure ...
a. boron concentration is equalized throughout the RCS prior to taking a sample to
verify cold shutdown boron conditions.
b. all inactive portions of the RCS are below 200'F prior to complete RCS
depressurization.
c. RCS and SG temperatures are equalized prior to any subsequent RCP restart.
d. RCS temperatures do not increase during the required 29 hour vessel soak period.
ANSWER:
b. all inactive portions of the RCS are below 200$F prior to complete RCS
depressurization.
Harris Nuclear Plant
Common Written Questions
QUESTION: 51 Given the following conditions:
8
- During a reactor startup, power has been stabilized at 10- amps.
- Main Feed Pump 'A' is operating and maintaining SG levels at program level.
- Main Feed Pump 'B' is secured.
- Subsequently, SG 'B' level increases to 85%.
Which of the following is the expected status of the following pumps?
a. * Main Feed Pump 'A' RUNNING
* Motor Driven AFW Pumps OFF
* Turbine Driven AFW Pump OFF
b. * Main Feed Pump 'A' OFF
* Motor Driven AFW Pumps RUNNING
* Turbine Driven AFW Pump OFF
c. * Main Feed Pump 'A' OFF
0 Motor Driven AFW Pumps OFF
0 Turbine Driven AFW Pump RUNNING
d. * Main Feed Pump 'A' OFF
* Motor Driven AFW Pumps RUNNING
* Turbine Driven AFW Pump RUNNING
ANSWER:
b. * Main Feed Pump 'A' OFF
"* Motor Driven AFW Pumps RUNNING
"* Turbine Driven AFW Pump OFF
Harris Nuclear Plant
Common Written Questions
QUESTION: 52 Given the following conditions: 0 A loss of offsite power has occurred. 0 Both Emergency Diesel Generators are loaded.
ALB-024-3-2, DIESEL GENERATOR A TROUBLE, alarms.
0 An operator is sent to investigate and reports the following conditions:
S Turbo Oil Press 28 psig and stable S Lube Oil Press 30 psig and stable S Fuel Oil Press 1.5 psig and stable S Day Tank Level 56% and slowly decreasing S Starting Air Pressure 227 psig and slowly decreasing S Jacket Water Pressure 17 psig and stable S Control Air Pressure 53 psig and stable
Which of the following components should have automatically started based on these conditions?
a. Lube Oil Circulating Pump b. Auxiliary Lube Oil Pump c. Fuel Oil Transfer Pump d. Starting Air Compressor ANSWER: b. Auxiliary Lube Oil Pump
Harris Nuclear Plant
Common Written Questions
QUESTION: 53 Given the following conditions:
- PRZ pressure is 1685 psig.
- PRT pressure is 15 psig.
Which of the following indications support a diagnosis that a PRZ PORV is stuck open?
TEMP
DOWNSTREAM
PRZ LEVEL OF PORV
a. Increasing 613 0 F
b. Increasing 250°F
C. Decreasing 613 0 F
d. Decreasing 250 0 F
ANSWER:
b. Increasing 250°F
Harris Nuclear Plant
Common Written Questions
QUESTION: 54 Given the following conditions:
- A Reactor Startup is being performed.
- Initial Source Range Count Rate was 200 count per second (cps).
- 2500 pcm has been inserted into the core by withdrawing control rods and Source
Range Count Rate has increased to 400 cps.
- Rod withdrawal is continued, and an additional 1250 pcm is added to the core.
Which of the following identifies the approximate condition of the core?
a. The reactor is subcritical with a stable count rate of 500 cps b. The reactor is subcritical with a stable count rate of 600 cps c. The reactor is subcritical with a stable count rate of 800 cps d. The reactor is critical with an increasing count rate
ANSWER:
c The reactor is subcritical with a stable count rate of 800 cps
Harris Nuclear Plant
Common Written Questions
QUESTION: 55 During a plant cooldown and depressurization in preparation for a refueling, the SIS Accumulators are depressurized and then drained. The normal drain path for the SIS Accumulators is through the Reactor Coolant Drain Tank ...
a. to the Recycle Holdup Tank. b. to the Waste Holdup Tank. c. via the Spent Fuel Pool Cooling System to the Refueling Water Storage Tank. d. via the Spent Fuel Pool Cooling System to the Transfer Canal.
ANSWER:
a. to the Recycle Holdup Tank.
Harris Nuclear Plant
Common Written Questions
QUESTION: 56 Given the following conditions:
- The plant is in Hot Standby.
- Letdown flow is 105 gpm.
- CSIP 'B' is operating.
- A loss of 125 VDC Emergency Bus DP-1B-SB occurs.
With NO operator actions, which of the following is the response of the plant?
a. Seal injection will be lost b. Charging pump suction will shift to the RWST c. Letdown line flashing will occur d. RCS inventory will be lost
ANSWER:
d. RCS inventory will be lost
Harris Nuclear Plant
Common Written Questions
QUESTION: 57 Which of the following sets of conditions would require that the Reactor Coolant Pumps be secured?
a. o RCS is currently at 525F during a plant heatup
0 Operating CSIP has tripped
0
* CCW Heat Exchanger outlet temperature is 95 F
* ALB-5-1-2B, RCP THERM BAR HDR LOW FLOW, is NOT alarming
b. * RCS is currently at 375 0 F during a plant heatup
* Operating CSIP has tripped
* CCW Heat Exchanger outlet temperature is 112'F
0 ALB-5-1-2B, RCP THERM BAR HDR LOW FLOW, is alarming
c. o RCS is currently at 5250 F during a plant heatup
o CSIP 'A' is operating
* CCW Heat Exchanger outlet temperature is 108'F
0 ALB-5-1-2B, RCP THERM BAR HDR LOW FLOW, is NOT alarming
d. o RCS is currently at 375F during a plant heatup
* CSIP 'A' is operating
* CCW Heat Exchanger outlet temperature is 1220 F
0 ALB-5-1-2B, RCP THERM BAR HDR LOW FLOW, is alarming
ANSWER:
b. e RCS is currently at 375 0 F during a plant heatup
"* Operating CSJP has tripped
"* CCW Heat Exchanger outlet temperature is 112'F
"* ALB-5-1-2B, RCP THERM BAR HDR LOW FLOW, is alarming
Harris Nuclear Plant
Common Written Questions
QUESTION: 58 Given the following conditions:
A loss of offsite power has occurred. SG levels are being maintained constant using AFW in manual control. ERFIS is NOT available. SG pressures are at 885 psig and decreasing slowly. RCS pressure is 1935 psig and stable. Core exit thermocouples are 6240 F and stable. RCS hot leg temperatures are 605TF and stable. RCS cold leg temperatures are 532TF and decreasing slowly.
The operator is verifying natural circulation flow in EPP-004, "Reactor Trip Response." Which of the following describes the status of natural circulation flow per EPP-004?
a. The natural circulation criteria of EPP-004 has been met b. RCS cold leg temperature criteria has NOT been met c. RCS hot leg temperature criteria has NOT been met d. RCS subcooling criteria has NOT been met
ANSWER:
d. RCS subcooling criteria has NOT been met
Harris Nuclear Plant
Common Written Questions
QUESTION: 59 Which of the following would require that Independent Verification be performed in accordance with OPS-NGGC-1303, "Independent Verification?"
a. During Mode 5, a valve in the Containment Spray system is being repositioned for
testing and the OP lineup will be completed prior to Mode 4 entry
b. During Mode 1, a valve in the Main Steam system is being placed under clearance
and the valve is only accessible with a manlift
c. During Mode 4, a valve in CVCS inside containment is being positioned for
0
draining and the valve is located in an area where the temperature is 134 F
d. During Mode 3, a valve in CVCS is being placed under clearance and the valve is
located in a radiation field of 175 mRem/hr with an estimated verification time of
6 minutes
ANSWER:
b. During Mode 1, a valve in the Main Steam system is being placed under clearance
and is only accessible with a manlift
Harris Nuclear Plant
Common Written Questions
QUESTION: 60 Given the following conditions:
- Train 'A' RHR has just been placed in service in accordance with GP-007, "Normal
Plant Cooldown MODE 3 to MODE 5."
- Train 'B' RHR is still aligned for ECCS Mode.
- Interlock P-12 has been bypassed and the Condenser Steam Dumps are in operation.
- Train 'A' equipment is in operation.
- Both CSIPs are still available.
- RCP 'C' has been secured for the cooldown.
A loss of 6.9 KV Bus 1A-SA occurs and EDG 1A-SA fails to start. Which of the following describes the impact of the loss of Bus 1A-SA on the plant?
a. TDAFW Pump becomes inoperable b. RCPs 'A' and 'B' must be secured c. RHR cooling capability is temporarily lost d. Condenser steam dump capability is lost
ANSWER:
c. RHR cooling capability is temporarily lost
Harris Nuclear Plant
Common Written Questions
QUESTION: 61 Given the following conditions:
- FRP-P. 1, "Response to Imminent Pressurized Thermal Shock," is being performed.
- Safety Injection CANNOT be terminated due to inadequate RCS subcooling.
- However, RCS subcooling is adequate to start an RCP.
Which of the following describes the bases for RCP operation under these conditions?
a. Provide additional RCS subcooling b. Provide mixing of injection water and reactor coolant c. Supply additional heat input into the RCS d. Provides normal sprays for the depressurization
ANSWER:
b. Provide mixing of injection water and reactor coolant
Harris Nuclear Plant
Common Written Questions
QUESTION: 62 Given the following conditions:
- REM-3502A, Containment RCS Leak Detection Radiation monitor, is in service.
- REM-3502B, Containment Pre-Entry Purge Radiation monitor, is in service.
Which of the following describes the effect on these monitors if a Containment Isolation Phase 'A' actuation occurs?
a. * REM-3502A remains in service
* REM-3502B remains in service
b. * REM-3502A remains in service
* REM-3502B is isolated
c. * REM-3502A is isolated
* REM-3502B remains in service
d. e REM-3502A is isolated
* REM-3502B is isolated
ANSWER:
c. e REM-3502A is isolated
* REM-3502B remains in service
Harris Nuclear Plant
Common Written Questions
QUESTION: 63 Given the following conditions: "* A LOCA has occurred inside Containment, resulting in a reactor trip and a safety
injection.
"* A transition has just been made from EPP PATH- I to FRP-P. 1, "Response to
Imminent Pressurized Thermal Shock."
- Containment pressure is 7 psig and increasing slowly.
- All RCPs have been secured.
- Pressurizer level is off-scale low.
- RVLIS Full Range indicates 88%.
- Core exit thermocouples are 240SF and decreasing
- RCS cold leg temperatures are 2300 F and decreasing.
- RCS pressure is 285 psig and stable.
- ERFIS indicates subcooling is 177TF.
- RHR HX header flows are both 0 gpm.
* SG levels are as follows:
SG LEVEL
A 32%
B 10%
C 26%
Which of the following actions should be taken in accordance with FRP-P. 1, "Response to Imminent Pressurized Thermal Shock?"
a. Maintain total AFW flow >210 KPPH until at least one (1) SG is >40% level b. Secure AFW flow to all SGs c. Maintain cold leg injection flow, but secure one (1) CSIP d. Return to EOP-PATH-1 ANSWER: a. Maintain total AFW flow > 210 KPPH until at least one (1) SG is > 40% level
Harris Nuclear Plant
Common Written Questions
QUESTION: 64 Given the following conditions:
- A loss of secondary heat sink has occurred.
- Attempts are made to restore main feedwater using FRP-H. 1, "Response to Loss of
Secondary Heat Sink."
- All RCPs are stopped.
- SG level wide range levels are all below 5%.
- Core exit thermocouple temperatures are increasing.
- PRZ pressure is 2180 psig and increasing rapidly.
Which of the following describes the sequence of actions to be taken?
a. * Actuate Safety Injection
* Verify all PRZ PORVs automatically open when pressure increases
b. * Actuate Safety Injection
* Open all PRZ PORVs after verifying Safety Injection flowpath
c. o Open all PRZ PORVs
* Verify Safety Injection automatically actuates when pressure decreases
d. o Open all PRZ PORVs
o Actuate Safety Injection after verifying the PRZ PORVs are open
ANSWER:
b. o Actuate Safety Injection
* Open all PRZ PORVs after verifying Safety Injection flowpath
Harris Nuclear Plant
Common Written Questions
QUESTION: 65 Given the following conditions:
- Reactor power is 8%.
- The turbine is at 1800 rpm, in preparations for synchronizing to the grid.
- A reactor trip occurs.
Which of the following describes why the Main Turbine must be tripped under these conditions?
a. Prevent an uncontrolled RCS cooldown b. Generate an additional reactor trip signal c. Minimize the depletion of SG inventory d. Minimize the pressure increase in the RCS
ANSWER:
a. Prevent an uncontrolled RCS cooldown
Harris Nuclear Plant
Common Written Questions
QUESTION: 66 Given the following conditions:
- PRZ pressure is being controlled in automatic at 2235 psig.
- Pressure transmitter PT-444 fails high.
- Approximately 10 seconds after the failure, the operator places PK-444A in
MANUAL.
Which of the following actions is the operator required to take to restore PRZ pressure to 2235 psig?
a. Raise controller output to cause heaters to energize and spray valves to close b. Raise controller output to cause spray valves to open and heaters to deenergize c. Lower controller output to cause heaters to energize and spray valves to close d. Lower controller output to cause spray valves to open and heaters to deenergize
ANSWER:
c. Lower controller output to cause heaters to energize and spray valves to close
Harris Nuclear Plant
Common Written Questions
QUESTION: 67 The plant is in Mode 3 with the Shutdown Banks withdrawn when the following events occur:
- The reactor trip breakers open.
- ALB-15-1-4,60 KVA UPS TROUBLE, is NOT alarming.
- ALB-15-1-5, 7.5 KVA UPS TROUBLE, is NOT alarming.
- ALB-15-2-2, PIC 1-2-3-4-9-10-13-14 POWER FAILURE, alarms.
- ALB-15-3-2, PIC 5-6-7-8-11-12-15-16 POWER FAILURE, is NOT alarming.
- ALB-15-4-3, PIC 17-18 POWER FAILURE, alarms.
- ALB-15-5-3, PIC 19 POWER FAILURE, is NOT alarming.
- Most lights in the top row of Trip Status Light Boxes are energized.
- Several lights in each of the other rows of Trip Status Light Boxes are energized.
Which of the following buses have been lost?
a. Instrument Bus S-I b. Instrument Bus S-II c. UPS Bus UPP-1A d. UPS Bus UPP-1B
ANSWER:
a. Instrument Bus S-I
Harris Nuclear Plant
Common Written Questions
QUESTION: 68 Given the following conditions: "* The crew diagnosed a SG tube leak. "* REM-1BD-3527, Steam Generator Blowdown, went into high (RED) alarm. "* In response to the alarm on REM-1BD-3527, the crew performed the required actions
of AOP-016, "Excessive Primary Plant Leakage," Attachment 1, "Primary-To
Secondary Leak."
Which of the following describes the expected indicated trend on REM-1BD-3527 after the completion of Attachment 1?
a. Stabilizes and then decreases b. Continues to indicate current SG radiation levels c. Increases to full scale d. Stabilizes and then increases
ANSWER:
a. Stabilizes and then decreases
Harris Nuclear Plant
Common Written Questions
QUESTION: 69 Given the following conditions:
- FRP-C. 1, "Response to Inadequate Core Cooling," is being performed following a
small break LOCA.
- Containment pressure is 8.5 psig.
- Core exit thermocouples are >14000 F.
- All efforts to establish SI flow have failed.
- The crew has started RCP 'C' in an attempt to lower core exit temperatures, but
temperatures have remained above 1300TF.
- SG 'C' level is 55%.
- SGs 'A' and 'B' are off-scale low.
Which of the following actions should be taken?
a. Open the PRZ PORVs and RCS vent valves b. Start RCPs 'A' and 'B' one at a time c. Close any open PRZ PORVs and RCS vent valves d. Refill and repressurize the SI Accumulators for continued injection
ANSWER:
a. Open the PRZ PORVs and RCS vent valves
Harris Nuclear Plant
Common Written Questions
QUESTION: 70 Given the following conditions: . The unit is in the Source Range during a reactor startup.
- Power is lost to Instrument Bus S-rnI.
- A reactor trip occurs.
Which of the following signals will cause a reactor trip?
a. Source Range High Count Rate b. Intermediate Range High Flux c. Power Range Neutron Flux (Low Setpoint) d. Turbine Trip ANSWER: d. Turbine Trip
Harris Nuclear Plant
Common Written Questions
QUESTION: 71 After plant control is completely shifted to the Auxiliary Control Panel in accordance with AOP-004, "Remote Shutdown", which of the following actions will the operators have to manually perform?
a. Align CSIP suction to the RWST b. Transfer control of the EDGs to the local control panels c. Open the reactor trip breakers d. Block SIAS to the Emergency Sequencers
ANSWER:
a. Align CSIP suction to the RWST
Harris Nuclear Plant
Common Written Questions
QUESTION: 72 Given the following conditions: "* During a plant startup, Main Feed Water is aligned to the SGs through the Feed Reg
Valve Bypass FCVs.
"* The controller for FCV-479, SG 'A' Feed Reg Valve Bypass FCV (FK-479.1), has
just been placed in AUTO.
"* The controller for FCV-489, SG 'B' Feed Reg Valve Bypass FCV (FK-489.1), is still
in MANUAL.
"* The controller for FCV-499, SG 'C' Feed Reg Valve Bypass FCV (FK-499.1), is still
in MANUAL.
"* FCV-479 begins going open. Which of the following failures could have caused the response of FCV-479?
a. SG 'A' Feed Flow Channel FT-475 failing low b. SG 'A' Steam Flow Channel FT-476 failing high c. SG 'A' Level Channel LT-476 failing high d. Power Range Channel N-44 failing high
ANSWER:
d. Power Range Channel N-44 failing high
Harris Nuclear Plant
Common Written Questions
QUESTION: 73 Which of the following describes why RCP trip criteria is included in PATH-2?
a. Protect against operator misdiagnosis since RCS pressure should not decrease to
the trip criteria during a SGTR
b. Decrease leakage from the RCS since the total leakage for the duration of the
SGTR is less than it would have been with the RCPs in service
c. Prevent heatup of the RCS since a heatup of the RCS due to the RCPs being in
service increases leakage to the ruptured SG
d. Protect the RCPs from operating with inadequate AP across the number one RCP
seal as a result of the RCS depressurization from the SGTR
ANSWER:
a. Protect against operator misdiagnosis since RCS pressure should not decrease to
the trip criteria during a SGTR
Harris Nuclear Plant
Common Written Questions
QUESTION: 74 Which of the following describes how the Emergency Sequencer is reset following a loss of AC power to 6.9 KV Bus 1A-SA which results in actuation of the Sequencer UV Program?
a. The operator resets the program by turning the SI Reset switch to RESET at least
2.5 minutes after Load Block 9 is completed
b. The operator resets the program by placing both Reactor Trip Breaker A-SA and
Reactor Trip Breaker B-SB to the closed position momentarily after all actuation
signals have been cleared
c. The program automatically resets when Auxiliary Bus D To Emergency Bus A
SA Breaker 104 and Emergency Bus A-SA To Aux Bus D Tie Breaker 105 SA are
closed during the restoration of offsite power
d. The program automatically resets when Diesel Generator A-SA Breaker 106 SA is
opened during the restoration of offsite power
ANSWER:
c. The program automatically resets when Auxiliary Bus D To Emergency Bus A
SA Breaker 104 and Emergency Bus A-SA To Aux Bus D Tie Breaker 105 SA are
closed during the restoration of offsite power
Harris Nuclear Plant
Common Written Questions
QUESTION: 75 Given the following conditions:
- FRP-S. 1, "Response to Nuclear Power Generation / ATWS," is being performed.
- The operating crew is about to exit FRP-S. 1.
Boration should continue even after exiting FRP-S. 1 to ensure ...
a. adequate shutdown margin is established since the criteria for exiting FRP-S. 1 is
only that the reactor be subcritical.
b. the reactor becomes subcritical since the criteria for exiting FRP-S. 1 is only that
the power range channels indicate < 5%.
c. cold shutdown boron concentration is achieved since additional boron, beyond that
needed to make the reactor subcritical, is required to compensate for the cooldown
portion of the recovery.
d. refueling boron concentration is achieved since additional boron, beyond that
needed to make the reactor subcritical, is required to allow for core offloading to
inspect for fuel damage.
ANSWER:
a. adequate shutdown margin is established since the criteria for exiting FRP-S. 1 is
only that the reactor be subcritical.
Harris Nuclear Plant
RO Written Questions
QUESTION: 76 Given the following conditions: "* Spent resin is being sluiced from the Cation Demineralizer to a Spent Resin Storage
Tank.
"* The operator reports that it appears that a pipe in the overhead of a hallway is plugged
with resin.
"* HP reports the results of a radiation survey as follows:
* 2500 mr/hr on contact with pipe * 1200 mr/hr @ 18 inches from the pipe * 5 mr/hr at floor level below the pipe
Which one of the following describes the required radiological postings?
a. NO postings are required because a ladder is required to access the pipe area b. Very High Radiation Area with red flashing light c. High Radiation Area with a red flashing light d. High Radiation Area, but NO red flashing light required
ANSWER:
c. High Radiation Area with a red flashing light
Harris Nuclear Plant
RO Written Questions
QUESTION: 77 Given the following conditions:
- A makeup to the PRT is in progress per OP-100, "Reactor Coolant System."
- Both RC-161, RMW TO CNMT, and RC-167, RMW TO PRT, are open.
Which of the following signals will automatically terminate the PRT makeup AND how will the valves respond?
a. * ONLY a Phase A signal
0 RC-161, RMW TO CNMT, closes
& RC-167, RMW TO PRT, remains open
b. o ONLY a Phase A signal
* RC-161, RMW TO CNMT, remains open
o RC-167, RMW TO PRT, closes
c. o EITHER a Phase A signal OR a high PRT level
* RC-161, RMW TO CNMT, closes
* RC-167, RMW TO PRT, remains open
d. * EITHER a Phase A signal OR a high PRT level
o RC-161, RMW TO CNMT, remains open
* RC-167, RMW TO PRT, closes
ANSWER:
a. o ONLY a Phase A signal
"* RC-161, RMW TO CNMT, closes
"* RC-167, RMW TO PRT, remains open
Harris Nuclear Plant
RO Written Questions
QUESTION: 78 Given the following conditions:
- An Emergency Boration of the RCS is required to be performed.
- Boric Acid Pump 'B' is operating.
Which of the following will result in the SLOWEST boration of the RCS in accordance with AOP-002, "Emergency Boration?"
a. Open both 1CS-283, Boric Acid To Boric Acid Blender FCV-1 13A and 1CS-155,
Make Up To VCT FCV-l 14A
b. Open both 1CS-283, Boric Acid To Boric Acid Blender FCV-1 13A and 1CS-156,
Make Up To CSIP Suction FCV-l 13B
c. Open either 1CS-291, Suction From RWST LCV-1 15B or 1CS-292, Suction From
RWST LCV-1 15D
d. Open 1CS-278, Emergency Boric Acid Addition
ANSWER:
a. Open both 1CS-283, Boric Acid To Boric Acid Blender FCV-1 13A and 1CS-155,
Make Up To VCT FCV-1 14A
Harris Nuclear Plant
RO Written Questions
QUESTION: 79 The compensating voltage on Intermediate Range (IR) channel N-35 is set too low, resulting in N-35 stabilizing at 2 x 10-10 amps during a reactor shutdown. When IR channel N-36 drops below 5 x 10-11 amps ...
a. BOTH SR NIs will automatically energize. b. ONLY SR channel N-31 will automatically energize. c. ONLY SR channel N-32 will automatically energize. d. NEITHER SR NI will automatically energize.
ANSWER:
d. NEITHER SR NI will automatically energize.
Harris Nuclear Plant
RO Written Questions
QUESTION: 80 Which of the following is an acceptable condition in accordance with AOP-0 10, "Feedwater Malfunctions?"
a. * Power level at 65%
* No (0) Heater Drain Pumps operating
o One (1) train of Feed Water Pumps (FWP, CBP, CP) operating
b. * Power level at 98%
0 One (1) Heater Drain Pump operating
* Two (2) trains of Feed Water Pumps (FWP, CBP, CP) operating
c. * Power level at 92%
* No (0) Heater Drain Pumps operating
* Two (2) trains of Feed Water Pumps (FWP, CBP, CP) operating
d. * Power level at 70%
0 One (1) Heater Drain Pump operating
* One (1) train of Feed Water Pumps (FWP, CBP, CP) operating
ANSWER:
b. * Power level at 98%
"* One (1) Heater Drain Pump operating
"* Two (2) trains of Feed Water Pumps (FWP, CBP, CP) operating
Harris Nuclear Plant
RO Written Questions
QUESTION: 81 Which of the following describes the automatic operation of 1SA-506, Service Air Header Isolation Valve?
a. Opens if Instrument Air pressure decreases to < 90 psig b. Closes if Service Air pressure decreases to < 90 psig c. Opens if Service Air pressure decreases to < 90 psig d. Closes if Instrument Air pressure decreases to < 90 psig
ANSWER:
d. Closes if Instrument Air pressure decreases to < 90 psig
Harris Nuclear Plant
RO Written Questions
QUESTION: 82 Given the following conditions:
- PATH-2 is being performed in response to a tube rupture on SG 'C'.
- SI termination criteria is being checked.
- If SI termination criteria cannot be met, a transition to another procedure is required.
Over the last several minutes:
- RCS subcooling has gone from 60 0 F to 63°F.
- Levels in SG 'A' and 'B' have increased from 45% to 50%.
- PRZ level has increased from 43% to 47%.
- RCS temperature has gone from 472SF to 468TF.
- RCS pressure has gone from 885 psig to 880 psig.
The Unit-SCO asks if "RCS pressure is stable or increasing." Which of the following describes how the operator should responsd AND what actions should be taken?
a. e RCS pressure is still decreasing.
* The cooldown must be stablized to determine if RCS pressure has stabilized.
b. * RCS pressure is still decreasing.
* SI cannot be terminated and a transition to the appropriate procedure must be
made.
c. o RCS pressure is stable.
o The cooldown must be stabilized in order to verify the trend.
d. o RCS pressure is stable.
* SI can be terminated.
ANSWER:
d. o RCS pressure is stable.
o SI can be terminated.
Harris Nuclear Plant
RO Written Questions
QUESTION: 83 Which of the following are the lower and upper limits on H2 concentration for placing the Electric Hydrogen Recombiners in service in accordance with OP-125, "Post Accident Hydrogen System?"
a. 0.5% - 4.0% b. .0.5% - 6.0% c. 1.0% - 4.0% d. 1.0% - 6.0%
ANSWER:
a. 0.5% - 4.0%
Harris Nuclear Plant
RO Written Questions
QUESTION: 84 Given the following conditions: "* The plant is operating at 100% power. "* The Condenser Zone 1 low-pressure turbine boot seal ruptures, causing a turbine trip
and reactor trip.
"* A complete loss of Zone 1 and Zone 2 vacuum occurs. Which of the following will automatically actuate to stabilize RCS temperature AND why are the Condenser Steam Dumps NOT used?
a. o Steam Generator PORVs
* Ensure the condenser does NOT reach saturation conditions
b. o Steam Generator PORVs
* Protect the condenser from an overpressure condition
c. * Atmospheric Steam Dumps
* Ensure the condenser does NOT reach saturation conditions
d. o Atmospheric Steam Dumps
* Protect the condenser from an overpressure condition
ANSWER:
b. o Steam Generator PORVs
o Protect the condenser from an overpressure condition
Harris Nuclear Plant
RO Written Questions
QUESTION: 85 Which of the following sets of conditions would require that AOP-022, "Loss of Service Water," be performed?
a. e ALB-002-1-1, EMER SERV WTR PMPS HDR STR HIGH AP OR LOSS OF
PWR, in alarm
* Local strainer AP indicating 20 psid
* ESW header pressure indicating 51 psig
b. * ALB-002-3-2, SW BSTR PUMP A AUTO START FAIL/OVERRIDE, in
alarm
* ALB-002-3-3, SW BSTR PUMP A O/C TRIP OR CLOSE CKT TROUBLE,
in alarm
* SI actuated
c. * ALB-002-6-2, SERV WTR PUMPS DISCHARGE VLV NOT FULL OPEN,
in alarm
* NSW Pump 'A' has been started in Priming Mode
* NSW Pump discharge valve 10% open
d. * ALB-002-7-5, COMPUTER ALARM SERVICE WATER, in alarm
* Main Reservoir temperature indicating 94 0 F
* Aux Reservoir temperature indicating 88'F
ANSWER:
a. * ALB-002-1-1, EMER SERV WTR PMPS HDR STR HIGH AP OR LOSS OF
PWR, in alarm
"* Local strainer AP indicating 20 psid
"* ESW header pressure indicating 51 psig
Harris Nuclear Plant
RO Written Questions
QUESTION: 86 Given the attached form from OST-1093 (next page) and the following conditions: "* Maintenance has been performed on 1CS-752 SB, Charging/SI Pump B-SB Alternate
Miniflow.
"* A full flow test of the valve has been performed in accordance with OST-1093,
"CVCS/SI System Operability Train B."
"* Stroke time in open direction was 5.06 seconds. "* Stroke time in closed direction was 8.02 seconds. Which of the following conditions apply to the results of the test?
a. e Declare the valve operable
* No additional paperwork is required
b. * Retest the valve if no mechanical failures are known to exist
* If the valve is within limits on retest, declare the valve operable
* No additional paperwork is required
c. * Retest the valve if no mechanical failures are known to exist
* If the valve is within limits on retest, declare the valve operable
* Initiate a Condition Report identifying the test results
d. * Declare the valve inoperable
0 Initiate a Condition Report identifying the test results
ANSWER:
c. * Retest the valve if no mechanical failures are known to exist
"* If the valve is within limits on retest, declare the valve operable
"* Initiate a Condition Report identifying the test results
Harris Nuclear Plant
RO Written Questions
QUESTION: 87 Given the following conditions:
- The plant is operating at 90% power.
- Control Bank 'C' Rod D-4, located at the edge of the core near PR NI N-43, has
slipped to approximately 6 steps off the bottom of the core and appears to be stuck.
Which of the following parameters would be the LEAST LIKELY to aid in detecting the rod misalignment?
a. Quadrant Power Tilt Ratio calculations b. Power Range Nuclear Instrument indications c. Axial Flux Difference indications d. Core Exit Thermocouple temperature indications
ANSWER:
c. Axial Flux Difference indications
Harris Nuclear Plant
RO Written Questions
QUESTION: 88 Given the following conditions:
- The unit is in a Refueling Outage.
- A spent fuel assembly is attached to the manipulator crane.
- A failure of the Reactor Vessel permanent cavity seal ring causes cavity level to drop
approximately 3" every minute.
- The Refueling Crew is in the process of placing the assembly in the Reactor Vessel
when a Loss of Off-Site Power occurs.
Upon the loss of off-site power there is NO means for ...
a. making up to the cavity. b. monitoring radiological levels inside Containment. c. monitoring the reactivity condition of the core. d. placing the fuel assembly in the vessel.
ANSWER:
d. placing the fuel assembly in the vessel.
Harris Nuclear Plant
RO Written Questions
QUESTION: 89 FRP-J.1, "Response to High Containment Pressure," monitors the status of the ESW Booster Pumps. Which of the following is the concern if ESW Booster pumps are NOT running while high containment pressure conditions exist?
a. ESW Pump runout b. Flooding of safety equipment in containment c. Reduced containment cooling capability d. Radioactivity release to the environment
ANSWER:
d. Radioactivity release to the environment
Harris Nuclear Plant
RO Written Questions
QUESTION: 90 Given the following conditions:
- A plant cooldown is in progress.
- All three (3) RCPs are operating.
- ALB-010-8-5A, "CMPTR ALARM RX COOLANT," is in alarm and investigation
reveals RCP 'C' Radial Bearing Temp has exceeded the warning alarm setpoint and is approaching the alarm limit of 2200 F.
- Seal injection to each RCP is approximately 10 gpm.
Which of the following conditions would direct opening the RCP No. 1 seal bypass valve in accordance with OP-100, "Reactor Coolant System?"
a. * RCS pressure 850 psig
* RCP 'A' No. 1 seal leakoff 1.2 gpm
* RCP 'B' No. 1 seal leakoff 0.9 gpm
* RCP 'C' No. 1 seal leakoff 1.3 gpm
b. * RCS pressure 1060 psig
* RCP 'A' No. 1 seal leakoff 0.7 gpm
* RCP 'B' No. 1 seal leakoff 0.9 gpm
0 RCP 'C' No. i seal leakoff 0.8 gpm
c. e RCS pressure 640 psig
* RCP 'A' No. 1 seal leakoff 1.2 gpm
* RCP 'B' No. 1 seal leakoff 1.1 gpm
* RCP 'C' No. 1 seal leakoff 1.1 gpm
d. * RCS pressure 1110 psig
* RCP 'A' No. 1 seal leakoff 1.2 gpm
0 RCP 'B' No. 1 seal leakoff 0.8 gpm
* RCP 'C' No. 1 seal leakoff 0.9 gpm
ANSWER:
a. * RCS pressure 850 psig
"* RCP 'A' No. 1 seal leakoff 1.2 gpm
"* RCP 'B' No. 1 seal leakoff 0.9 gpm
"* RCP 'C' No. 1 seal leakoff 1.3 gpm
Harris Nuclear Plant
RO Written Questions
QUESTION: 91 A valve lineup calls for a valve to be OPEN and by the difference of grease on the stem and the stem length exposure the valve appears open. The operator verifying this valve open should ...
a. sign the valve off as open. b. try to open the valve to ensure it is full open. c. move the valve closed and then reopen the valve. d. observe down stream system flow or pressure to ensure open.
ANSWER:
c. move the valve closed and then reopen the valve.
Harris Nuclear Plant
RO Written Questions
QUESTION: 92 Which of the following is the basis for the Technical Specification limit of 31 gpm on Controlled Leakage?
a. Sufficiently low to ensure early detection of additional leakage b. Allows limited known leakage with the ability to detect additional leakage c. Ensures safety injection flow is greater than that analyzed for a LOCA d. Keeps dose to a small fraction of limits in the event of a SGTR or steam line break
ANSWER:
c. Ensures safety injection flow is greater than that analyzed for a LOCA
Harris Nuclear Plant
RO Written Questions
QUESTION: 93 The unit is operating at 100% power. If 125 VDC Bus 1A-SA deenergizes due to a fault on the bus ...
a. the reactor will trip due to an undervoltage (UV) trip of Train SA reactor trip
breaker.
the reactor will trip due to a shunt trip of Train SA reactor trip breaker.
b.
C.
an undervoltage trip signal will NOT be capable of opening Train SA reactor trip
breaker.
d. a shunt trip signal will NOT be capable of opening Train SA reactor trip breaker.
ANSWER:
d. a shunt trip signal will NOT be capable of opening Train SA reactor trip breaker.
Harris Nuclear Plant
RO Written Questions
QUESTION: 94 A waste gas release is in progress when the WPB Stack 5 PIG radiation monitor, REM 1WV-3546, exceeds the high alarm setpoint. Which of the following describes how the release will be automatically terminated?
a. Waste Gas Decay Tanks E & F to Plant Vent, 3WG-229, CLOSES b. Running Waste Gas Compressor TRIPS c. Filtered Exhaust Fans, E-46, E-47, E-48, and E-49 TRIP d. Gas Decay Tanks to Plant Vent Isolation Valve, 3WG-230, CLOSES
ANSWER:
a. Waste Gas Decay Tanks E & F to Plant Vent, 3WG-229, CLOSES
Harris Nuclear Plant
RO Written Questions
QUESTION: 95 Given the following conditions: "* The plant is at 22% power during a shutdown. "* Source Range Channel N-31 has been declared inoperable as a result of failing to
meet Operational Test Criteria of MST-I0169.
"* The test was performed, per GP-006, during a Tech Spec 3.0.3 required shutdown
(i.e., the shutdown must continue).
"* OWP-RP-19 has been performed, which places the LEVEL TRIP BYPASS switch in
the BYPASS position and verifies the associated light on the Bypass Permissive Light Panel.
"* The I&C Supervisor states that both control and instrument power must be removed
from the drawer to replace a bistable module.
Assuming the instrument and control power are removed for the remainder of the shutdown, the shutdown continues and ...
a. the reactor trips when the fuses are removed. b. the reactor trips when power is reduced below P-10. c. the reactor trips when power is reduced below P-6. d. NO reactor trip occurs.
ANSWER:
c. the reactor trips when power is reduced below P-6.
Harris Nuclear Plant
RO Written Questions
QUESTION: 96 Which of the following describes the operation of 1CS-50, Letdown to VCT/Demin?
a. Automatically bypasses the CVCS Demnineralizers when TE-143, LP Letdown
Temperature, exceeds 135 0F to prevent an inadvertent dilution event
b. Automatically bypasses the CVCS Demineralizers when TE-143, LP Letdown
Temperature, exceeds 135 0 F to prevent damage to the demineralizer resin
c. Automatically bypasses the CVCS Demineralizers when TE-144, Letdown HX
Outlet Temperature, exceeds 135 0F to prevent an inadvertent dilution event
d. Automatically bypasses the CVCS Demineralizers when TE-144, Letdown HX
Outlet Temperature, exceeds 135 0 F to prevent damage to the demineralizer resin
ANSWER:
b. Automatically bypasses the CVCS Demineralizers when TE-143, LP Letdown
Temperature, exceeds 135TF to prevent damage to the demineralizer resin
Harris Nuclear Plant
RO Written Questions
QUESTION: 97 Given the following conditions:
- During Mode 3 operations, a large break LOCA occurred concurrently with a loss of
offsite power.
- Both EDGs started and loaded.
- The BOP operator secured all running AFW pumps with SG levels all at 55%.
- SI has NOT been reset and offsite power has NOT been restored.
If SG narrow range levels decrease to the following levels,
SG LEVEL
A 22%
B 17%
C 27%
which of the following describes the expected AFW pump operation?
a. * MDAFW Pump 1A-SA running
* MDAFW Pump 1B-SB running
* TDAFW Pump running
b. * MDAFW Pump 1A-SA running
* MDAFW Pump lB-SB running
* TDAFW Pump secured
c. o MDAFW Pump 1A-SA secured
* MDAFW Pump lB-SB secured
* TDAFW Pump running
d. o MDAFW Pump 1A-SA secured
* MDAFW Pump lB-SB secured
0 TDAFW Pump secured
ANSWER:
c. * MDAFW Pump 1A-SA secured
"* MDAFW Pump lB-SB secured
"* TDAFW Pump running
Harris Nuclear Plant
RO Written Questions
QUESTION: 98 Given the following conditions: "* A small break LOCA occurred and SI has been terminated in accordance with EPP
008, "SI Termination."
"* SI Reinitiation criteria has been met. Which of the following should be isolated PRIOR TO re-opening the BIT valves to prevent CSIP runout conditions?
a. Charging b. Seal injection c. Normal miniflow d. Alternate miniflow
ANSWER:
a. Charging
Harris Nuclear Plant
RO Written Questions
QUESTION: 99 Which of the following Emergency Operating Procedure actions may the operating crew perform before being directed to perform the step in the appropriate EOP in accordance with the EOP User's Guide?
a. Isolating AFW to a ruptured SG and closing the Main Steam Isolation Valve from
the ruptured SG immediately upon completion of the Immediate Actions of
PATH-1
b. Throttling AFW to supply all three (3) SGs at 100 KPPH per SG with one AFW
pump during the performance of EPP-FRP-S. 1, "Response to Nuclear Power
Generation - ATWS," with all SG narrow range levels below 5%
c. Restoring AFW to a faulted SG after it has been isolated to establish the maximum
cooldown rate available during the performance of PATH-2
d. Throttling AFW flow to all three (3) SGs at 50 KPPH per SG to maintain intact
SG levels between 25% and 50% immediately upon completion of the Immediate
Actions of PATH-1
ANSWER:
d. Throttling AFW flow to all three (3) SGs at 50 KPPH per SG to maintain intact
SG levels between 25% and 50% immediately upon completion of the Immediate
Actions of PATH-1
Harris Nuclear Plant
RO Written Questions
QUESTION: 100 Given the following conditions: "* Following a large break LOCA, SI has been reset. "* The crew has just completed the performance of EPP-010, "Transfer to Cold Leg
Recirculation," and have transitioned back to PATH-1.
"* A loss of offsite power occurs following the transition to PATH-1. Core cooling will be ...
a. automatically re-established immediately when the Emergency Diesel Generator
output breaker closes.
b. automatically re-established within minutes after the Emergency Diesel Generator
output breaker closes.
c. lost until the operator re-opens the RHR suction valves after the Emergency Diesel
Generator output breaker closes.
d. lost until the operator restarts the RHR pumps after the Emergency Diesel
Generator output breaker closes.
ANSWER:
d. lost until the operator restarts the RHR pumps after the Emergency Diesel
Generator output breaker closes.
Harris Nuclear Plant
SRO Written Questions
QUESTION: 76 Given the following conditions: "* Both RHR pumps are inoperable. "* At 0700 today, during repair efforts, a Maintenance person exited the area after
receiving a Total Effective Dose Equivalent of 5800 mRem.
"* At 0730 today, a plant shutdown was commenced due to both RHR pumps being
inoperable.
When are the notifications to the NRC required to be completed by for these events?
a. 00745 today for the plant shutdown
0800 today for the over-exposure
b. * 0745 today for the plant shutdown
* 0700 tomorrow for the over-exposure
C. 1130 today for the plant shutdown
0800 today for the over-exposure
d. * 1130 today for the plant shutdown
* 0700 tomorrow for the over-exposure
ANSWER:
d. * 1130 today for the plant shutdown
* 0700 tomorrow for the over-exposure
Harris Nuclear Plant
SRO Written Questions
QUESTION: 77 Given the following conditions:
- The plant is operating at 60% power.
- RCS Tavg is on program.
- PRZ level is on program.
- The Median Tavg input to PRZ level fails high.
- The operator places LK-459, PRZ Master Level Controller, in MANUAL.
In order to control PRZ level at program level, the operator should be directed to control PRZ level at ...
a. 36%. b. 41%. c. 46%. d. 51%.
ANSWER:
c. 46%.
Harris Nuclear Plant
SRO Written Questions
QUESTION: 78 Given the following conditions: "* A LOCA occurred several hours ago, resulting in a start of both Containment Spray
Pumps.
"* Only one (1) Containment Spray Pump is currently running due to actions taken in
EPP-012, "Loss of Emergency Coolant Recirculation."
"* A transition has just been made to FRP-J. 1, "Response to High Containment
Pressure."
"* Containment Pressure is 14 psig. Which of the following actions should be taken?
a. Restart the second Containment Spray Pump if Containment pressure does NOT
decrease below 10 psig before exiting FRP-J.1.
b. Restart the second Containment Spray Pump since pressure is above 10 psig.
c. Continue operation with one Containment Spray Pump.
d. Continue operation with one Containment Spray Pump unless Containment
pressure begins increasing, then start the second pump.
ANSWER:
c. Continue operation with one Containment Spray Pump.
Harris Nuclear Plant
SRO Written Questions
QUESTION: 79 In addition to Radiation Levels inside containment, which of the following parameters are checked when determining whether an entry is required to be made into EPP-FRP-J.3, "Response to Containment High Radiation Level?"
a. Containment Pressure, Containment Sump Levels, and Containment Spray Pump
status
b. ONLY Containment Pressure and Containment Sump Levels
c. ONLY Containment Pressure and Containment Spray Pump status
d. ONLY Containment Sump Levels and Containment Spray Pump status
ANSWER:
b. ONLY Containment Pressure and Containment Sump Levels
Harris Nuclear Plant
SRO Written Questions
QUESTION: 80 Given the following conditions:
- Just prior to shift change, the oncoming Reactor Operator calls in sick.
- The shift schedule shows the oncoming crew at minimum complement with the
Reactor Operator, but there is a Licensed Operator (CO) scheduled for the RAB.
The S-SO should ...
a. use the RAB CO in the control room and replace the RAB whenever possible.
b. use the RAB CO in the control room and call in a replacement RAB within two
hours.
c. hold the off-going CO until he can ensure a replacement will arrive within two
hours.
d. hold the off-going CO until a replacement can relieve him.
ANSWER:
d. hold the off-going CO until a replacement can relieve him.
Harris Nuclear Plant
SRO Written Questions
QUESTION: 81 Given the following conditions: "* During refueling operations, NO spent fuel assembly is attached to the manipulator
crane gripper.
"* Refueling personnel discover that they are unable to move the bridge or trolley. Which of the following is the most likely cause of the inability to move the bridge or trolley?
a. The gripper is NOT in the FULL UP position inside the mast b. The gripper is NOT at least 12 inches up inside the mast c. The gripper is failed in the ENGAGED position d. A slack cable condition is sensed by the gripper
ANSWER:
b. The gripper is NOT at least 12 inches up inside the mast
Harris Nuclear Plant
SRO Written Questions
QUESTION: 82 Which one of the following describes the bases for the off-site power distribution Technical Specification that requires two (2) independent off-site power sources?
a. The requirement can only be satisfied by the off-site transmission lines that feed
the SUTs directly (Cary Regency Park and Cape Fear North)
b. The requirement can only be satisfied by the off-site transmission lines that do not
feed the respective north or south switchyard bus through a jumper
c. The requirement is satisfied as long as the switchyard alignment is such that power
is available from the off-site transmission network to both SUTs regardless of the
number of transmission lines available
d. The requirement is satisfied as long as there are two separate off-site transmission
lines that can power the SUTs (either through the switchyard or directly)
ANSWER:
d. The requirement is satisfied as long as there are two separate off-site transmission
lines that can power the SUTs (either through the switchyard or directly)
Harris Nuclear Plant
SRO Written Questions
QUESTION: 83 Given the following conditions: "* During a refueling outage, the SRO-Fuel Handling reports that the crew is having
difficulties loading several fuel assemblies in the vicinity of the hot legs due to the flow through the piping.
"* He has requested that the RHR system be secured to allow loading the assemblies. Which of the following identifies how long the RHR system may be secured under these conditions?
a. 1 hour per 2-hour period b. 1 hour per 4-hour period c. Up to2hours d. Up to 4 hours
ANSWER:
a. 1 hour per 2-hour period
Harris Nuclear Plant
SRO Written Questions
QUESTION: 84 An action that deviates from the Facility Operating License requirements for fire protection is deemed necessary to protect the public health and safety. Which of the following is required, according to PRO-NGGC-0200, "Procedure Use and Adherence?"
a. The deviation shall be approved by the Manager - Operations prior to performing
the protective action
b. The deviation shall be approved by the Superintendent - Shift Operations prior to
performing the protective action
c. The state and counties must be notified as soon as possible after performing the
protective action and within 60 minutes in all cases
d. The NRC must be notified prior to performing the protective action
ANSWER:
b. The deviation shall be approved by the Superintendent - Shift Operations prior to
performing the protective action
Harris Nuclear Plant
SRO Written Questions
QUESTION: 85 Given the following conditions: "* Following a reactor trip and safety injection concurrent with a loss of offsite power, a
transition has been made to EPP-015, "Uncontrolled Depressurization Of All Steam Generators."
"* Emergency Diesel Generator lB-SB has tripped and cannot be restarted. "* The TD AFW pump has tripped on overspeed and cannot be reset. "* MDAFW pump 1A-SA is tagged out. "* SG 'A' narrow level is 6%. "* SG 'B' and 'C' narrow range levels are off-scale low. "* Core exit thermocouple temperatures are all between 7050 F and 7200 F. Which of the following actions should be taken?
a. Continue in EPP-015, "Uncontrolled Depressurization Of All Steam Generators" b. Transition to EPP-001, "Loss of AC Power to 1A-SA and lB-SB Buses" c. Transition to EPP-FRP-C.1, "Response to Inadequate Core Cooling" d. Transition to EPP-FRP-H.1, "Response to Loss of Secondary Heat Sink"
ANSWER:
d. Transition to EPP-FRP-H.1, "Response to Loss of Secondary Heat Sink"
Harris Nuclear Plant
SRO Written Questions
QUESTION: 86 Given the following conditions:
- The unit is in Mode 5 with the RCS filled.
- RHR Train 'A' is in operation.
- RHR Train 'B' is operable, but not in operation.
- SG wide range levels are:
SG LEVEL
A 81%
B 68%
C 63%
- Maintenance requests that RHR Pump 'B' be removed from operable status for
several hours for minor maintenance.
Which of the following describes the acceptability of removing RHR Pump 'B' from service under these conditions?
a. It may NOT be done because the SGs are not an adequate heat sink under these
conditions.
b. It may NOT be done because two RHR trains are required at all times for Mode 5.
c. It may be done as long as the RCS remains filled.
d. It may be done as long as RCS temperature remains below 200TF.
ANSWER:
a. It may NOT be done because the SGs are not an adequate heat sink under these
conditions.
Harris Nuclear Plant
SRO Written Questions
QUESTION: 87 Given the following conditions: "* Safety injection is being terminated in accordance with EPP-008, "SI Termination." "* The operator reports 1SI-3, BIT Outlet, is closed as directed, but 1SI-4, BIT Outlet,
will NOT close.
"* An operator unsuccessfully attempts to locally close 1SI-4. Which of the following actions should be taken?
a. e Unlock and close 1SI-2, BIT Inlet, ONLY
* Establish normal charging flow while waiting for 1SI-2 to be closed
b. * Unlock and close 1SI-2, BIT Inlet, ONLY
* Wait until 1SI-2 is closed before establishing normal charging flow
c. * Unlock and close BOTH 1SI-1, BIT Inlet, and 1SI-2, BIT Inlet
* Establish normal charging flow while waiting for 1SI-1 and 1SI-2 to be closed
d. * Unlock and close BOTH ISI-1, BIT Inlet, and 1SI-2, BIT Inlet
* Wait until 1SI-1 and 1SI-2 are closed before establishing normal charging flow
ANSWER:
d. * Unlock and close BOTH 1S-1, BIT Inlet, and 1SI-2, BIT Inlet
* Wait until 1SI-1 and 1SI-2 are closed before establishing normal charging flow
Harris Nuclear Plant
SRO Written Questions
QUESTION: 88 Given the following conditions:
- At 0530, RCS temperature was being maintained at 550TF.
- A small break LOCA occurred.
- At 0545, the crew is ready to commence a cooldown to cold shutdown in accordance
with EPP-009, "Post LOCA Cooldown and Depressurization."
- RCS temperature at 0545 is 4900 F.
Which of the following identifies the lowest allowable temperature of the RCS at 0630 if the crew begins the MAXIMUM permissible cooldown rate AND the basis for this temperature limit?
a. 450TF to ensure that a transition is NOT required to be made to FRP-P. 1,
"Response to Imminent Pressurized Thermal Shock"
b. 4500 F to ensure that Technical Specification cooldown limits are NOT exceeded
c. 415TF to ensure that a transition is NOT required to be made to FRP-P. 1,
"Response to Imminent Pressurized Thermal Shock"
d. 415TF to ensure that Technical Specification cooldown limits are NOT exceeded
ANSWER:
b. 450TF to ensure that Technical Specification cooldown limits are NOT exceeded
Harris Nuclear Plant
SRO Written Questions
QUESTION: 89 Given the following conditions:
- Following a loss of offsite power, EPP-001, "Loss of AC Power to 1A-SA and lB-SB
Buses," is being performed.
- Safety Injection has been actuated and reset.
- Attachment 5, "6.9 KV Emergency Bus Breakers," has been completed and all
breakers have been verified open.
- The SGs are being depressurized.
- Several minutes later, Emergency Diesel Generator 1A-SA is started.
- SG pressures are stabilized.
- ESW Pump 1A-SA is started and the valve alignment for Header 'A' has been
verified.
Plant conditions are now:
- EDG 1A-SA is running.
- ESW Pump 1A-SA is running.
- NO other pumps are running.
- NO SI valves have repositioned from their "at power" position.
- RCS pressure is 1400 psig.
- RCS temperature is 492TF.
- RCS subcooling is 96TF.
- PRZ level is 6%.
Which of the following identifies the procedure(s) to be used for recovery from this condition?
a. EPP-002, "Loss Of All AC Power Recovery Without SI Required"
b. EPP-003, "Loss Of All AC Power Recovery With SI Required"
c. EOP-PATH-1 and AOP-025, "Loss of One Emergency AC Bus or One
Emergency DC Bus," performed concurrently
d. EOP-PATH-1 and FRP-I.2, "Response to Low Pressurizer Level," performed
concurrently
ANSWER:
b. EPP-003, "Loss Of All AC Power Recovery With SI Required"
Harris Nuclear Plant
SRO Written Questions
QUESTION: 90 Conditions meeting the Emergency Classification criteria for a Notification of Unusual Event have been determined to have existed, but no longer exist. As the Site Emergency Coordinator you should ...
a. declare and terminate the event in a single notification message.
b. declare the event in a notification message and terminate the event in a followup
message.
c. notify the NRC of the conditions, but NO notifications to the state and county
would be performed.
d. notify Licensing of the need to generate an LER, but no other notifications would
be performed.
ANSWER:
a. declare and terminate the event in a single notification message.
Harris Nuclear Plant
SRO Written Questions
QUESTION: 91 A reactor startup is being performed following a mid-cycle outage per GP-004, "Reactor Startup (Mode 3 to Mode 2)." Estimated Critical Conditions are as follows:
TIME 1830
BORON CONC. 1215 ppm
CONT BANK 'C' POSTION 218 steps
CONT BANK 'D' POSTION 90 steps
ECC - 500 PCM POSITION 45 steps on Bank 'D'
ECC + 500 PCM POSITION 197 steps on Bank 'D'
ROD INSERTION LIMIT 0 steps on Bank 'D'
The Actual Critical Conditions are as follows:
TIME 1836
BORON CONC. 1198 ppm
CONT BANK 'C' POSTION 110 steps
CONT BANK 'D' POSTION 0 steps
Which of the following actions must be taken?
a. Shut down the reactor using OP-104, "Rod Control System," AND borate, as
needed, to increase RCS boron concentration to 1215 ppm
b. Maintain critical conditions AND borate, as needed, to increase RCS boron
concentration to 1215 ppm
c. Shut down the reactor using OP-104, "Rod Control System," AND initiate
Emergency Boration per AOP-002, "Emergency Boration"
d. Trip the reactor AND initiate Emergency Boration per AOP-002, "Emergency
Boration"
ANSWER:
c. Shut down the reactor using OP-104, "Rod Control System," AND initiate
Emergency Boration per AOP-002, "Emergency Boration"
Harris Nuclear Plant
SRO Written Questions
QUESTION: 92 Given the following conditions:
- A small break LOCA has occurred.
- Entry has been made into FRP-C. 1, "Response to Inadequate Core Cooling."
- Core exit thermocouples are all indicating between 740 OF and 760 OF and rising
slowly.
- RCS pressure has stabilized at 805 psig.
- PZR level is off-scale low.
- RVLIS Full Range is indicating 32% and lowering slowly.
- NO CSIPs are available.
- SG narrow range levels are all off-scale low.
- Total AFW flow to the SGs is 240 KPPH.
Which of the following actions should be taken?
a. Dump steam to cooldown and depressurize the RCS to cause the SI accumulators
to dump
b. Open the RCS Head Vent valves to depressurize the RCS to cause the SI
accumulators to dump
c. Start an RCP immediately to provide forced cooling flow
d. Open the PRZ PORVs to depressurize the RCS to cause the SI accumulators to
dump
ANSWER:
a. Dump steam to cooldown and depressurize the RCS to cause the SI accumulators
to dump
Harris Nuclear Plant
SRO Written Questions
QUESTION: 93 Given the following conditions:
- A large break LOCA has occurred.
- EPP-012, "Loss of Emergency Coolant Recirculation," is being performed.
- One (1) CSIP is operating with a flow rate of 520 gpm.
- One (1) RHR pump is operating with a flow rate of 3350 gpm.
- Time after trip and SI is 73 minutes.
- SI CANNOT be terminated due to insufficient subcooling.
Which of the following actions should be taken to establish SI flow greater than the minimum required for decay heat removal?
a. Stop the CSIP b. Start the standby CSIP c. Manually throttle high head SI flow d. Stop the RHR pump
ANSWER:
d. Stop the RHR pump
Harris Nuclear Plant
SRO Written Questions
QUESTION: 94 Given the following conditions:
- A reactor startup is in progress.
- Power level is stable at 10- 8 amps.
- Electrical Maintenance reports there is a potential problem with the inverter for
Instrument Bus IDP-1A-SI and recommends placing the bus on the alternate power supply (PP- 1A21 1-SA).
Which of the following describes the effect of permitting this re-alignment?
a. NO reactor trip occurs, but the reactor startup is delayed due to C-1, Intermediate
Range Rod Stop
b. NO reactor trip occurs, but the reactor startup is delayed due to C-2, Power Range
Overpower Rod Stop
c. Reactor trip on Intermediate Range High Flux
d. Reactor trip on Power Range High Flux Low Setpoint
ANSWER:
c. Reactor trip on Intermediate Range High Flux
Harris Nuclear Plant
SRO Written Questions
QUESTION: 95 Given the following conditions:
- A reactor trip and safety injection occurred.
- During the performance of PATH-1, an ORANGE path was noted on the Core
Cooling status tree and a transition was made to the appropriate procedure.
Which of the following describes how the CSF status trees should be monitored at this point?
a. Suspend monitoring until actions have been completed for the ORANGE path
condition
b. Monitor for information only until actions have been completed for the ORANGE
path condition
c. Monitor every 10 to 20 minutes
d. Monitor continuously
ANSWER:
d. Monitor continuously
Harris Nuclear Plant
SRO Written Questions
QUESTION: 96 Given the following conditions:
- The plant is operating at 50% power
- Train 'A' safety equipment is in service
- ALB 24-1-2, 6.9kV EMER BUS A-SA TROUBLE, in alarm
- ALB 25-1-2, 6.9kV EMER BUS B-SB TROUBLE, in alarm
- AEP-2-8, DEGRADED VOLTAGE, in alarm
- AEP-2-9, DEGRADED VOLTAGE, in alarm
"* Emergency 6.9 kV Buses lA-SA and lB-SB both indicating approximately 6500
volts
"* Emergency 480V Buses all indicating approximately 450 volts Which of the following Emergency Buses will be first to be supplied by its EDG AND which procedure will be used to direct this action?
a. e Emergency Bus A-SA
* AOP-028, "Grid Instability"
b. * Emergency Bus A-SA
* OP-155, ""Diesel Generator Emergency Power System"
c. * Emergency Bus B-SB
* AOP-028, "Grid Instability"
d. o Emergency Bus B-SB
o OP-155, ""Diesel Generator Emergency Power System"
ANSWER:
c. o Emergency Bus B-SB
* AOP-028, "Grid Instability"
Harris Nuclear Plant
SRO Written Questions
QUESTION: 97 Given the following conditions: "* A LOCA outside containment has resulted in unsafe radiological conditions in the
RAB.
"* The crew has taken all the actions of EPP-013, "LOCA Outside Containment," to
isolate the break.
Which of the following is the PRIMARY indication used in EPP-013 that the actions taken have been successful AND which procedure should be transitioned to when the isolation is successful?
a. * RAB sump level alarms clearing
* Transition to PATH-1
b. * RCS pressure increasing
* Transition to PATH-1
c. * RAB sump level alarms clearing
* Transition to EPP-008, "SI Termination"
d. * RCS pressure increasing
* Transition to EPP-008, "SI Termination"
ANSWER:
b. * RCS pressure increasing
* Transition to PATH-I
Harris Nuclear Plant
SRO Written Questions
QUESTION: 98 Given the following conditions:
- A reactor trip occurred from 23% power.
- Shutdown Bank 'B' Rod L-5 is indicating 228 steps.
- Control Bank 'C' Rod K-8 is indicating 6 steps.
- All other rods have the Rod Bottom Lights lit.
- RCS boron concentration at the time of the trip was 845 ppm.
- The plant is to be maintained at no-load Tavg.
Which of the following actions should be taken AND what is the MINIMUM RCS boron concentration that must be achieved?
a. Emergency Borate to raise RCS boron concentration to 1307 ppm b. Emergency Borate to raise RCS boron concentration to 2282 ppm c. Normal Borate to raise RCS boron concentration to 1307 ppm d. Normal Borate to raise RCS boron concentration to 2282 ppm ANSWER: a. Emergency Borate to raise RCS boron concentration to 1307 ppm
Harris Nuclear Plant
SRO Written Questions
QUESTION: 99 Given the following conditions:
- The plant is operating at 40% power.
- A fire alarm has been received.
Which of the following conditions would require that a plant shutdown be required at the earliest time?
a. e RHR Pump 1A-SA has been out-of-service for 18 hours for maintenance
* The fire requires de-energizing Emergency Bus 1A-SA
b. * RHR Pump 1A-SA has been out-of-service for 18 hours for maintenance
0 The fire is contained in the CSIP 1A-SA pump room
c. * Containment Spray Pump lB-SB has been out-of-service for 18 hours for
maintenance
* The fire requires de-energizing Aux Bux B
d. * Containment Spray Pump lB-SB has been out-of-service for 18 hours for
maintenance
* The fire is contained in the CSIP lA-SA pump room
ANSWER:
c. * Containment Spray Pump 1A-SA has been out-of-service for 18 hours for
maintenance
* The fire requires de-energizing Aux Bux B
Harris Nuclear Plant
SRO Written Questions
QUESTION: 100 Given the following plant conditions:
- A small break LOCA has occurred.
- A transition has been made to EPP-009, "Post LOCA Cooldown and
Depressurization."
- Containment pressure is 6.1 psig.
- RCS subcooling is 55 0 F by ERFIS.
- PRZ level is 31%.
- Both CSIPs are injecting through the BIT.
- Both RHR pumps are secured.
- The operators are depressurizing the RCS to refill the pressurizer to > 40% when
subcooling is noted to decrease to 35T.
Which of the following actions should be taken?
a. Continue the depressurization in EPP-009 b. Stop the depressurization and continue in EPP-009 c. Stop the depressurization and transition to PATH-1 d. Reinitiate SI and transition to PATH-1
ANSWER:
a. Continue the depressurization in EPP-009
Draft Submittal
(Pink Paper)
Written Exam Sample outlines
SHEARON HARRIS
EXAM 2002-301
50-400
AUGUST 26 - 29, 2002
ES-401 PWR SRO Examination Outline Form ES-401-3 kacility: HARRIS Date of Exam: 8/26/2002 Exam Level: SRO
K/A Category Points
Tier Group K K K K K K A A A A G Point
1 2 3 4 5 6 1 2 3 4 Total
1 1 2 1 3 5
Emergency& 2 1 2 1 2 6 4 16
Abnormal Plant 3 0 0 0 ....... 0 2 1 3
Evolutions Tier 3 4 7 6 171
Totals
2 1 1 2 2 2 2 2 1 2 1 1 3 19
Plant 2 2 0 2 2 1 1 2 1 1 3 2 17
Systems 3 1 1 0 1 0 0 1 0 0 0 0 4
Tier
To 4 3 4 5 3 2
Totals 4 5 40
3 Generic Knowledge and Abilities Cat I Cat 2 Cat 3 Cat 4
5 4 3 5 17
Notes:
1 Ensure that at least two topics from every K/A category are sampled within each tier
(i.e., the "Tier Totals" in each K/A category shall not be less than two).
2 The point total for each group and tier in the proposed outline must match that specified in the table.
The final point total for each group and tier may deviate by +/-1 from that specified in the table based
on NRC revisions. The final exam must total 100 points.
3 Select topics from many systems; avoid selecting more than two or three K/A topics from a
given system unless they relate to plant-specific priorities.
4 Systems/evolutions within each group are identified on the associated outline.
5 The shaded areas are not applicable to the category/tier.
6* The generic K/As in Tiers 1 and 2 shall be selected from Section 2 of the K/A Catalog, but the
topics must be relevant to the applicable evolution or system.
7 On the following pages, enter the K/A numbers, a brief description of each topic, the topics'
importance ratings for the SRO license level, and the point totals for each system and category.
K/As below 2.5 should be justified on the basis of plant-specific priorities. Enter the tier totals
for each category in the table above.
8 Shaded topics on individual Tier / Group pages indicate topic is SRO ONLY.
4k)
NUREG- 1021, Revision 8, Supplement 1
( (\K (
PWR SRO Examination Outline Form ES-401-31
ES-401
S-401 _Emergency and Abnormal Plant Evolutions - Tier 1/Group 1
Al A2 G K/A Topic(s) Imp. Points
E/APE # / Name / Safety Function K1 K2 K3
AK 1.16 - Definition / application of power defect 3.4 1
000001 Continuous Rod Withdrawal / 1 X
X 2.4.6 - Knowledge of symptom based EOP strategy 4.0 1
000003 Dropped Control Rod / 1 4.4 1 000005 Inoperable/Stuck Control Rod / 1 X- AA2.O3 - Actions ifmore than one rod is stucki inoperable
EA1.01 - Control of RCS pressure and temp to avoid violating 3.8 1
000011 Large Break LOCAl 3 (PSA)
LX PTS limits during a LBLOCA
W/E04 LOCA Outside Containment / 3 EA2.01 - on of poceduesduring response to LOCA
Xoutside containment,
X EA2.02 - Adherence to procedures and operations within limits 4ý0
W/EO 1 & E02 Rediagnosis & SI Termination 3
in license during SI Termination
X AA1.08 - Operate / monitor SG LCS during loss of RCS flow 2.9 1
000015/17 RCP Malffinctions / 4 3.7
XEK1.02 - Normal, abnormal, and emergency procedures
WE09&E10 Natural Cue. / 4
X associated with NC operations 1
4.2
X 2.20 AbIlitt execbteprocedurestep
000024 Emergency Boration/1
3.5. 1
000026 Loss of Component Cooling Water 8S-----------------AI3.0l - Conditions that will initiate automatic operation npofo e
1 L X SWS valves to CCW coolers elto si.ewe
EK3.11 - Reasons for s sesdn
initiating Emergency Boration 4.3 1
000029 Anticipated Transient w/o Scram / 1 X eo vale
X_-000040AA1.22 EK0(X2. - He t - Operate / monitor load sequencer status lights
4.1 3.0 1
000040 (W/E12) Steam Line Rupture - Excessive Heat
000ransfer / 4 during a steam line rupture
XEK2.02 - Heat removal systems, and relationship between proper 4.0 1
/EO8 RCS Overcooling - PTS /4 /- operation of systems during PTS event
Xl AA2*ý Loissofvacuum reqiiring actor I turbine trip :4 A 1
000051 Loss of Condenser Vacuum / 4 3.6 1
000055 Station Blackout / 6 X 2.4.18-Knowledge of specific bases for EOPs
X AA2.15 - Determine / interpret that a loss of AC has occurred 4.1 1
000057 Loss of Vital AC Elec. Inst. Bus / 6 1
X AA2.03 - Failure modes, symptoms, and causes of misleading 36
000059 Accidental Liquid RadWaste Rel. / 9
on liquid rad monitor R
1tindications
X AA2.06 - Length of time after loss of SWS flow to component 3.1 1
00062 Loss of Nuclear Service Water 4 Iuntil damage to component occurs
AA2.13 codtosrqdrn
- Fire60 mrecy plant shutdown::ý 4A.4 I
0007Pan ieOnstX
AK3 .09 - Transfer to local control - charging pumps, flow 4.4 1
control, PZR heaters, and boric acid transfer pumps
---- --- ---
000068_____Control____Room________ 0 Cl E /
i2.310l to reduce
Perfomprocedure ivexlevelsoUof 3U3
000069 (W/E14) Loss of CTMT Integrity /5
radion a guard -aaist personnel exposure
X 000074 EAI.05 - Operate / monitor PZR PORV during ICC 4.1 2
000074 (W/E06&E07) Inad. Core Cooling / 4 (PSA)
W/E6E2.O
-Selcton f rocduesduring res ponse to 4.2
:L:-:'16er~cr6-Cooling
000076 High Reactor Coolant Activity/ 9 X 2.4.18 - Knowledge of specific bases for EOPs 3.6 1
lK/A CategoryTotals: )2 i 1 L A3..L4 9. 5 Group PointTotal: j 24
NUREG- 1021, Revision 8, Supplement 1
C C (
fES-401 PWR SRO Examination Outline Form ES-401-3
Emergency and Abnormal Plant Evolutions - Tier 1/Group 2
E/APE # / Name / Safety Function K1 K2 rK3 Al A2 G K/A Topic(s) Imp. Points
000007 Reactor Trip - Stabilization - Recovery / 1 X EKL.03 - Reasons for isolating the TG after a reactor trip 4.0 1 000008 Pressurizer Vapor Space Accident / 3 X AA2.12 - Determine / interpret PZR level indicators 3.7 1 000009 Small Break LOCA / 3 EA2.0. - Action to take duing SBLOCA, based on temp and 4.1
__ _ _ _ _ _ _ _ __ ipressure
W/E03 LOCA Cooldown - Depress. /4 (PSA) X 2A.22 - Bases for prioritizing safety functions in EOPs 4.0 W/E 11 Loss of Emergency Coolant Recirc. / 4 (PSA) IA..02 -Adhereence to procedures and operations within limits 42
X..... . license
,.in during Loss of Emergency Coolant Recire
000022 Loss of Reactor Coolant Makeup / 2 X 2.1.25 - Obtain / interpret station reference materials such as 3.1 1
graphs, monographs, and tables.
000025 Loss of R-R System / 4 X AK2.05 - Interrelationship between Loss of RHR and reactor 2.6 1
building sump
000027 Pressurizer Pressure Control System X AA2.15 - Actions if PCS instrument fails high 4.0 1
Malfunction / 3
000032 Loss of Source Range NI / 7 000033 Loss of Intermediate Range NI /.7...4 - aecogniabnr t indications for pa ters forienty- 43
____________, _ level conditionsi* foremergency :abnormal-procedures. i
000037 Steam Generator Tube Leak / 3 X AA1. 13 - Operate / monitor SGBD rad monitor during tube leak 4.0 000038 Steam Generator Tube Rupture /3 X EK3.08 - Criteria for securing RCPs during a SGTR 4.2 1 000054 Loss of Main Feedwater / 4 (PSA) > Z4.......O ej......
t.ti.n.herarch oord.in.tipn.with ,1 40
Ssupp...... ..
W/E05 Inadequate Heat Transfer - Loss of Secondary X EK2.01 - Components, control safety systems, including auto 3.9 1
Heat Sink / 4 manual operation related to Loss of Heat Sink
000058 Loss of DC Power / 6 X AA2.03 - Determine / interpret DC loads lost; impact on ability 3.9 1
to operate / monitor plant systems
000060 Accidental Gaseous Radwaste Rel. / 9
000061 ARM System Alarms / 7
W/E16 High Containment Radiation /9 : ] EA2. Sqeetion ofprocedures during resp6oseto Hig 3.3
X CntinmntRadiation,___
000065 Loss of Instrument Air / 8 X AA 1.02 - Operate / monitor components served by IAS to 2.8 1
minimize drain on system
1K/A Category Totals: I 11 2 I 1 1 2 1 I6lGroup 4 Point Total: - 16
NUREG-1021, Revision 8, Supplement 1
( ( ( ES-401 PWR SRO Examination Outline Form ES-401-3
I_ Emergency and Abnormal Plant Evolutions - Tier 1/Group 3
E/APE # / Name / Safety Function KL K,2 K3 Al A2 G K/A Topic(s) Imp. Points
000028 Pressurizer Level Malfunction /2 AA2.13. Det.nineactualPZR leve, giventmcompensated ..
000036 Fuel Handling Accident / 22228 Knowledge .. ofnewand.spent. fe.ov.ement.procedures 3 .5
000056 Loss of Off-site Power / 6 X AA2.22 - Determine / interpret lube oil pump indicators and low 3.6 1
pressure alarms on EDG
W/E13 Steam Generator Over-pressure / 4
W/E 15 Containment Flooding / 5
[K/A Category Totals: ]0101 0II2Jil GroupPointTotal: 3]
NUREG-1021, Revision 8, Supplement I
( ( (
ES-401 PWR SRO Examination Outline Form ES-401
Plant Systems - Tier 2/Group I
System # / Name KI K2 K3 K4 K5 K6 Al A2 A3 A4 G K/A Topic(s) IMp. Points
001 Control Rod Drive X K4.11 - Design for resetting CRDM ckt breakers 2.9 2
X K5.42 - Definition of Tave and no load Tave 3.0
003 Reactor Coolant Pump X A4.08 - Operate / monitor RCP cooling supplies 2.9 1 004 Chemical and Volume Control X K3.08 - Effect of loss of CVCS on RCP seal 3.8 2
injection
X A1.06 - Monitor changes in VCT level 3.2
013 Engineered Safety Features Actuation X K6.01 - Effect of loss of sensors / detectors on 3.1 1
ESFAS
014 Rod Position Indication X 2.1.11 - Less than one hour tech spec actions 3.8 1 015 Nuclear Instrumentation X K5.06 - Implication of subcritical multiplication on 3.7 1
I_ INIS
017 In-core Temperature Monitor X A3.01 - Incore temp indications of normal, natural, 3.8 1
Iand interrupted circulation of RCS
022 Containment Cooling X K2.01 - Power supplies to containment cooling 3.1 2
X K4.01 - Design for cooling of containment 3.0
__enetrations
026 Containment Spray X A2.08 - Determination of when spray can be 3.7 1
Xsecured
056 Condensate X A2.04 - Predict impacts of loss of condensate 2.8 1
pvumps 059 Main Feedwater X K1.02 - Cause / effect between MFW and AFW 3.4 1
061 Auxiliary/Emergency Feedwater (PSA) X K6.01 - Effect of loss of controllers / positioners on 2.8 1
AFW
063 DC Electrical Distribution X K2.01 - Power to major DC loads 3.1 1 068 Liquid Radwaste X 2.1.32 - Explain / apply system limits and 3.8 1
orecautions
071 Waste Gas Disposal X 2.2.25 - Bases in tech specs for LCOs and safety 3.7 1
limits
072 Area Radiation Monitoring X K3.01 - Effect of loss of ARM on containment vent 3.4 1
1 isolation
[K/A Category Totals: i 1 2 12 2[ 2 [ 221 1 1 3 [1Group Point Total: - 19 NUREG- 1021, Revision 8, Supplement 1
( (
ES-401 PWR SRO Examination Outline Form ES-4O1-*31
Plant Systems - Tier 2/Grou1p 2
System # / Name K1 K2 K3 K4 K5 K6 Al A2 A3 A4 G K/A Topic(s) Imp. Points
002 Reactor Coolant X K6.03 - Effect of loss on RVLIS 3.6 1 006 Emergency Core Cooling X A4.08 - Operate / monitor ESF, including reset 4.3 1 010 Pressurizer Pressure Control X K1.06 - Cause / effect between PCS and CVCS 3.1 1 011 Pressurizer Level Control X K5.06 - Indicated charging flow - seal flow plus 3.2 1
actual charging flow
012 Reactor Protection X K11.08 - Cause / effect between RPS and MFW 3.1 1 0 16 Non-nuclear Instrumentation X K4.03 - Design of input to control systems 2.9 1 027 Containment Iodine Removal 028 Hydrogen Recombiner and Purge Control 029 Containment Purge XK4.02 - Design regarding negative pressure in 3.1 1
containment
033 Spent Fuel Pool Cooling A2.03 - Predict impact of abnormal spent fuel pool 3.5 1
water level or loss of level
034 Fuel Handling Equipment 035 Steam Generator X IA3.01 - Monitor SG water level control 3.9 1 039 Main and Reheat Steam X A1.05 - Predict effect of changes on RCS Tave 3.3 1 055 Condenser Air Removal 1K3.01 - Effect of loss of CARS on main condenser 2.7 1 062 AC Electrical Distribution X 2.1.27 - Knowledge of system purpose / function 2.9 1 064 Emergency Diesel Generator X K3.03 - Effect of loss of EDG on manual loads 3.9 1 073 Process Radiation Monitoringg X A4.02 - Operate / monitor rad monitor panel 3.7 1 075 Circulating Water X 2.1.25 - Obtain / interpret station reference 3.1 1
materials such as graphs and tables
079 Station Air 086 Fire Protection X A 1.01 - Predict effect of changes on fire header 3.3 1
1___ 1 pressure
103 Containment X A4.06 - Operate / monitor containment personnel 2.9
airlock door
JK/A Category Totals: 2 0 2 1 2 i1 i 1 2i 1 1 1 1 3 1 2 1Group Point Total: 17 NUREG- 1021, Revision 8, Supplement 1
( ( K?
ES-401 PWR SRO Examination Outline Form ES-40 1-3
Plant Systems - Tier 2 /Group 3
System # / Name K1 K2 K3 K4 K5 K6 Al A2 A3 A4 G K/A Topic(s) Imp. Points
ý005Ksidual Heat Removal 1(4.08 - Lineup for piggy back mode with HP 3.5
injection
007 Pressurizer Relief/Quench Tank 008 Component Cooling Water X K2.02 - Power to CCW pumps, including 3.2 1
emergency backup
041 Steam Dump/Turbine Bypass Control 045 Main Turbine Generator X A1.06 - Expected response of secondary plant 3.7 1
parameters followina TG trip
076 Service Water 078 Instrument Air 1K1.03 - Cause / effect between IAS and 3.4 1
x containment air
.K/ACategory Totals: 1I 1 0 1 0 0 1 [ o00 0 0 [Group PointTotal: [ 4
Plant-Specific Priorities
System/Topic Recommended Replacement for ... Reason Points
[Plant-Specific Priority Total: (limit 10) [ NUREG- 1021, Revision 8, Supplement 1
ES-401 Generic Knowledge and Abilities Outline (Tier 3) Form ES-401-5
Facility: HARRIS Date of Exam: 08/26/2002 Exam Level: SRO
Category K/A # Topic Imp. Points
.1.33 eco~ ze 1nd icat~ons .. yst n 6peratixg param eters*vh chcar ' 4.0
2.1.25 Obtain / interpret station reference materials such as graphs, 3.1 1
monographs, and tables.
X..d go f conditio s and l1... .t ..... in i .. .. .. . ........... .1
Conduct of Kf0
2.1.2 Knowledge of operator responsibilities during all modes of 4.0 1
Operations operations.
:21.4 :s inr'iMts 1..4 1.
Total
o2l.g ass t se I aCOs
and'ai..... ii 37 1
2.2.24 Analyze the effect of maintenance activities on LCO status. 3.8 1
Knowledge of tagging and clearance procedures. 3.8 1
Equipment
Equpontro
Control 72.2.1326
!:2:.... Knowiagt*re~ljh~drntlrisf~tiye &e4dfremti&ts* 37, 1
Total 4
:2.134 i'ýonfridil, ,nluln 1p
3f,'fnfi
: :. ... p im ssi ujn q . . .... . ...
2.3.10 Perform procedures to reduce excessive levels of radiation and 3.3 t
Radiation guard against personnel exposure.
Ability to control radiation releases. 3.2 1
Control 2.3.11
Total 3
[::421 WJid"o' "0 i"" t6.asi ht ...s Of 4.3 1
2.4.2 System setpoints, interlocks, and automatic actions associated with 4.1 1
EOP entry conditions.
Knowledge of event based EOP mitigation strategies. 3.8 1
Emergency 2.4.7
..o6 g o........ .... . ....
Procedures/ NR':
Plan . .. k 42i&o'M1 l idJs t 4Af erina pse ii°Pn~gameters ..
S:whcb&?~jilon*foS ere~i& anin*n¢*[
Total 5
Tier 3 Point Total [ 17
NUREG- 1021, Revision 8, Supplement I
ES-401 PWR RO Examination Outline Form ES-401-4 kacility: HARRIS Date of Exam: 8/26/2002 Exam Level: RO
K/A Category Points
Tier Group K K K K K K A A A A G Point
1 2 3 4 5 6 1 2 3 4 * Total
1.3 2 16
Emergency& 2 3 2 3 3 4 2 17
Abnormal Plant 0 0 0 2 1 0 3
Evolutions Tier 5 8 8 .. .43
Totals 4 8 3
2 1 2 3 2 2 2 2 2 2 2 2 2 23
Plant 2 2 1 2 2 1 1 2 2 1 3 3 20
Systems 3 1 1 0 1 0 0 1 0 0 2 2 8
Tier
Toa 5 5 4 5 3 3 5 4 3 7 7 51
Tota--s
3 Generic Knowledge and Abilities Cat 1 Cat 2 Cat 3 Cat 4
4 3 3 3 13
Notes:
I Ensure that at least two topics from every K/A category are sampled within each tier
(i.e., the "Tier Totals" in each K/A category shall not be less than two).
2 The point total for each group and tier in the proposed outline must match that specified in the table.
The final point total for each group and tier may deviate by +/- I from that specified in the table based
on NRC revisions. The final exam must total 100 points.
3 Select topics from many systems; avoid selecting more than two or three K/A topics from a
given system unless they relate to plant-specific priorities.
4 Systems/evolutions within each group are identified on the associated outline.
5 The shaded areas are not applicable to the category/tier.
6* The generic K/As in Tiers I and 2 shall be selected from Section 2 of the K/A Catalog, but the
topics must be relevant to the applicable evolution or system.
7 On the following pages, enter the K/A numbers, a brief description of each topic, the topics'
importance ratings for the RO license level, and the point totals for each system and category.
K/As below 2.5 should be justified on the basis of plant-specific priorities. Enter the tier totals
for each category in the table above.
8 Shaded topics on individual Tier / Group pages indicate topic is RO ONLY.
NUREG- 1021, Revision 8, Supplement I
( (1 (1
ES-401 PWR RO Examination Outline Form ES-401-41
Emergency and Abnormal Plant Evolutions - Tier 1/Group 1
E/APE # / Name / Safety Function KI K2 K3 Al A2 G K/A Topic(s) Imp. Points
000005 Inoperable/Stuck Control Rod / 1 X . ... AKI 02 Implications of inoperable / stuck rod on flux tilt 3.1 1 000015/17 RCP Malfunctions / 4 X AA1.08 - Operate / monitor SG LCS during loss of RCS flow 3.0 1 W/E09&EIO Natural Ciro. / 4 EKL.02 - Normal, abnormal, and emergency procedures 3.3 1
X associated with NC operations
000024 Emergency Boration I 1oratioo
1AK2.01 - Interrelationshipbetwelen Emergency and 2
x valVes
000026 Loss of Component Cooling Water / 8 X AK3.01 - Conditions that will initiate automatic operation of 3.2 1
SWS valves to CCW coolers
000027 Pressurizer Pressure Control System X AA2.15 - Actions if PCS instrument fails high 3.7 1 Malfunction / 3 000040 (W/E 12) Steam Line Rupture - Excessive Heat X 000040AA1.22 - Operate / monitor load sequencer status lights 3.0 1 Transfer / 4 during a steam line rapture
W/EO8 RCS Overcooling - PTS / 4 X EK2.02 - Heat removal systems, and relationship between proper 3.6 1
operation of systems during PTS event
000051 Loss of Condenser Vacuum /4 301A - toss ofeam p cafcndeser 21
vacuum 000055 Station Blackout /6 X 2.4.18 - Knowledge of specific bases for EOPs 2.7 1
000057 Loss of Vital AC Elec. Inst. Bus! 6 X AA2.15 - Determine / interpret that a loss of AC has occurred 3.8 1 000062 Loss of Nuclear Service Water / 4 X AA2.06 - Length of time after loss of SWS flow to component 2.8 1
until damage to component occurs
000067 Plant Fire On-site / 9
000068 Control Room Evac. / 8 X AK3.09 - Transfer to local control - charging pumps, flow 3.9 1
control, PZR heaters, and boric acid transfer pumps
000069 (W/E14) Loss of CTMT Integrity / 5 W 4EIK3 , Normal -abnormal an eergen1cy proedps
associated ihhg oianetpesr
000074 (W/E06&E07) Inad. Core Cooling / 4 (PSA) X 000074 EA1.05 - Operate / monitor PZR PORV during ICC 3.9 1
000076 High Reactor Coolant Activity / 9 X 2.4.18 - Knowledge of specific bases for EOPs 2.7 1
K/A Category Totals: L2 2 41 3[3I 2IGroup Point Total: 116
NUREG- 1021, Revision 8, Supplement 1
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ES-401 PWR RO Examination Outline Form ES-401
Emergency and Abnormal Plant Evolutions - Tier I/Group 2
Kl E2 K3 Al A2 G K/A Topic(s) Imp. Points
E!APE # / Name /Safety Function
X AK 1.16 - Definition / application of power defect 3.0 1
000001 Continuous Rod Withdrawal/ 1 000003 Dropped Control Rod / 1 X 2.4.6 - Knowledge of symptom based EOP strategy 3.1 1
EK1.03 - Reasons for isolating the TG after a reactor trip 3.7 1
000007 Reactor Trip - Stabilization - Recovery / 1 X 000008 Pressurizer Vapor Space Accident / 3 X AA2.12 - Determine / interpret PZR level indicators 3.4 1 000009 Small Break LOCA / 3 _ EK3.20 - Tech speeRCS leGaage limits . 3f5 000011 Large Break LOCA / 3 (PSA) EA1.01 - Control of RCS pressure and temp to avoid violating 3.7 1
0PTS limits during a LBLOCA
W/E04 LOCA Outside Containment / 3
4
Emergency Coolant/ Recirc.4
Loss ofCooldown/Depress.
W/E11 LOCA W/E03 W/EO I & E02 Rediagnosis & SI Termination/3 /EA 03 - mpsaentycsys'°msfeincudiant0,
man*al operationl reate jt: SI Terminatio*n
X 2.1.25 - Obtain / interpret station reference materials such as 2.8 1
000022 Loss of Reactor Coolant Makeup / 2
graphs, monographs, and tables.
X AK2.05 - Interrelationship between Loss of RHR and reactor 2.6 1
00025 Loss of RHR System / 4
building SUmp
X EK3.11 - Reasons for initiating Emergency Boration 4.2 1
000029 Anticipated Transient w/o Scram / 1 000032 Loss of Source R ange NI / 7 ........ ............. . ........... ..... ....._ .._
t
000033 Loss of Intermediate Range NI / 7 ____ > ) (KI'
X AA1.13 - Operate / monitor SGBD rad monitor during tube leak 3.9 1
000037 Steam Generator Tube Leak / 3
X EK3.08 - Criteria for securing RCPs during a SGTR 4.1 1
000038 Steam Generator Tube Rupture / 3
000054 Loss of Main Feedwater / 4 (PSA)
X EK2.01 - Components, control / safety systems, including auto 3.7 1
W/E05 Inadequate Heat Transfer - Loss of Secondary
Heat Sink / 4 manual operation related to Loss of Heat Sink
X AA2.03 - Determine / interpret DC loads lost; impact on ability 3.5 1
000058 Loss of DC Power / 6
to operate / monitor plant systems
X AA2.03 - Failure modes, symptoms, and causes of misleading 3.1 1
000059 Accidental Liquid RadWaste Rel. / 9
indications on liquid rad monitor
.6 Determine -interpretv for release of mcI 3.6
000060 Accidental Gaseous Radwaste Rel. / 9
000061 ARM System Alarms/ 7
W/E 16 High Containment Radiation /9
K/A Categor Totals: 13 2 3 3 4 2 Group Point Total: 17
NUREG- 1021, Revision 8, Supplement 1
/*
(
ES-401 PWR RO Examination Outline Form ES-401-4
Emergency and Abnormal Plant Evolutions - Tier 1/Group 3
E/APE # / Name / Safety Function KI K2 K3 Al A2 G K/A Topic(s) Imp. Points
000028 Pressurizer Level Malfunction / 2 000036 Fuel Handling Accident / 8 AA* .04 - Oprate / monitor fuel handling equipment during 3.1 1
I
..incident
000056 Loss of Off-site Power / 6 X AA2.22 - Determine / interpret lube oil pump indicators and low 3.4 1
pressure alarms on EDG
000065 Loss of Instrument Air / 8 X AA 1.02 - Operate / monitor components served by IAS to 2.6 1
minimize drain on system
W/E 13 Steam Generator Over-pressure /4 W/E 15 Containment Flooding / 5
1-1-1-1-1-1-4 I I
t. I
I-F-I-I-I-I-I 1 I
I-I-f -f-f-I-I t I
I I
4-4-4-i-I-I-f I t
t I
I T
I-4-4-4-{-+-t 1 T
1-1-1-1-4-4-4 I I
I-I-I-I-I-I-i I I
K/A Category Totals: 1 0 1 0 1 0 1 2 1 1 0 1Group Point Total: I 3
NUREG-1021, Revision 8, Supplement 1
( C (
PWR RO Examination Outline Bonn ES-401-4
ES-401
Plant Systems - Tier 2/Grou 1
K1 K2 K3 1(4 K5 K6 Al A2 A3 A4 G K/A Topic(s) Imp. Points
System # / Name
X. K2.02 - One line diagramofPS to trip breakers 3.6 3
001 Control Rod Drive
X K4.11 - Design for resetting CRDM ckt breakers 2.7
X K5.42 - Definition of Tave and no load Tave 2.9
XX----1.03 -Cause! [ffect between RCP and RCP seas 3.3 2
003 Reactor Coolant Pump
X A4.08 - Operate / monitor RCP cooling supplies 3.2
K3.08 - Effect of loss of CVCS on RCP seal 3.6 3
004 Chemical and Volume Control VX iniection
X A1.06 - Monitor changes in VCT level 3.0
X A3.08 - Monitor CVCS changes on reactor power 3.9
XK6.01 - Effect of loss of sensors / detectors on 2.7 2
013 Engineered Safety Features Actuation
ESFAS
K
"'X, A101- Predict impacts of LOCA on ESFAS 4
K5.06 - Implication of subcritical multiplication on 3.4 2
0K15 Nuclear Instrumentation
NIS
Ar3. Operatemotr NlS'trip bypasses: .
A3.01 - Incore temp indications of normal, natural, 3.6
017 In-core Temperature Monitor
and interrupted circulation of RCS
022 Containment Cooling X I--(2.01 - Power supplies to containment cooling 3.0 2
K4.01 - Design for cooling of containment 2.5
X
p1enetrations
056 Condensate A2.04 - Predict impacts of loss of condensate 2.6 1
pumps
K1.02 - Cause effect between MFW and AFW 3.4 2
059 Main Feedwater X
( ýn n, .........
Al...3.-Power.leel ........ .restrctions
. ... fr opiraton of... . . . 21
___MWpulmps Iand valves
n X2102 - Power.supplieo AW electric pumps ,3.7 2
061 Auxiliary/Emergency Feedwater
XK6.01 - Effect of loss of controllers / positioners on 2.5
(PSA)
AFW
068 Liquid Radwaste 2.1.32 - Explain / apply system limits and 3.4 1
precautions
X 2.2.25 - Bases in tech specs for LCOs and safety 2.5 1
071 Waste Gas Disposal
lim its 3 1
- Effect of loss of ARM on containment vent 3.2 t
A Radiation Monitoring---K3.01
I072 Area I isolation
212 12121212 1 2 1 2 IGroup PointTotal: 23 1
K/A Category Totals: j21 3J2
NUREG- 1021, Revision 8, Supplement 1
( ( (
ES-401 PWR RO Examination Outline Form ES-401-4
Plant Systems - Tier 2/Group 2
System # / Name Ki K2 K3 K4 K5 K6 Al A2 A3 A4 G K/A Topic(s) m Points
002 Reactor Coolant X K6.03 - Effect of loss on RVLIS 3.1 1 006 Emergency Core Cooling X A4.08 - Operate / monitor ESF, including reset 4.2 1 010 Pressurizer Pressure Control X- K1.06 - Cause / effect between PCS and CVCS 2.9 1 011 Pressurizer Level Control X K5.06 - Indicated charging flow - seal flow plus 2.9 1
actual charging flow
012 Reactor Protection X K1.08 - Cause / effect between RPS and MFW 2.9 1 014 Rod Position Indication X 2.1.11 - Less than one hour tech spec actions 3.0 1 016 Non-nuclear Instrumentation X K4.03 - Design of input to control systems 2.8 1 026 Containment Spray X A2.08 - Determination of when spray can be 3.2 1
secured
029 Containment Purge X K4.02 - Design regarding negative pressure in 2.9 1
containment
033 Spent Fuel Pool Cooling X A2.03 - Predict impact of abnormal spent fuel pool 3.1 1
water level or loss of level
035 Steam Generator X I A3.01 - Monitor SG water level control 4.0 1 1 039 Main and Reheat Steam X A 1.05 - Predict effect of changes on RCS Tave 3.2 1 055 Condenser Air Removal X K3.01 - Effect of loss of CARS on main condenser 2.5 1 062 AC Electrical Distribution X 2.1.27 - Knowledge of system purpose / function 2.8 1 063 DC Electrical Distribution X K2.01 - Power to major DC loads 2.9 1 064 Emergency Diesel Generator X K3.03 - Effect of loss of EDG on manual loads 3.6 1 073 Process Radiation Monitoring X A4.02 - Operate / monitor rad monitor panel 3.7 1
075 Circulating Water X 2.1.25 - Obtain / interpret station reference 2.8 1
materials such as graphs and tables
079 Station Air A....._-
__ / moitior-cr0s__ie it IAS'_7
086 Fire Protection X A1.01 - Predict effect of changes on fire header 2.9 1
pressure
K/A Category Totals: 2 1j 2j 2 1 [ 121 1 3 3 3j2 [Group Point Total: 20
NUREG-1021, Revision 8, Supplement 1
( ( (
ES-401 PWR RO Examination Outline Form ES-401-4j
SPlant Systems - Tier 2/Group 3
K1 K2 K3 K4 K5 K6 Al A2 A3 A4 G K/A Topic(s) Imp. Points
System # / Name
X K4.08 - Lineup for piggy back mode with HP 3.1 1
005 Residual Heat Removal
injection
X A4.04 Operate / monitor PZR vent valve 2.6 1
007 Pressurizer Relief/Quench Tank -
X K2.02 - Power to CCW pumps, including 3.0 1
008 Component Cooling Water
emergency backup
027 Containment Iodine Removal 028 Hydrogen Recombiner and Purge I I.....2i* K~[g
- Knowlege of system purpose I function 2.8 1
028HyroenReominr2.1Puge27
Control 034 Fuel Handling Equipment
Bypass Control ___________________
041 Steam Dump/Turbine
C A1.06 - Expected response of secondary plant 3.3 1
045 Main Turbine Generator
parameters following TG trip
076 Service Water X..4...ow geo f a
-- K1.03 - Cause/ effect between IAS and 3.3 1
078 Instrument Air-
nX containment air
X-A4.06 - Operate / monitor containment personnel 2.7 1
103 Containment X airlock door
1 i o10 i1 0 0o 1 0 2 2 Group PointTotal: 8
[./ACategory Totals:
Plant-Specific Priorities
System/Topic Recommended Replacement for ... Reason Points
jPlant-Specific Priority Total: (limit 10)
NUREG- 1021, Revision 8, Supplement 1
ES-401 Generic Knowledge and Abilities Outline (Tier 3) Form ES-401-5
Facility: HARRIS Date of Exam: 08/26/2002 Exam Level: RO
K/A # Topic Imp. Points
Category
Z 1.29 Ky v alv*e* ups.3t4
2.1.25 Obtain / interpret station reference materials such as graphs, 2.8 1
monographs, and tables.
Conduct of 2b x ta4
2.1.2 Knowledge of operator responsibilities during all modes of 3.0 1
Operations
operations.
Total 4
2,212 K1owled1gedfsiilagi~ropdtres 3.00
2.2.24 Analyze the effect of maintenance activities on LCO status. 2.6 1
3.6 1
Equipment 2.2.13 Knowledge of tagging and clearance procedures.
Control
Total 3
2.3.10 Perform procedures to reduce excessive levels of radiation and 2.9 1
guard against personnel exposure.
2.7 1
Radiation 2.3.11 Ability to control radiation releases.
Control
Total 3
Plol
2.4.2 System setpoints, interlocks, and automatic actions associated with 3.9 1
EOP entry conditions.
3.1 1
Emergency 2.4.7 Knowledge of event based EOP mitigation strategies.
Procedures/
Plan
Total 3
Tier 3 Point Total
J 13
NUREG- 1021, Revision 8, Supplement I
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