Text

August 2002 Exam 50-400/2002-301 Draft Ro/Sro Written Exam & Outlines
	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)
	v • d • e

{{#Wiki_filter:Draft Submittal (Pink Paper) Reactor Operator Written Exam CQ.JD Qjp~z~arQ~0f (2 5) 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 d. 0 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 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. The plant is to be placed in Hot Standby with T-avg less than 5000F 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. b. 5.63 6.25 6.88 7.50 C. d. 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 c. Until Jacket Water Cooler Outlet temperature exceeds 1850F d. Until Lube Oil Cooler Outlet temperature exceeds 1850F ANSWER: a. One (1) minute

Harris Nuclear Plant Common Written Questions QUESTION: 19 The plant is operating at 100% power with the following conditions: Ambient Temp 35 OF 20 OF 10 OF CT Basin Temp 64 OF 60 OF 58 OF Which of the following conditions? 1600 a. Full Open b. Full Open c. Half Open d. Half Open ANSWER: b. Full Open describes the correct CT Deicing Gate Valve alignment for these 2000 Full Open Half Open Full Open Half Open Half Open Time 1200 1600 2000

Harris Nuclear Plant Common Written Questions 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% QUESTION:

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? a. Rods initially step in at 8 steps per minute to reduce Tavg to 5530F b. Rods initially step in at 8 steps per minute to reduce Tavg to 5570F c. Rods initially step in at 72 steps per minute to reduce Tavg to 5530F d. Rods initially step in at 72 steps per minute to reduce Tavg to 5570F ANSWER: d. Rods initially step in at 72 steps per minute to reduce Tavg to 5570F

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 HIGH AH-2A LIT CNMT FAN HIGH AH-2B OFF CNMT FAN LOW AH-2A OFF CNMT FAN LOW AH-2B OFF SW BSTR PUMP START A 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 HIGH AH-2A CNMT FAN HIGH AH-2B CNMT FAN LOW AH-2A CNMT FAN LOW AH-2B SW BSTR PUMP START A a. LIT b. LIT C. OFF d. OFF ANSWER: c. OFF LIT LIT OFF OFF OFF OFF LIT LIT OFF OFF OFF OFF LIT OFF LIT 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. 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 H2 B8 H14 P8 Group 2 Rod F6 F10 K10 K6 Steam 186 186 192 180 Steps 186 198 186 180 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 S

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. "Fuel Handling Accident." b. 18 feet above the fuel assemblies. "Fuel Pool Cooling System." c. 21 feet above the fuel assemblies. "Fuel Handling Accident." d. 21 feet above the fuel assemblies. "Fuel Pool Cooling System." ANSWER: b. 18 feet above the fuel assemblies. "Fuel Pool Cooling System." Makeup should be initiated using AOP-0 13, Makeup should be initiated using OP- 116, Makeup should be initiated using AOP-013, Makeup should be initiated using OP- 116, Makeup should be initiated using OP-1 16,

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 'A' and C' RCPs are running. 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. 5530F. b. 5570F. c. 562°F. d. 5640F. ANSWER: c. 5620F.

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 b. e SG pressure INCREASES

Turbine First Stage Impulse Pressure

SG pressure DECREASES Turbine First Stage Impulse d. e SG pressure DECREASES

Turbine First Stage Impulse

ANSWER: b. o SG pressure INCREASES INCREASES DECREASES Pressure INCREASES Pressure DECREASES

Turbine First Stage Impulse Pressure DECREASES

C. S

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:

During a reactor startup, power has been stabilized at 10-8 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: A loss of offsite power has occurred. Both Emergency Diesel Generators are loaded. ALB-024-3-2, DIESEL GENERATOR A TROUBLE, alarms. An operator is sent to investigate and reports the following conditions: Turbo Oil Press Lube Oil Press Fuel Oil Press Day Tank Level Starting Air Pressure Jacket Water Pressure Control Air Pressure 28 psig and stable 30 psig and stable 1.5 psig and stable 56% and slowly decreasing 227 psig and slowly decreasing 17 psig and stable 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 0 0 0 S S S S S S S

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 a. Increasing b. Increasing C. Decreasing d. Decreasing ANSWER: b. Increasing OF PORV 613 0F 250°F 613 0F 2500F 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

CCW Heat Exchanger outlet temperature is 950F
ALB-5-1-2B, RCP THERM BAR HDR LOW FLOW, is NOT alarming

b.

RCS is currently at 3750F 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 5250F 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 1220F

0 ALB-5-1-2B, RCP THERM BAR HDR LOW FLOW, is alarming ANSWER: b. e RCS is currently at 3750F 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 6240F 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 draining and the valve is located in an area where the temperature is 134 0F 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 2300F 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 >14000F.
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: Given the following conditions: . The unit is in the Source Range

Power is lost to Instrument Bus

during S-rnI. a reactor startup.

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 70

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 600F 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 940F
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 2200F.

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. b. C. the reactor will trip due to a shunt trip of Train SA reactor trip breaker. 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 1 CS-50, Letdown to VCT/Demin? a. Automatically bypasses the CVCS Demnineralizers when TE-143, LP Letdown Temperature, exceeds 1350F to prevent an inadvertent dilution event b. Automatically bypasses the CVCS Demineralizers when TE-143, LP Letdown Temperature, exceeds 1350F 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 1350F 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? today for the today for the plant shutdown over-exposure b.

0745
0700

C. 1130 0800 today for the plant shutdown tomorrow for the over-exposure today for the plant shutdown today for the over-exposure d.

1130 today for the plant shutdown

0700 tomorrow for the over-exposure

ANSWER: d.

1130

0700

today for the plant shutdown tomorrow for the over-exposure a. 00745 0800

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 7050F and 7200F. 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 A B C LEVEL 81% 68% 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 4900F.

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. 4500F 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 1 A-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 BORON CONC. CONT BANK 'C' POSTION CONT BANK 'D' POSTION ECC - 500 PCM POSITION ECC + 500 PCM POSITION ROD INSERTION LIMIT 1830 1215 ppm 218 steps 90 steps 45 steps on Bank 'D' 197 steps on Bank 'D' 0 steps on Bank 'D' The Actual Critical Conditions are as follows: TIME BORON CONC. CONT BANK 'C' POSTION CONT BANK 'D' POSTION 1836 1198 ppm 110 steps 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- 1 A21 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 550F 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 7 3 4 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 4 5 40 Totals 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 ( ES-401 PWR SRO Examination Outline Form ES-401-31 S-401 _Emergency and Abnormal Plant Evolutions - Tier 1/Group 1 E/APE # / Name / Safety Function K1 K2 K3 Al A2 G K/A Topic(s) Imp. Points 000001 Continuous Rod Withdrawal / 1 X AK 1.16 - Definition / application of power defect 3.4 1 000003 Dropped Control Rod / 1 X 2.4.6 - Knowledge of symptom based EOP strategy 4.0 1 000005 Inoperable/Stuck Control Rod / 1 X- AA2.O3 - Actions ifmore than one rod is stucki inoperable 4.4 1 000011 Large Break LOCAl 3 (PSA) EA1.01 - Control of RCS pressure and temp to avoid violating 3.8 1 LX PTS limits during a LBLOCA W/E04 LOCA Outside Containment / 3 EA2.01 - on of poceduesduring response to LOCA Xoutside containment, W/EO 1 & E02 Rediagnosis & SI Termination 3 X EA2.02 - Adherence to procedures and operations within limits 4ý0 in license during SI Termination 000015/17 RCP Malffinctions / 4 X AA1.08 - Operate / monitor SG LCS during loss of RCS flow 2.9 1 WE09&E10 Natural Cue. / 4 XEK1.02 - Normal, abnormal, and emergency procedures 3.7 X associated with NC operations 000024 Emergency Boration/1 X 2.20 AbIlitt execbteprocedurestep 4.2 1 000026 Loss of Component Cooling Water 8 S-----------------AI3.0l - Conditions that will initiate automatic operation of 3.5 1 1 L X SWS valves to CCW coolers 000029 Anticipated Transient w/o Scram / 1 X EK3.11 - Reasons for initiating Emergency Boration 4.3 1 000040 (W/E12) Steam Line Rupture - Excessive Heat X_-000040AA1.22 - Operate / monitor load sequencer status lights 3.0 1 000ransfer / 4 during a steam line rupture /EO8 RCS Overcooling - PTS /4 XEK2.02 - Heat removal systems, and relationship between proper 4.0 1 /- operation of systems during PTS event 000051 Loss of Condenser Vacuum / 4 Xl AA2*ý Loissofvacuum reqiiring actor I turbine trip :4 A 1 EK0(X2. 4.1 - He t eo vale s sesdn elto si.ewe np o e . 000055 Station Blackout / 6 X 2.4.18-Knowledge of specific bases for EOPs 3.6 1 000057 Loss of Vital AC Elec. Inst. Bus / 6 X AA2.15 - Determine / interpret that a loss of AC has occurred 4.1 1 000059 Accidental Liquid RadWaste Rel. / 9 X AA2.03 - Failure modes, symptoms, and causes of misleading 36 1 1tindications on liquid rad monitor R 00062 Loss of Nuclear Service Water 4 X AA2.06 - Length of time after loss of SWS flow to component 3.1 1 Iuntil damage to component occurs 0007Pan ieOnstX AA2.13 -Fire60 codtosrqdrn mrecy plant shutdown::ý 4A.4 I 000068_____Control____Room________

--- --- AK3 .09 - Transfer to local control - charging pumps, flow 4.4 1 0 C l E / control, PZR heaters, and boric acid transfer pumps 000069 (W/E14) Loss of CTMT Integrity /5 i2.310l Perfomprocedure to reduce ivexlevelsoUof 3U3 r adion a guard -aaist personnel exposure 000074 (W/E06&E07) Inad. Core Cooling / 4 (PSA) X 000074 EAI.05 - Operate / monitor PZR PORV during ICC 4.1 2 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..... . ,.in license 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) ej > Z4.......O ...... t.ti.n. herarch oord.in.tipn.with 40 ,1 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 6 I 4 lGroup 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 I and 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 21 [ 2 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 A 1.05 - 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 2 i 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 ý005K sidual 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/A Category Totals: 1I 1 0 1 0 0 1 [ o0 0 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 chc ar .. ' 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 Equipment 2.2.13 Knowledge of tagging and clearance procedures. 3.8 1 Equpontro 7 Control !:2:.... 26 Knowiagt*re~ljh~drntlrisf~tiye &e4dfremti&ts* 37, 1 Total 4

2.134

i 'ýonfridil, ,nluln 3f,'fnfi 1p

:. ...

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. Control 2.3.11 Ability to control radiation releases. 3.2 1 Total 3 [::421 W Jid"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. Emergency 2.4.7 Knowledge of event based EOP mitigation strategies. 3.8 1 ..o 6 g o........ .. .. . .... Procedures/ NR': Plan . .. k 42i&o'M1 l idJs t er a 4Af in pse ii°P n~ 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 8 4 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 1AK2.01 - Interrelationshipbetwelen Emergency 1oratioo 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

( ( C ES-401 PWR RO Examination Outline Form ES-401 Emergency and Abnormal Plant Evolutions - Tier I/Group 2 E!APE # / Name /Safety Function Kl E2 K3 Al A2 G K/A Topic(s) Imp. Points 000001 Continuous Rod Withdrawal/ 1 X AK 1.16 - Definition / application of power defect 3.0 1 000003 Dropped Control Rod / 1 X 2.4.6 - Knowledge of symptom based EOP strategy 3.1 1 000007 Reactor Trip - Stabilization - Recovery / 1 X EK1.03 - Reasons for isolating the TG after a reactor trip 3.7 1 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 spe eRCS 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 W/E03 LOCA Cooldown/Depress. / 4 W/E11 Loss of Emergency Coolant Recirc.4 W/EO I & E02 Rediagnosis & SI Termination/3 /EA 03 - mpsaentycsys'°msfeincudiant0, man*al operationl reate jt: SI Terminatio*n 000022 Loss of Reactor Coolant Makeup / 2 X 2.1.25 - Obtain / interpret station reference materials such as 2.8 1 graphs, monographs, and tables. 00025 Loss of RHR System / 4 X AK2.05 - Interrelationship between Loss of RHR and reactor 2.6 1 building SUmp 000029 Anticipated Transient w/o Scram / 1 X EK3.11 - Reasons for initiating Emergency Boration 4.2 1 000032 L oss of Source R ange N I / 7 ........ ............. . ........... ..... ..... .. _ _ 000033 Loss of Intermediate Range NI / 7 ____ > ) (KI' t 000037 Steam Generator Tube Leak / 3 X AA1.13 - Operate / monitor SGBD rad monitor during tube leak 3.9 1 000038 Steam Generator Tube Rupture / 3 X EK3.08 - Criteria for securing RCPs during a SGTR 4.1 1 000054 Loss of Main Feedwater / 4 (PSA) W/E05 Inadequate Heat Transfer - Loss of Secondary X EK2.01 - Components, control / safety systems, including auto 3.7 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 3.5 1 to operate / monitor plant systems 000059 Accidental Liquid RadWaste Rel. / 9 X AA2.03 - Failure modes, symptoms, and causes of misleading 3.1 1 indications on liquid rad monitor 000060 Accidental Gaseous Radwaste Rel. / 9 .6 Determine -interpretv for release of mcI 3.6 000061 ARM System Alarms/ 7 W/E 16 High Containment Radiation /9 K/A Categor Totals: 1 3 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 NUREG-1021, Revision 8, Supplement 1 /* K/A Category Totals: 1 0 1 0 1 0 1 2 1 1 0 1Group Point Total: I 3

( C ( ES-401 PWR RO Examination Outline Bonn ES-401-4 Plant Systems - Tier 2/Grou 1 System # / Name K1 K2 K3 1(4 K5 K6 Al A2 A3 A4 G K/A Topic(s) Imp. Points 001 Control Rod Drive X. K2.02 - One line diagramofPS to trip breakers 3.6 3 X K4.11 - Design for resetting CRDM ckt breakers 2.7 X K5.42 - Definition of Tave and no load Tave 2.9 003 Reactor Coolant Pump XX----1.03 -Cause! [ffect between RCP and RCP seas 3.3 2 X A4.08 - Operate / monitor RCP cooling supplies 3.2 004 Chemical and Volume Control K3.08 - Effect of loss of CVCS on RCP seal 3.6 3 VX iniection X A1.06 - Monitor changes in VCT level 3.0 X A3.08 - Monitor CVCS changes on reactor power 3.9 013 Engineered Safety Features Actuation XK6.01 - Effect of loss of sensors / detectors on 2.7 2 ESFAS K "'X, A101- Predict impacts of LOCA on ESFAS 4 0K15 Nuclear Instrumentation K5.06 - Implication of subcritical multiplication on 3.4 2 NIS Ar3. Operatemotr NlS'trip bypasses: . 017 In-core Temperature Monitor A3.01 - Incore temp indications of normal, natural, 3.6 and interrupted circulation of RCS 022 Containment Cooling X I--(2.01 - Power supplies to containment cooling 3.0 2 X K4.01 - Design for cooling of containment 2.5 p1enetrations 056 Condensate A2.04 - Predict impacts of loss of condensate 2.6 1 pumps 059 Main Feedwater X K1.02 - Cause effect between MFW and AFW 3.4 2 Al...3.-Power.leel restrctions fr opiraton of 21 ý n n , ... ( .. .... ... .... ..... ... ......... ___MWpulmps Iand valves 061 Auxiliary/Emergency Feedwater n X2102 - Power.supplieo AW electric pumps ,3.7 2 (PSA) XK6.01 - Effect of loss of controllers / positioners on 2.5 AFW 068 Liquid Radwaste 2.1.32 - Explain / apply system limits and 3.4 1 precautions 071 Waste Gas Disposal X 2.2.25 - Bases in tech specs for LCOs and safety 2.5 1 lim its 3 1 072 Area Radiation Monitoring---K3.01 - Effect of loss of ARM on containment vent 3.2 t I A I isolation K/A Category Totals: j21 3J2 212 12121212 1 2 1 2 IGroup PointTotal: 23 1 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 [ 3j2 121 1 3 3 [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 System # / Name K1 K2 K3 K4 K5 K6 Al A2 A3 A4 G K/A Topic(s) Imp. Points 005 Residual Heat Removal X K4.08 - Lineup for piggy back mode with HP 3.1 1 injection 007 Pressurizer Relief/Quench Tank X A4.04 - Operate / monitor PZR vent valve 2.6 1 008 Component Cooling Water X K2.02 - Power to CCW pumps, including 3.0 1 emergency backup 027 Containment Iodine Removal 028HyroenReominr2.1Puge27 - Knowlege of system purpose I function 2.8 1 028 Hydrogen Recombiner and Purge I I.....2i* K~[g Control 034 Fuel Handling Equipment 041 Steam Dump/Turbine Bypass Control ___________________ 045 Main Turbine Generator C A1.06 - Expected response of secondary plant 3.3 1 parameters following TG trip 076 Service Water X..4...ow geo f a 078 Instrument Air- -- K1.03 - Cause/ effect between IAS and 3.3 1 nX containment air 103 Containment X-A4.06 - Operate / monitor containment personnel 2.7 1 X airlock door [./ACategory Totals: 1 i o1 0 i1 0 0o 1 0 2 2 Group PointTotal: 8 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 Category K/A # Topic Imp. Points 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 Operations 2.1.2 Knowledge of operator responsibilities during all modes of 3.0 1 operations. Total 4 2,212 K1owled1ge dfsiilagi~ropdtres 3.0 0 2.2.24 Analyze the effect of maintenance activities on LCO status. 2.6 1 Equipment 2.2.13 Knowledge of tagging and clearance procedures. 3.6 1 Control Total 3 2.3.10 Perform procedures to reduce excessive levels of radiation and 2.9 1 guard against personnel exposure. Radiation 2.3.11 Ability to control radiation releases. 2.7 1 Control Total 3 Plol 2.4.2 System setpoints, interlocks, and automatic actions associated with 3.9 1 EOP entry conditions. Emergency 2.4.7 Knowledge of event based EOP mitigation strategies. 3.1 1 Procedures/ Plan Total 3 Tier 3 Point Total J 13 NUREG- 1021, Revision 8, Supplement I }}

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