ML103400274
| ML103400274 | |
| Person / Time | |
|---|---|
| Site: | Palo Verde  |
| Issue date: | 11/05/2010 |
| From: | Division of Nuclear Materials Safety IV |
| To: | Arizona Public Service Co |
| References | |
| PV-2001-11 | |
| Download: ML103400274 (274) | |
Text
Rev 0 1 ID: Q22660 Points: 1.00 Given the following conditions:
Unit 1 is operating at 100% power.
B05 indicates VOPT on all 4 channels are tripped.
Phase current lights on B05 are lit.
CRS has entered SPTAs.
Which ONE of the following describes the next appropriate action?
A. Direct an AO to manually trip RTCBs.
B. Borate the RCS at full charging pump capacity.
C. Open Load Center Breakers L3 and L10 from B01.
D. RO will depress all 4 Reactor Trip pushbuttons on B05.
Answer: D Question 1 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 3.00 System ID: 22660 User-Defined ID: Q22660 Cross Reference Number:
Topic: Q22660 RO ATWS L3L10 BREAKER OPS Num Field 1: 3.9 Num Field 2: 4.1 Text Field: 4.1 029 EA1.11 Comments: Proposed reference to be provided to applicant during examination:
NONE KA STATEMENT:
Ability to operate and monitor the following as they apply to a ATWS:
Manual opening of the CRDS breakers.
TECHNICAL
REFERENCE:
40EP-9EO01 SPTAs JUSTIFICATION:
A. Incorrect - Only the FRP directs this action per RC-1 CEA Insertion step 3.
B. Incorrect - This would be correct if opening the LC Breakers L3 and L10 was unsuccessful. This is the next contingency to be taken per the SPTAs step 2.
C. Incorrect - This would be correct if depressing the RT pushbuttons on B05 was unsuccessful. This is the next contingency to be taken per the SPTAs step 2.
D. Correct - This action is directed in the SPTAs, step 2. This will manually open the RTCBs from the control room.
Rev 0 Question 1 Table-Item Links Q - 10CFR Sections 55.41 (7) Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
Q - Reactor Operator K & A Importance classification 3.9 Q - Cognitive Level Memory Q - Question Source New O - RO Tier Group Designations Tier 1 Group 1 Q - RO Exam 2010 RO NRC Associated objective(s):
Given a reactor trip describe the EOP expectation concerning the operation of the input breakers for L03 and L10
Rev 0 2 ID: Q6331 Points: 1.00 Given the following conditions:
A LOCA event results in a Reactor trip.
Containment Pressure is 3.5 psig and rising.
The SPTAs are in progress.
RCS Subcooling indicates 10 °F and is continuing to degrade.
Which ONE of the following describes the guidance regarding the operation of the RCPs?
A. Trip Two RCPs now (in SPTAs).
B. Trip Four RCPs now (in SPTAs).
C. The CRS shall not direct tripping of RCPs until an EOP is entered.
o D. The running RCPs shall remain operating until saturation conditions exist (0 F subcooling).
Answer: B
Rev 0 Question 2 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 2.00 System ID: 6331 User-Defined ID: Q6331 Cross Reference Number:
Topic: Q6331 GEN EOP LBLOCA RCP operation without NPSH Num Field 1: 2.6 Num Field 2: 2.7 Text Field: 41011EK202 Comments: Proposed reference to be provided to applicant during examination:
NONE TECHNICAL
REFERENCE:
40EP-9EO01 SPTAs KA STATEMENT: 4.1 011EK2.02 Knowledge of the interrelations between the pumps and the following: Large break LOCA: Pumps.
JUSTIFICATION:
A. Incorrect - All RCPs are to be secured with subcooling < 24 0F.
Candidate may confuse the trip 2 leave 2 strategy with RCS pressure remaining below the SIAS setpoint.
B. Correct - This is the SPTA contingency for loss of subcooling. RCPs should not be operated without adequate subcooling.
C. Incorrect - The expectation is that these pumps will be secured prior to exiting the SPTAs. Candidate may think that this is an early step of the LOCA EOP.
D. Incorrect - This does not meet the standards set by the EOP Technical Guideline. Candidate may understand loss of subcooling as < 0 0F subcooling, not the procedurally directed < 24 0F.
Rev 0 Question 2 Table-Item Links Q - 10CFR Sections 55.41 (10) Administrative, normal, abnormal, and emergency operating procedures for the facility.
55.41 (7) Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
Q - Reactor Operator K & A Importance classification 4.1 Q - Cognitive Level Comprehension / Anal Q - Question Source PV Bank Not Modified O - RO Tier Group Designations Tier 1 Group 1 Q - RO Exam 2010 RO NRC Associated objective(s):
describe the EOP expectation concerning the operation of the RCPs when subcooling is lost
Rev 0 3 ID: Q22668 Points: 1.00 Given the following conditions:
Unit 1 is operating at 100% power.
The power supply for CHE-P01 (E Charging Pump) is aligned to PBA-S03.
Charging Pump selector switch is in the 1-2-3 position.
PBA-S03 experiences a LOP.
Assuming NO operator action, which ONE of the following describes the plant response?
A. PZR Level will remain constant.
B. PZR Level will lower until letdown isolates on high temperature.
C. PZR Level will immediately rise due to letdown isolating on the LOP.
D. PZR Level will initially lower then return to program as letdown adjusts to the new charging flow.
Answer: B
Rev 0 Question 3 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 3.00 System ID: 22668 User-Defined ID: Q22668 Cross Reference Number:
Topic: Q22668 LOP EFFECTS ON CH AND PZR LEVEL Num Field 1: 3.2 Num Field 2: 3.7 Text Field: 4.2 022 AA2.02 Comments: Proposed reference to be provided to applicant during examination:
NONE KA STATEMENT:
Ability to determine and interpret the following as they apply to the Loss of Reactor Coolant Makeup: Charging pump problems TECHNICAL
REFERENCE:
40AO-9ZZ12 Degraded Electrical AOP, Operator Information Manual.
JUSTIFICATION:
A. Incorrect - Candidate may believe that letdown will be able to adjust quickly enough to maintain the PZR level constant.
B. Correct - A/E CCPs will be secured on the LOP, therefore charging flow will be 44 gpm and initial letdown flow is 78 gpm plus 8.8 gpm of seal bleedoff flow. Charging vs. Letdown mismatch will cause PZR level to initially lower until Letdown isolates on high temp at the regen Hx. Letdown temps rise due to the reduced charging flow available to cool the letdown flow.
C. Incorrect - Candidate may think the LOP will cause letdown isolation valves to close.
D. Incorrect - Candidate may not understand the charging vs. letdown relationship as it relates to rise in temp over time. The may think that letdown will back off enough to maintain letdown in service.
Rev 0 Question 3 Table-Item Links Q - 10CFR Sections 55.41 (6) Design, components, and functions of reactivity control mechanisms and instrumentation.
Q - Reactor Operator K & A Importance classification 3.2 Q - Cognitive Level Comprehension / Anal Q - Question Source New O - RO Tier Group Designations Tier 1 Group 1 Q - RO Exam 2010 RO NRC Associated objective(s):
Describe the automatic functions associated with the Charging Pumps.
Rev 0 4 ID: Q43744 Points: 1.00 Given the following conditions:
Unit 1 RCS level is at mid-loop at 101 ft. 4 in.
SDC is aligned to A Train with LPSI A running.
SDC TRAIN A/B FLOW LO (2B06A) is alarming.
LPSI A amps are oscillating.
Which ONE of the following actions will mitigate this condition?
A. Place SDC Train B in service.
B. Line up make-up from the RWT.
C. Start CS Pump A and Secure LPSI Pump A.
D. Start CS Pump A in parallel operation with LPSI A.
Answer: B Question 4 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 4.00 System ID: 43744 User-Defined ID: Q43744 Cross Reference Number:
Topic: Q43744 LOSS OF SDC Num Field 1: 2.9 Num Field 2: 2.8 Text Field: 4.2 025 AA1.22 Comments: Proposed reference to be provided to applicant during examination:
NONE KA STATEMENT:
Ability to operate and / or monitor the following as they apply to the Loss of Residual Heat Removal System: Obtaining of water from BWST for LPI system TECHNICAL
REFERENCE:
40AL-9RK2B, SDC Train A/B FLOW LO Alarm Response.
JUSTIFICATION:
A. Incorrect - Placing the other train in service will result in the issue being exhibited by the other SDC train.
B. Correct - Raising level will eliminate the vortexing issue that is causing the flow oscillations.
C. Incorrect - Shifting pumps will not correct the issue of vortexing, it will result in the issue being exhibited by the other pump.
D. Incorrect - Parallel pump operations will make the issue worse by raising flow.
Rev 0 Question 4 Table-Item Links Q - 10CFR Sections 55.41 (7) Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
Q - Reactor Operator K & A Importance classification 2.9 Q - Cognitive Level Comprehension / Anal Q - Question Source New O - RO Tier Group Designations Tier 1 Group 1 Q - RO Exam 2010 RO NRC Associated objective(s):
Identify the limitations associated with the SDC Low Flow Alarms for the RK, SESS and ERFDADS systems.
Rev 0 5 ID: Q43773 Points: 1.00 Given the following conditions:
Unit 1 has tripped from 100% power.
RDT level and pressure are rising slowly.
Pressurizer Relief Valves Status Panel on B04 shows a 9% Red LED lit for PSV-200.
Which ONE of the following describes the Pressurizer level response and the Pressurizer level Technical Specification 3.4.9?
Pressurizer level will ____(1)____. Enter Technical Specification 3.4.9 when Pressurizer level reaches ____(2)____.
A. (1) rise (2) 65%
B. (1) rise (2) 56%
C. (1) lower (2) 27%
D. (1) lower (2) 25%
Answer: B
Rev 0 Question 5 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 3.00 System ID: 43773 User-Defined ID: Q43773 Cross Reference Number:
Topic: Q43773 LCO 3.4.9 PZR SAFETY LIFTS Num Field 1: 3.4 Num Field 2: 3.7 Text Field: 2.2.40 Comments: Proposed reference to be provided to applicant during examination:
NONE TECHNICAL
REFERENCE:
40DP-9AP06, SPTA Tech Guideline, TS 3.4.9 KA STATEMENT:
Ability to apply Technical Specifications for a system. Pressurizer Vapor Space Accident.
JUSTIFICATION:
A. Incorrect - Level will rise post trip. 65% is the top of the SPTAs PZR level Setpoint.
B. Correct - Level will rise and the high level TS limit is 56%.
C. Incorrect - Level will rise and the 25% level represents the Low Level Heater cutout setpoint.
D. Incorrect - Level will rise but the 27% level represents the low level TS limit of 27%.
Rev 0 Question 5 Table-Item Links Q - 10CFR Sections 55.41 (10) Administrative, normal, abnormal, and emergency operating procedures for the facility.
Q - Reactor Operator K & A Importance classification 3.4 Q - Cognitive Level Comprehension / Anal Q - Question Source New O - RO Tier Group Designations Tier 1 Group 1 Q - RO Exam 2010 RO NRC Associated objective(s):
Describe LCO 3.4.9, Pressurizer, and its basis.
Rev 0 6 ID: Q22656 Points: 1.00 Given the following conditions:
Unit 1 is operating at 100% power.
Pressurizer level is 52% and stable.
Pressurizer pressure is 2295 psia and rising slowly.
Pressurizer Pressure Master Controller has a 0 output.
Pressurizer Main Spray valves RC-PV-100E and 100F are closed.
All pressurizer heaters are energized ON.
RRS system Tavg is selected to Tavg 1 (Loop 1).
Which ONE of the following describes the current condition?
A. Pressurizer Main Spray valves are failed closed.
B. Pressurizer Pressure Master Controller has failed.
C. Selected Pressurizer Level Instrument has failed high.
D. Selected Pressurizer Pressure Transmitter has failed low.
Answer: D
Rev 0 Question 6 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 3.00 System ID: 22656 User-Defined ID: Q22656 Cross Reference Number:
Topic: Q22656 RO APE PPCS Malfunction Num Field 1: 3.8 Num Field 2: 3.9 Text Field: 4.2 027 AK2.03 Comments: Proposed reference to be provided to applicant during examination:
NONE TECHNICAL
REFERENCE:
Operators Information Manual.
KA STATEMENT:
Knowledge of the interrelations between the Pressurizer Pressure Control Malfunctions and the following: Controllers and positioners.
JUSTIFICATION:
A. Incorrect - Spray valves are closed due to failing low PZR Press instrument, PPCS Master Control thinks pressure is lower than setpoint.
B. Incorrect - RRS system provides the output to PLCS for Tavg level setpoint, therefore failure of master pressure controller would not support this condition.
C. Incorrect - Level transmitter failure would cause pressurizer heaters to energize on insurge, but actual pressure would have spray valves open, PPCS Master Control thinks pressure is lower than setpoint.
D. Correct - PPCS Master Control thinks pressure is lower than setpoint, therefore pressure will rise due to the heaters on and spray valves shut.
Rev 0 Question 6 Table-Item Links Q - 10CFR Sections 55.41 (7) Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
Q - Reactor Operator K & A Importance classification 2.6 Q - Cognitive Level Comprehension / Anal Q - Question Source New O - RO Tier Group Designations Tier 1 Group 1 Q - RO Exam 2010 RO NRC Associated objective(s):
Describe the response of the Pressurizer Pressure Control System to a failure of an input transmitter.
Rev 0 7 ID: Q22625 Points: 1.00 Given the following conditions:
Initial Conditions:
Unit 1 is operating at 100% power.
CRS is performing 40AO-9ZZ02 (Excessive RCS leakage).
All Three (3) Charging Pumps are operating.
Letdown is still in service.
Leakrate is approximately 70 gpm.
Subsequently:
A 12 Finger CEA slips, resulting in a Reactor Trip.
40EP-9EO01 (SPTAs) are in progress.
No operator actions have been taken.
Pressurizer Pressure is 2000 psia and stable, and was never any lower.
CTMT Pressure is 0.5 psig and slowly rising.
RU-1 is in ALERT alarm and slowly trending up.
Assuming No Operator Action, which ONE of the following Pressurizer Level responses would the Reactor Operator observe?
PZR level is 26% and will A. lower until Safety Injection actuates and restores level.
B. restore to 33% with charging pumps due to automatic PLCS control.
C. rise until in a solid condition or charging pumps are manually secured.
D. restore to 52.5% with charging pumps due to automatic PLCS control.
Answer: B
Rev 0 Question 7 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 5 Difficulty: 4.00 System ID: 22625 User-Defined ID: Q22625 Cross Reference Number:
Topic: Q22625 EOP RO SBLOCA PZR LEVEL Response Num Field 1: 3.2 Num Field 2: 3.5 Text Field: 4.4 A16 AK2.1 Comments: Proposed reference to be provided to applicant during examination: NONE TECHNICAL
REFERENCE:
40AO-9ZZ04, RCP Emergencies KA STATEMENT: 4.1 009 EK3.10 Knowledge of the reasons for the following responses as they apply to the small break LOCA: Observation of PZR Level.
JUSTIFICATION:
A. Incorrect - The current RCS Leakrate is within the capacity of the charging system, therefore pressure and level will not lower to the point of SIAS.
B. Correct - Charging Capacity is 132 gpm and Leakrate plus seal bleedoff is approx 109 gpm. Automatic PLCS control will maintain at the 0% power setpoint of 33%.
C. Incorrect - Letdown is still in service therefore PLCS will maintain level at the 0% power setpoint of 33%.
D. Incorrect - The 100% setpoint level for PLCS is 52%, candidate may use this for the setpoint.
Rev 0 Question 7 Table-Item Links Q - 10CFR Sections 55.41 (5) Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.
Q - Reactor Operator K & A Importance classification 3.4 Q - Cognitive Level Comprehension / Anal Q - Question Source New O - RO Tier Group Designations Tier 1 Group 1 Q - RO Exam 2010 RO NRC Associated objective(s):
Given conditions of LOCA describe how the plant would respond to various types of RCS leaks in accordance with 40EP-9EO03.
Rev 0 8 ID: Q22663 Points: 1.00 Given the following conditions:
Unit 1 has been tripped from 100% power due to a SGTR.
Which ONE of the following describes when the affected SG will be isolated.
A. Thot < 540 0F.
B. Tcold < 540 0F.
C. Thot < 540 0F. and RCS pressure within 50 psia of faulted SG.
D. Tcold < 540 0F and RCS pressure within 50 psia of faulted SG.
Answer: A
Rev 0 Question 8 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 2.00 System ID: 22663 User-Defined ID: Q22663 Cross Reference Number:
Topic: Q22663 RO EPE SGTR WHEN TO ISOLATE Num Field 1: 4.1 Num Field 2: 4.7 Text Field: 4.1 038 EA2.01 Comments: Proposed reference to be provided to applicant during examination: NONE TECHNICAL
REFERENCE:
40EP-9EO04 SGTR KA STATEMENT:
Ability to determine or interpret the following as they apply to a SGTR:
When to isolate one or more S/Gs.
JUSTIFICATION:
A. Correct - This is done to prevent lifting an MSSV after the SG is isolated.
B. Incorrect - Isolation is done when Thot < 540 0F. Candidate may not understand the Tcold vs. Thot and why Thot is used for isolation.
C. Incorrect - Isolation is done when Thot < 540 0F. RCS pressure is lowered to within 50 psia but not a criteria for isolating. The pressure portion of the procedure is used to ensure that the size of the leak is reduced and to minimize the dilution of the RCS.
D. Incorrect - Isolation is done when Thot < 540 0F. RCS pressure is lowered to within 50 psia but not a criteria for isolating. The pressure portion of the procedure is used to ensure that the size of the leak is reduced and to minimize the dilution of the RCS.
Rev 0 Question 8 Table-Item Links Q - 10CFR Sections 55.41 (5) Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.
Q - Reactor Operator K & A Importance classification 4.1 Q - Cognitive Level Memory Q - Question Source New O - RO Tier Group Designations Tier 1 Group 1 Q - RO Exam 2010 RO NRC Associated objective(s):
Given a SGTR is occurring describe how and why a ruptured SG will be isolated and confirmed to be the leaking SG in accordance with 40EP-9EO04.
Rev 0 9 ID: Q7243 Points: 1.00 Given the following conditions:
Unit 1 is operating at 100% power.
2A RCP Upper Thrust Bearing temperature is observed to be in alarm at 230 0F and rising.
In accordance with 40AO-9ZZ04 (RCP Emergencies), which ONE of the following actions can be performed to slow the rate of temperature rise?
A. Start the 2A hydraulic oil lift pump PO2C.
B. Start the standby Nuclear Cooling Water pump.
C. Increase the output of CHN-FIC-243, Seal Injection Flow Controller.
D. Stop all but one Normal Chiller; ensure that only one Nuclear Cooling Water outlet valve is open.
Answer: A
Rev 0 Question 9 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 1 Difficulty: 2.00 System ID: 7243 User-Defined ID: Q7243 Cross Reference Number: 40AO-9ZZ04, RCP EMERG Topic: Q7243 RCP Thrust bearing failure indications Num Field 1: 4.0 Num Field 2: 4.2 Text Field: 2.4.11 Comments: Proposed reference to be provided to applicant during examination:
NONE Technical
Reference:
40AO-9ZZ04, RCP Emergencies K&A:
Knowledge of abnormal condition procedures as it relates to RCP malfunctions.
Justification:
A. Correct - RCP Emergencies procedurally directs this action for a high temperature bearing condition.
B. Incorrect - RCP Emergencies procedurally directs this in the case of high temperature and loss of cooling water.
C. Incorrect - These are reverse acting controllers, increasing output would cause a reduction in flow and increase temperature.
D. Incorrect - directed in the case of high temperature and a loss of cooling water.
Rev 0 Question 9 Table-Item Links Q - 10CFR Sections 55.41 (10) Administrative, normal, abnormal, and emergency operating procedures for the facility.
Q - Reactor Operator K & A Importance classification 4.0 Q - Cognitive Level Memory Q - Question Source PV Bank NOT Modified O - RO Tier Group Designations Tier 1 Group 1 Q - RO Exam 2009 RO Exam 2010 RO NRC Associated objective(s):
Given RCP motor amps and Upper Thrust Bearing Temperature
Rev 0 10 ID: Q43812 Points: 1.00 Given the following conditions:
Unit 1 has been in a Blackout condition for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
The crew is performing action of 40EP-9EO08 (Blackout).
Attempts to restore power from other sources have been unsuccessful.
The following parameters exist:
REP CET indicated 589°F and stable.
RCS pressure indicates 1660 psia and slowly lowering.
Pressurizer level indicates 7% and slowly lowering.
Which ONE of the following actions (if any) should be taken by the crew to ensure Natural Circulation conditions exist?
A. No action required.
B. Raise RCS pressure.
C. Raise Pressurizer level.
D. Control steam generator feed and steam.
Answer: D
Rev 0 Question 10 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 3.00 System ID: 43812 User-Defined ID: Q43812 Cross Reference Number:
Topic: Q43812 Blackout action to maintain subcooling Num Field 1: 4.1 Num Field 2: 4.4 Text Field: 4.1 055 EK1.02 Comments: Proposed reference to be provided to applicant during examination: Steam Tables Technical
Reference:
40EP-9EO08, BLACKOUT / 40DP-9AP13. BO Tech Guideline K&A:
Knowledge of the operational implications of the following concepts as they apply to the Station Blackout : Natural circulation cooling.
Justification:
A. Incorrect - Conditions per Blackout of 24 0F superheat are not met, therefore action is required to restore subcooling.
B. Incorrect - Raising RCS pressure would improve subcooling, but the conditions of blackout do not support energizing of heaters.
C. Incorrect - Raising PZR level will not support Natural Circulation conditions, also the conditions of blackout do not support the use of a charging pump.
D. Correct - Per Step 21 Blackout EOP, if the conditions are met, ENSURE proper control of steam generator steaming and feeding.
Rev 0 Question 10 Table-Item Links Q - 10CFR Sections 55.41 (10) Administrative, normal, abnormal, and emergency operating procedures for the facility.
55.41 (8) Components, capacity, and functions of emergency systems.
Q - Reactor Operator K & A Importance classification 4.1 Q - Cognitive Level Comprehension / Anal Q - Question Source New O - RO Tier Group Designations Tier 1 Group 1 Q - RO Exam 2010 RO NRC Associated objective(s):
state the action necessary to maintain subcooling margin
Rev 0 11 ID: Q43745 Points: 1.00 Given the following conditions:
Unit 1 has tripped from 100% power due to a LOOP.
SPTAs are in progress.
Which ONE of the following describes the use of alarm response procedures.
Alarm response procedures may be suspended until...
A. SPTAs are exited.
B. the plant stabilizes.
Answer: B
Rev 0 Question 11 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 3.00 System ID: 43745 User-Defined ID: Q43745 Cross Reference Number:
Topic: Q43745 ARP use in LOOP Num Field 1: 4.2 Num Field 2: 4.1 Text Field: 2.4.31 Comments: Proposed reference to be provided to applicant during examination: None Technical
Reference:
40DP-9AP16 (EOP users guide)
K&A:
Knowledge of annunciator alarms and indications, and use of the response instructions. LOOP Justification:
A. Incorrect - There is no requirement to resume Alarm Response based on procedure entry or exit. Candidate may assume that changing procedures may mean stable plant conditions.
B. Correct - This is described in 40DP-9AP16 (EOP users guide) step 28.
C. Incorrect - There is no requirement to resume Alarm Response based on procedure entry or exit. Candidate may assume that changing procedures may mean stable plant conditions.
D. Incorrect - LOOP EOP does not direct using ARP, various alarms are entry conditions for this EOP.
Rev 0 Question 11 Table-Item Links Q - 10CFR Sections 55.41 (10) Administrative, normal, abnormal, and emergency operating procedures for the facility.
Q - Reactor Operator K & A Importance classification 4.2 Q - Cognitive Level Memory Q - Question Source New O - RO Tier Group Designations Tier 1 Group 1 Q - RO Exam 2010 RO NRC Associated objective(s):
Given that an ORP is being implemented describe the use of an AL when the reactor trips or when performing an EOP in accordance with 40DP-9AP16.
Rev 0 12 ID: Q43746 Points: 1.00 Given the following conditions:
Unit 1 is operating in Mode 2.
The supply breaker to inverter PNC-N13 is inadvertently opened at PKC-M43.
Based on these conditions, PNC-D27 will:
A. lose its alternate power supply but remain energized.
B. automatically align to its alternate power supply and remain there until manually transferred back to the inverter.
C. automatically align to its alternate power supply and then, when the inverter is restored, automatically transfer back to its normal source.
D. be de-energized for as long as the supply breaker is open. Power can be restored by manually swapping the inverter output to its alternate supply.
Answer: D
Rev 0 Question 12 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 2.00 System ID: 43746 User-Defined ID: Q43746 Cross Reference Number:
Topic: Q43746 PN Unit 1 inverter failure Num Field 1: 3.7 Num Field 2: 3.7 Text Field: 4.2 057 AA1.01 Comments: Proposed reference to be provided to applicant during examination: None Technical
Reference:
40AO-9ZZ13 (Loss of Class Instrument or Control Power)
K&A:
Ability to operate and / or monitor the following as they apply to the Loss of Vital AC Instrument Bus: Manual inverter swapping.
Justification:
A. Incorrect - The supply breaker for the inverter is the normal power supply and the backup supply must be closed manually.
B. Incorrect - Unit 1 does not have a static transfer switch, This would be correct in Units 2 and 3.
C. Incorrect - Unit 1 does not have a static transfer switch, and the inverters are not self normalizing, therefore even if Unit 1 had the static transfer switch, the inverter would have to be manually aligned upon restoration of power.
D. Correct - Unit 1 does not have a static transfer switch, so all lineups will be manual.
Rev 0 Question 12 Table-Item Links Q - 10CFR Sections 55.41 (7) Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
Q - Reactor Operator K & A Importance classification 3.7 Q - Cognitive Level Memory Q - Question Source New O - RO Tier Group Designations Tier 1 Group 1 Q - RO Exam 2010 RO NRC Associated objective(s):
Explain the operation of the Static Transfer Switch which is provided on Inverters in Units 2 and 3. Manual Transfer Switch in Unit 1.
Rev 0 Question 13 Info Question Type: Multiple Choice Status: Active Always select No on test?
Authorized for No practice?
Points: 1.00 Time to 2
Complete:
Difficulty: 2.00 System ID: 22669 User-Defined Q22669 ID:
Cross Reference Number:
Topic: Q22669 Loss of DC battery charger.
Num Field 1: 3.7 Num Field 2: 4.1 Text Field: 4.2 058 AA2.01 Comments: Proposed reference to be provided to applicant during examination: None Technical
Reference:
40OP-9PK01(125 VDC 1E Electrical System)
K&A:
Ability to determine and interpret the following as they apply to the Loss of DC Power: That a loss of dc power has occurred; verification that substitute power sources have come on line.
Justification:
A. Incorrect - In a normal alignment, Battery Charger B DC amps would be approximately 120 DC amps.
B. Incorrect - If the B battery output breaker had tripped open, Battery B DC amps would be 0 DC amps.
C. Correct - In a normal alignment, Battery Charger B DC amps would be approximately 120 DC amps.
D. Incorrect - In a normal alignment Battery Charger BD DC amps would be 0 DC amps. This is normal indication when the BD Battery Charger is not aligned. The BD battery charger is not normally aligned.
Rev 0 Question 13 Table-Item Links Q - 10CFR Sections 55.41 (7) Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
Q - Reactor Operator K & A Importance classification 3.7 Q - Cognitive Level Comprehension / Anal Q - Question Source New O - RO Tier Group Designations Tier 1 Group 1 Q - RO Exam 2010 RO NRC Associated objective(s):
Describe the Control Room indications associated with the Class IE 125 VDC Power system.
Rev 0 14 ID: Q10499 Points: 1.00 Given the following conditions:
Unit 1 has tripped from 100% power due to a Loss of Offsite Power (LOOP) event.
The crew has cross connected Essential Cooling Water (EW) system "A" to Nuclear Cooling Water per Standard Appendix 63.
Which ONE of the following describes the reason for throttling closed EWA-HCV-53 (SDCHX A Outlet Isolation) to a minimum EW system flow of 8500 gpm or until the RCP LO NC FLOW alarms are clear?
A. Maintain operability of the EW "A" system.
B. Provide the NCW priority loads with adequate flow.
C. Prevent EW pump damage due to operating in a runout condition.
D. Ensure adequate flow to the Fuel Pool Heat Exchangers if aligned.
Answer: B Question 14 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 3.00 System ID: 10499 User-Defined ID: Q10499 Cross Reference Number:
Topic: Q10499 X-tie EW to NC during a LOOP Num Field 1: 4.0 Num Field 2: 4.2 Text Field: 4. 2 062 AK3.03 Comments: Proposed reference to be provided to applicant during examination: None Technical
Reference:
SA-63 and Tech Guide/ 40DP-9AP12, LOOP Tech Guide /
K&A: Knowledge of the reasons for the following responses as they apply to the Loss of Nuclear Service Water: Guidance actions contained in EOP for Loss of nuclear service water Justification:
A. Incorrect - EW is INOP when cross tied.
B. Correct - this is to ensure adequate flow to NCW priority loads.
C. Incorrect- not for runout but examinee could believe this to be true when taking systems out of normal alignment.
D. Incorrect- One train of EW not sized to provide both NCW and Fuel Pool cooling, this would require aligning the other train.
Rev 0 Question 14 Table-Item Links Q - 10CFR Sections 55.41 (4) Secondary coolant and auxiliary systems that affect the facility.
55.41 (8) Components, capacity, and functions of emergency systems.
Q - Reactor Operator K & A Importance classification 4.0 Q - Cognitive Level Memory Q - Question Source PV Bank Not Modified O - RO Tier Group Designations Tier 1 Group 1 Q - RO Exam RO EXAM QUESTION 2008 2010 RO NRC Associated objective(s):
Cross tie EW to NC while in the LOOP/LOFC EOP.
Rev 0 15 ID: Q43742 Points: 1.00 Which ONE of the following describes why the AOP directs aligning a Fuel Building Essential AFU to the auxiliary building during a complete loss of Instrument Air?
A. The essential AFU simply augments the normal AFUs, ensuring a negative pressure is maintained.
B. Due to the number of additional heat loads (motors and piping), additional air movement is necessary.
C. The essential AFU will circulate air below the 100' elevation, which may contain nitrogen from leaking air connections.
D. Since the normal AHUs and AFUs trip on a loss of Instrument Air, the Fuel Building essential AFUs will ensure a negative pressure is kept on the building.
Answer: C
Rev 0 Question 15 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 3.00 System ID: 43742 User-Defined ID: Q43742 Cross Reference Number:
Topic: Q43742 HF alignment on loss of IA Num Field 1: 3.0 Num Field 2: 3.2 Text Field: 4.2 065 AK3.04 Comments: Proposed reference to be provided to applicant during examination: None Technical
Reference:
40OP-9HF01( Fuel Building HVAC),
40AO9ZZ06 (Loss of IA)
K&A:
Knowledge of the reasons for the following responses as they apply to the Loss of Instrument Air: Cross-over to backup air supplies Justification:
A. Incorrect - The normal AFUs are not lined up to the aux bldg at all times, candidate may think that normal AFUs are always lined up to the lower levels of the aux building.
B. Incorrect - Heat generation is not a concern in this event, therefore the AHUs would not be needed to dissipate the heat. During a SIAS, the heat generated from the pumps and possibility of radioactive gasses is the reason for shifting suctions.
C. Correct - N2 buildup is a concern, the fuel building AFU has suctions on the lower levels of the Aux Building, therefore removing the N2.
D. Incorrect - The normal AHUs do not trip on loss of IA. Candidate may think that the dampers are air powered and therefore would be lost.
Rev 0 Question 15 Table-Item Links Q - 10CFR Sections 55.41 (10) Administrative, normal, abnormal, and emergency operating procedures for the facility.
55.41 (8) Components, capacity, and functions of emergency systems.
Q - Reactor Operator K & A Importance classification 3.0 Q - Cognitive Level Memory Q - Question Source New O - RO Tier Group Designations Tier 1 Group 1 Q - RO Exam 2010 RO NRC Associated objective(s):
Identify when the Nitrogen backup will begin supplying the system and what equipment should be monitored if the backup is in service.
Rev 0 16 ID: Q43747 Points: 1.00 Given the following conditions:
Initial Conditions:
Unit 2 has tripped from 100% power.
SG #1 is 1000 psia and lowering.
SG #2 is 800 psia and lowering.
PZR level is at 30% and slowly lowering.
Containment Pressure is 1 psig and rising.
At the time that the ORP is entered the conditions are as follows:
Containment pressure peaked and is stable at 9.8 psig.
Containment temperature is 185°F.
PZR level is 12% and rising.
RVUH level is 67%.
RCS subcooling is 98°F.
SG #1 is at 34% WR (rising) and being fed from AFW at 500 gpm.
SG #2 is below the indicated level.
Both HPSI pumps are injecting into the RCS.
Based on these conditions, you should obtain CRS concurrence and throttle HPSI...
A. immediately.
B. when PZR level reaches 15%.
C. when RVUH is equal to 100%.
D. when SG #1 Level are 45%-60% NR.
Answer: B
Rev 0 Question 16 Info Question Multiple Choice Type:
Status: Active Always select No on test?
Authorized No for practice?
Points: 1.00 Time to 2
Complete:
Difficulty: 3.00 System ID: 43747 User-Defined Q43747 ID:
Cross Reference Number:
Topic: Q43747 EOP ESD Thottling HPSI Num Field 1: 2.9 Num Field 2: 3.2 Text Field: 4.4 E05 EK1.1 Comments: Proposed reference to be provided to applicant during examination: NONE Technical
Reference:
40EP-9EO05, Excess Steam Demand, 40EP-9EO10 Appendix 2 SI Throttle Criteria K&A:
Knowledge of the operational implications of the following concepts as they apply to the (Excess Steam Demand) Components, capacity, and function of emergency systems.
Justification:
A. Incorrect - PZR level requirement is > 15% for Harsh CTMT conditions. Candidate may believe the 10% non harsh CTMT criteria is met.
B. Correct - PZR level requirement for throttling HPSI is > 15% level when in Harsh CTMT conditions.
C. Incorrect - RVUH level must be greater than 16% to throttle HPSI, which it is. Candidate may not understand RVUH and Plenum relationship.
D. Incorrect - The SG requirement is RESTORING to 45-60% NR level. Candidate may believe that SG levels must be in the band.
Rev 0 Question 16 Table-Item Links Q - 10CFR Sections 55.41 (10) Administrative, normal, abnormal, and emergency operating procedures for the facility.
55.41 (8) Components, capacity, and functions of emergency systems.
Q - Reactor Operator K & A Importance classification 3.0 Q - Cognitive Level Comprehension / Anal Q - Question Source New O - RO Tier Group Designations Tier 1 Group 1 Q - RO Exam 2010 RO NRC Associated objective(s):
describe the mitigating strategy outlined in the ESD EOP
Rev 0 17 ID: Q43748 Points: 1.00 Given the following conditions:
Unit 1 has tripped from 100% power.
SPTAs are complete and Reactor Trip EOP has been entered.
The following electrical buses are available:
NAN-S01.
NBN-S01.
PBA-S03.
PKA-M41.
PNA-D25.
Which ONE of the following describes the status of the Maintenance of Vital Auxiliaries safety function and actions (if any) required to satisfy the safety function?
The Maintenance of Vital Auxiliaries safety function is...
A. satisfied.
B. not satisfied, energize PNB-D26.
C. not satisfied, energize PKB-M42.
D. not satisfied, energize PKC-M43.
Answer: D
Rev 0 Question 17 Info Question Multiple Choice Type:
Status: Active Always select on No test?
Authorized No for practice?
Points: 1.00 Time to 2
Complete:
Difficulty: 3.00 System ID: 43748 User-Defined Q43748 ID:
Cross Reference Number:
Topic: Q43748 MVA SFSC RT Num Field 1: 3.0 Num Field 2: 3.4 Text Field: 4.4 E02 EK1.2 Comments: Proposed reference to be provided to applicant during examination:
NONE Technical
Reference:
40EP-9EO02, Reactor Trip K&A:
Knowledge of the operational implications of the following concepts as they apply to the (Reactor Trip Recovery) Normal, abnormal and emergency operating procedures associated with (Reactor Trip Recovery).
Justification:
A. Incorrect - not satisfied due to requiring a PK bus on the same side as the vital 4.16 kV bus.
B. Incorrect - not satisfied due to requiring a PK bus on the same side as the vital 4.16 kV bus. PNB-D26 is downstream of PBB-S04(Which is not available).
C. Incorrect - not satisfied due to requiring a PK bus on the same side as the vital 4.16 kV bus. PKB-M42 is downstream of PBB-S04.
D. Correct - not satisfied due to requiring a PK bus on the same side as the vital 4.16 kV bus. PKC-M43 is downstream of PBA-S03.
Rev 0 Question 17 Table-Item Links Q - 10CFR Sections 55.41 (10) Administrative, normal, abnormal, and emergency operating procedures for the facility.
Q - Reactor Operator K & A Importance classification 3.0 Q - Cognitive Level Comprehension / Anal Q - Question Source PV Bank Not Modified O - RO Tier Group Designations Tier 1 Group 1 Q - RO Exam 2010 RO NRC Associated objective(s):
analyze the MVA to determine if the SFSC acceptance criteria is satsified
Rev 0 18 ID: Q43818 Points: 1.00 Given the following conditions:
Unit 2 has tripped from 100% power.
SPTAs are complete and the CRS has entered 40EP-9EO06 (LOAF).
SG #1 pressure is 1195 psia.
SG #2 pressure is 1040 psia.
Based on the above indications, which ONE of the following actuations (if any) has occurred?
A. None.
B. AFAS-1 Only.
C. AFAS-2 Only.
D. AFAS-1 and AFAS-2.
Answer: D
Rev 0 Question 18 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 2.00 System ID: 43818 User-Defined ID: Q43818 Cross Reference Number:
Topic: Q43818 AFAS SETPOINT Num Field 1: 3.3 Num Field 2: 3.7 Text Field: 4.4 E06 EK2.1 Comments: Proposed reference to be provided to applicant during examination:
NONE Technical
Reference:
EOP Setpoint Document, LOIT Lesson Plan.
K&A:
Knowledge of the interrelations between the (Loss of Feedwater) and the following: Components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
Justification:
A. Incorrect - SG #1 level is 24% and SG #2 level is at 20%, candidate may not know the AFAS setpoint and believe that both SGs still greater than the setpoint.
B. Incorrect - SG #2 Pressure is 150 psia < SG #1, candidate may believe that only SG #1 will have the AFAS due to a D/P lockout (actual setpoint of 185 psid).
C. Incorrect - SG #2 level is at 20%, candidate may not know the AFAS setpoint and believe that SG #1 is still greater than the setpoint.
D. Correct - AFAS setpoint is 25.8% WR, therefore both SG #1 and SG
- 2 will initiate an AFAS signal.
Rev 0 Question 18 Table-Item Links Q - 10CFR Sections 55.41 (7) Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
Q - Reactor Operator K & A Importance classification 3.3 Q - Cognitive Level Memory Q - Question Source New O - RO Tier Group Designations Tier 1 Group 1 Q - RO Exam 2010 RO NRC Associated objective(s):
Describe what automatically initiates Auxiliary Feedwater Actuation Signal (AFAS) and its function.
Rev 0 19 ID: Q43761 Points: 1.00 Given the following conditions:
Unit 1 is operating at 100% power with CEAs ARO.
A single Group 5 CEA drops into the core.
Pulse Counter Rod Position Indication for the single Group 5 CEA 150 inches.
Actual Rod position for the single Group 5 CEA is 15 inches.
Which ONE of the following explains the difference between the actual and the pulse counter CEA position?
The A. Lower Group Stop (LGS) has NOT been met.
B. Lower Electrical Limit (LEL) has NOT been met.
C. Dropped Rod Contact (DRC) has NOT been activated.
D. Pulse Counters DO NOT reset until a Rod Motion demand is sensed.
Answer: C
Rev 0 Question 19 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 3.00 System ID: 43761 User-Defined ID: Q43761 Cross Reference Number:
Q43761 Explain the operation of the RSPT (Reed Switch Position Topic:
Transmitter).
Num Field 1: 3.7 Num Field 2: 3.9 Text Field: 4.2 003 AA2.01 Comments: Proposed reference to be provided to applicant during examination:
NONE Technical
Reference:
CEDMCS Lesson Plan, Operator Information Manual.
K&A:
Ability to determine and interpret the following as they apply to the Dropped Control Rod: Rod position indication to actual rod position.
Justification:
A. Incorrect - Lower Group Stop does not input to RPI but is part of the PC which will stop group motion at 3/4 step withdrawn.
B. Incorrect - Lower Electrical Limit does not input to RPI but is part of the RSPTs which do provide the DRC.
C. Correct - The DRC will send the signal to the Plant Computer which provides the Rod Position to B05.
D. Incorrect - Pulse Counters do provide RPI to the PC that will be indicated on B05, but not until the DRC is met. Candidate may confuse the reset when demand is sensed with the AMI reset associated with the RRS.
Rev 0 Question 19 Table-Item Links Q - 10CFR Sections 55.41 (2) General design features of the core, including core structure, fuel elements, control rods, core instrumentation and coolant flow.
Q - Reactor Operator K & A Importance classification 3.7 Q - Cognitive Level Comprehension / Anal Q - Question Source New O - RO Tier Group Designations Tier 1 Group 2 Q - RO Exam 2010 RO NRC Associated objective(s):
Explain the operation of the RSPT (Reed Switch Position Transmitter).
Rev 0 20 ID: Q22435 Points: 1.00 Given the following conditions:
Unit 1 was initially at 100% power (250 EFPD).
CRS has directed an Emergency Boration due to a loss of Shutdown Margin.
1000 gallons of boron have been sent to the Charging Pump suction at 40 gpm.
Turbine load has been adjusted to maintain RCS temperature on program.
From memory, using the STA reactivity control worksheet approximate values.
Which ONE of the following describes the current condition?
A. Reactor Power is ~ 90%, Tave has risen and Tcold has risen.
B. Reactor Power is ~ 80%, Tave has lowered and Tcold has risen.
C. Reactor Power is ~ 90%, Tave has risen and Tcold has lowered.
D. Reactor Power is ~ 80%, Tave has lowered and Tcold has lowered.
Answer: B
Rev 0 Question 20 Info Question Multiple Choice Type:
Status: Active Always select No on test?
Authorized for No practice?
Points: 1.00 Time to 3
Complete:
Difficulty: 4.00 System ID: 22435 User-Defined Q22435 ID:
Cross Reference Number:
Topic: Q22435 Reactor power response to emergency boration Num Field 1: 3.6 Num Field 2: 3.9 Text Field: 4.2 024 AK1.02 Comments: Proposed reference to be provided to applicant during examination: NONE Technical
Reference:
Basic Operator knowledge Per UIC16 Rev 0.
Current Simulator Setup.
K&A: Knowledge of the operational implications of the following concepts as they apply to Emergency Boration: Relationship between boron addition and reactor power Justification: Per UIC16 Rev 0. Current Simulator Setup. Updated from 2008 RO NRC Exam.
U1C16 STA Reactivity Control Worksheet quick reference numbers or Core Data Book state, 53 gal of Boration per 1% power Tave is 585ºF (100%) to 564ºF (0%)
T cold is 558ºF (100%) to 564ºF (0%)
Boron addition rate of 40 gpm means xenon has had little effect at this time A. Incorrect - Candidate may not know the reference values and may not understand how the RCS will respond to a lowering Rx Power with Tcold being maintained on program with a Lowering Turbine Load.
B. Correct - (1000 gal)/(53 gal boration/ 1% power) = 18.9% power change. Tave will lower due to Thot lowering with Tcold remaining Constant.
C. Incorrect - Incorrect - Candidate may not know the reference values and may not understand how the RCS will respond to a lowering Rx Power with Tcold being maintained on program with a Lowering Turbine Load.
D. Incorrect - Incorrect - Candidate may not know the reference values and may not understand how the RCS will respond to a lowering Rx Power with Tcold being maintained on program with a Lowering Turbine Load.
Rev 0 Question 20 Table-Item Links Q - 10CFR Sections 55.41 (10) Administrative, normal, abnormal, and emergency operating procedures for the facility.
55.41 (5) Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.
55.41 (8) Components, capacity, and functions of emergency systems.
Q - Reactor Operator K & A Importance classification 3.6 Q - Cognitive Level Comprehension / Anal Q - Question Source PV Bank Not Modified O - RO Tier Group Designations Tier 1 Group 2 Q - RO Exam RO EXAM QUESTION 2008 2010 RO NRC Associated objective(s):
identify entry and exit of Emergency Boration
Rev 0 21 ID: Q43762 Points: 1.00 Which ONE of the following would result in a Loss of the Start Up Channel NIs?
Loss of power to...
A. 125 vdc Class Control Power (PK).
B. 125 vdc Non-Class Control Power (NK).
C. 120 vac Class Instrument and Control Power (PN).
D. 120 vac Non Class Instrument and Control Power (NN).
Answer: D Question 21 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 3.00 System ID: 43762 User-Defined ID: Q43762 Cross Reference Number:
Topic: Q43762 LOSS OF PWR ON SR NIs Num Field 1: 2.7 Num Field 2: 3.1 Text Field: 4.2 032 AK2.01 Comments: Proposed reference to be provided to applicant during examination:
NONE Technical
Reference:
Power Supply Table K&A:
Knowledge of the interrelations between the Loss of Source Range Nuclear Instrumentation and the following: Power supplies, including proper switch positions.
Justification:
A. Incorrect - Candidate may not know the power supply to the Start Up Nis and may pick the Class DC control power.
B. Incorrect - Candidate may not know the power supply to the Start Up Nis and may pick the Non Class DC control power.
C. Incorrect - PN is the power supply to the Log Safety Channels.
Candidate may not know the power supply to the Start Up Nis and may pick the Class AC instrument and control power.
D. Correct - The Power Supplies to the Startup Channel NIs is NNN-D11 and D12.
Rev 0 Question 21 Table-Item Links Q - 10CFR Sections 55.41 (7) Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
Q - Reactor Operator K & A Importance classification 2.7 Q - Cognitive Level Memory Q - Question Source New O - RO Tier Group Designations Tier 1 Group 2 Q - RO Exam 2010 RO NRC Associated objective(s):
Describe how the Nuclear Instrumentation System is supported by the following systems:
- Containment Building HVAC (HC)
Rev 0 22 ID: Q7819 Points: 1.00 Given the following conditions:
Unit 1 was tripped due to a degrading condenser vacuum.
Immediately after the trip, the Secondary Operator became involved with the B01 report.
When this operator returns to B06, condenser backpressure is 6.0" HgA (highest shell).
Steam Generator levels are both 65% WR.
Assuming no other Operator actions on B06, which ONE of the following describes the condition of the secondary plant?
A. No feed with SBCS valves 1001 & 1004 maintaining SG pressure until Condenser pressure reaches 7.5 inches HgA.
B. Main Feed Pumps feeding in RTO with SBCS valves 1001 & 1004 maintaining SG pressure until Condenser pressure reaches 7.5 inches HgA.
C. No feed, SBCVs 1001 thru 1006 locked out to prevent Condenser over pressurization and SBCVs 1007 & 1008 maintaining SG pressure.
D. Main Feed Pumps feeding in RTO, SBCVs 1001 thru 1006 locked out to prevent Condenser over pressurization and SBCVs 1007 & 1008 maintaining SG pressure.
Answer: D
Rev 0 Question 22 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 3.00 System ID: 7819 User-Defined ID: Q7819 Cross Reference Number:
Topic: Q7819 AOP 40AO-9ZZ07 Impact of rising backpressure Num Field 1: 2.8 Num Field 2: 3.1 Text Field: 4.2 051 AK3.01 Comments: Proposed reference to be provided to applicant during examination:
NONE Technical
Reference:
40AO-9ZZ07, Loss of Vacuum K&A: Knowledge of the reasons for the following responses as they apply to the Loss of Condenser Vacuum: Loss of steam dump capability upon loss of condenser vacuum JUSTIFICATION:
A. Incorrect - Main feed pumps (MFPs) don't trip till 13.5 inches HgA, SBCS condenser valves (1001 - 1006) lock out on condenser interlock at 5.5 inches HgA.
B. Incorrect - MFPs are feeding but 1001 -1006 locked out at 5.5 inches HgA.
C. Incorrect - MFPs are supplying feedwater, but the SBCS valves 1007/1008 are correct.
D. Correct - MFPs are feeding in RTO with SBCS valves 1007/1008 maintaining RCS heat removal automatically to atmosphere.
Rev 0 Question 22 Table-Item Links Q - 10CFR Sections 55.41 (5) Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.
Q - Reactor Operator K & A Importance classification 2.8 Q - Cognitive Level Comprehension / Anal Q - Question Source PV Bank Not Modified O - RO Tier Group Designations Tier 1 Group 2 Q - RO Exam RO Exam Question 2007 2010 RO NRC Associated objective(s):
Describe the main condenser vacuum interlock associated with the SBCS System including how it is reset.
Rev 0 23 ID: Q43774 Points: 1.00 Given the following conditions:
RU-145 (Fuel Building Ventilation, Low Range Gas) is in ALERT alarm.
RU-31 (Spent Fuel Pool Area) is in HIGH alarm.
74RM-9EF41 (Radiation Monitoring System Alarm Response) requires the operator to verify FBEVAS and CREFAS have actuated.
Which ONE of the following describes the reason for performing this action?
A. FBEVAS and CREFAS both actuated off the HIGH alarm.
B. FBEVAS actuated off the HIGH alarm; CREFAS actuated off the cross trip.
C. CREFAS actuated off the HIGH alarm; FBEVAS actuated off the cross trip.
D. FBEVAS actuated off the HIGH alarm; CREFAS actuated off the ALERT alarm.
Answer: B Question 23 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 3.00 System ID: 43774 User-Defined ID: Q43774 Cross Reference Number:
Topic: Q43774 ARM RU-31 Num Field 1: 3.4 Num Field 2: 3.6 Text Field: 4.2 061 AK3.02 Comments: Proposed reference to be provided to applicant during examination:
NONE Technical
Reference:
74RM-9EF41, RMS Alarm Response Procedure.
K&A:
Knowledge of the reasons for the following responses as they apply to the Area Radiation Monitoring (ARM) System Alarms: Guidance contained in alarm response for ARM system.
Justification:
A. Incorrect - FBEVAS ONLY comes off the HIGH Alarm. The cross trip from FBEVAS initiated the CREFAS.
B. Correct - RU-31 HIGH Alarm initiated FBEVAS and FBEVAS will initiate a cross trip of CREFAS.
C. Incorrect - This is the opposite way of how the activations will occur.
D. Incorrect - An Alert Alarm will not actuate the CREFAS, only a HIGH Alarm.
Rev 0 Question 23 Table-Item Links Q - 10CFR Sections 55.41 (10) Administrative, normal, abnormal, and emergency operating procedures for the facility.
Q - Reactor Operator K & A Importance classification 3.4 Q - Cognitive Level Memory Q - Question Source New O - RO Tier Group Designations Tier 1 Group 2 Q - RO Exam 2010 RO NRC Associated objective(s):
Describe the operator's responsibility when acknowledging RMS alarms in accordance with 74RM-9EF41.
Rev 0 24 ID: Q8251 Points: 1.00 Given the following conditions:
Unit 1 is operating at 100% power.
Due to a hazardous gas condition, the SM and CRS have directed the evacuation of the control room.
The CRS directs you to take all initial operator actions prior to leaving the control room.
Which ONE of the following describes the appropriate action?
A. All RCPs must be tripped.
B. Ensure PLCS is in remote auto.
C. One RCP in each loop must be tripped.
D. Initiate a RPCB Large Load Reject from B04.
Answer: C
Rev 0 Question 24 Info Question Multiple Choice Type:
Status: Active Always select No on test?
Authorized No for practice?
Points: 1.00 Time to 3
Complete:
Difficulty: 4.00 System ID: 8251 User-Defined Q8251 ID:
Cross Reference 40AO-9ZZ18, SD OUTSIDE CR Number:
Topic: Q8251 AOP Immed actions following CR evac Num Field 1: 4.1 Num Field 2: 4.5 Text Field: 2.4.6 Comments: Proposed reference to be provided to applicant during examination: NONE Technical
Reference:
40AO-9ZZ18, Shutdown Outside Control Room K&A:
Knowledge of EOP mitigation strategies as it applies to Control Room Evacuation.
Justification:
A. Incorrect - Correct answer is trip 2 RCPs, candidate may think that tripping all 4 RCPs is required per step 2 of Shutdown outside the CR.
B. Incorrect - Step 4 states if time permits, place in PLCS master controller in Local Auto.
C. Correct - This is directed per step 2 of the procedure.
D. Incorrect - Step 1 states ensure the reactor is tripped and Step 3 states ensure Main Turbine is tripped, therefore the candidate could interpret this as Perform a RPCB.
Rev 0 Question 24 Table-Item Links Q - 10CFR Sections 55.41 (10) Administrative, normal, abnormal, and emergency operating procedures for the facility.
Q - Reactor Operator K & A Importance classification 3.7 Q - Cognitive Level Memory Q - Question Source New O - RO Tier Group Designations Tier 1 Group 2 Q - RO Exam 2010 RO NRC Associated objective(s):
Active Question Bank 2004 State the operator actions that are required to be performed prior to Control Room evacuation.
discuss the actions prior to Control Room evacuation for reasons other than a fire
Rev 0 25 ID: Q22624 Points: 1.00 Given the following conditions:
Unit 1 is operating at 100% power and stable.
PZR Level is 52% and stable.
2 charging pumps are operating.
Letdown flow is 56 gpm and lowering slowly.
RCS Pressure is 2225 psia and stable.
0 RCS Subcooling is 41 F and stable.
SG 1 level is 48% NR and slowly rising.
SG 2 level is 50% NR and stable.
CTMT Pressure is 0.4 psig and stable.
CTMT Humidity is rising slowly.
RU-1 CTMT Atmosphere monitor is slowly trending up.
RU- 139/140 Main Steam Line SG#1/#2 are stable.
Which ONE of the following describes the current event?
A. RCS Cold Leg Leak.
B. Stuck Open PZR Safety Valve.
C. Steam Generator Tube Rupture.
D. Excessive Steam Demand inside CTMT.
Answer: A
Rev 0 Question 25 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 3.00 System ID: 22624 User-Defined ID: Q22624 Cross Reference Number:
Topic: Q22624 Excessive RCS Leakage Location Num Field 1: 3.2 Num Field 2: 3.5 Text Field: 4.4 A16 AK2.1 Comments: Proposed reference to be provided to applicant during examination:
NONE Technical
Reference:
40EP-9EO03, LOCA K&A:
Knowledge of the interrelations between the (Excess RCS Leakage) and the following: Components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
Justification:
A. Correct - Charging/Letdown mismatch are indicative of a RCS Leak and the other conditions support RCS Cold Leg location.
B. Incorrect - Steam Space Leaks will cause PZR level to rise.
C. Incorrect - At power the Main Steam Line N-16 monitors would provide indication of a SGTR.
D. Incorrect - Rx Power and subcooling would rise on an ESD event due to the increase in steam demand resulting in a colder RCS.
Rev 0 Question 25 Table-Item Links Q - 10CFR Sections 55.41 (7) Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
Q - Reactor Operator K & A Importance classification 3.2 Q - Cognitive Level Comprehension / Anal Q - Question Source New O - RO Tier Group Designations Tier 1 Group 2 Q - RO Exam 2010 RO NRC Associated objective(s):
Given conditions of LOCA describe how the plant would respond to various types of RCS leaks in accordance with 40EP-9EO03.
Rev 0 26 ID: Q7654 Points: 1.00 Given the following conditions:
Unit 1 has tripped from 100% power.
AFA-P01(Essential Turbine Driven Aux Feed Pump) is out of service.
NAN-S03 and NAN-S04 are de-energized.
Emergency DG 'A' is out of service.
AFB-P01(Essential Motor Driven Aux Feed Pump) is controlling SG levels.
ADVs are being used to control SG pressures.
Which ONE of the following correctly identifies the Optimal Recovery Procedure to implement after the SPTAs are completed?
A. 40EP-9EO06 (LOAF).
B. 40EP-9EO08 (Blackout)
C. 40EP-9EO02 (Reactor Trip).
D. 40EP-9EO07 (LOOP/LOFC).
Answer: D
Rev 0 Question 26 Info Question Multiple Choice Type:
Status: Active Always select No on test?
Authorized No for practice?
Points: 1.00 Time to 2
Complete:
Difficulty: 3.00 System ID: 7654 User-Defined Q7654 ID:
Cross Reference Number:
Topic: Q7654 EOP Selecting proper ORP Num Field 1: 2.7 Num Field 2: 3.7 Text Field: 4.4 A13 AA2.1 Comments: Proposed reference to be provided to applicant during examination: NONE Technical
Reference:
40EP-9EO07, LOOP/LOFC K&A:
Ability to determine and interpret the following as they apply to the (Natural Circulation Operations) Facility conditions and selection of appropriate procedures during abnormal and emergency operations.
Justification:
A. Incorrect - AFB-P01 is controlling SG levels therefore LOAF entry conditions are not met.
B. Incorrect - PBB-S04 is energized by its EDG, therefore no Blackout condition exists.
C. Incorrect - Rx Trip Safety Functions not met due to not having a 13.8 kV AC energized therefore Rx Trip is not to be utilized.
D. Correct - NAN-S03 and S04 not being energized results in a loss of RCPs, therefore entry conditions for Loss of offsite power/Loss of forced circulation entry conditions are met.
Rev 0 Question 26 Table-Item Links Q - 10CFR Sections 55.41 (10) Administrative, normal, abnormal, and emergency operating procedures for the facility.
Q - Reactor Operator K & A Importance classification 2.7 Q - Cognitive Level Comprehension / Anal Q - Question Source PV Bank Not Modified O - RO Tier Group Designations Tier 1 Group 2 Q - RO Exam 2010 RO NRC Associated objective(s):
determine whether or not entry into the LOOP/LOFC EOP is appropriate
Rev 0 27 ID: Q10330 Points: 1.00 Given the following conditions:
You are making a tour of the Turbine Building.
You discover an Oil leak at Main Feed Pump B.
The oil has caught on FIRE.
You have made the required notifications.
In accordance with PVNGS plant policies and procedures which type of fire extinguisher is recommended for use on this class of fire?
A. CO2 extinguisher.
B. Pressurized water extinguisher.
C. Any type of fire extinguisher is acceptable.
D. ABC Multipurpose Dry Chemical extinguisher.
Answer: D Question 27 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 2.00 System ID: 10330 User-Defined ID: Q10330 Cross Reference Number:
Topic: Q10330 fire extinguishers and their uses Num Field 1: 3.4 Num Field 2: 3.7 Text Field: 4.2 067 AA1.08 Comments: Proposed reference to be provided to applicant during examination:
NONE Technical
Reference:
PVNGS Safety Manual, Site Access Training Handbook K&A: Ability to operate and / or monitor the following as they apply to the Plant Fire on Site: Fire fighting equipment used on each class of fire JUSTIFICATION:
A. Incorrect - This is used on electrical fires.
B. Incorrect - This is used on ash producing fires.
C. Incorrect - This is not in accordance with PVNGS procedures.
D. Correct - This is the correct extinguisher for this class of fire.
Rev 0 Question 27 Table-Item Links Q - 10CFR Sections 55.41 (7) Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
Q - Reactor Operator K & A Importance classification 3.4 Q - Cognitive Level Memory Q - Question Source New O - RO Tier Group Designations Tier 1 Group 2 Q - RO Exam RO Exam Question 2007 2010 RO NRC Associated objective(s):
FIRE EXTINGUISHERS AND THEIR USES
Rev 0 28 ID: Q43793 Points: 1.00 Which ONE of the following describes the NORMAL parameters associated with RCP seal water?
RCP Seal flow is provided by approximately ____(1)____ gpm from CVCS, ____(2)____
gpm from RCS journal bearing lower cavity with ____(3)____ gpm discharging to seal bleed off.
A. (1) 6.6 (2) 16.0 (3) 2.2 B. (1) 16.0 (2) 6.6 (3) 2.2 C. (1) 6.6 (2) 16.0 (3) 9.5 D. (1) 16.0 (2) 6.6 (3) 9.5 Answer: A Question 28 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 2.00 System ID: 43793 User-Defined ID: Q43793 Cross Reference Number:
Topic: Q43793 RCP SEAL FLOW RATES Num Field 1: 2.6 Num Field 2: 2.9 Text Field: 3.4 003 K1.04 Comments: Proposed reference to be provided to applicant during examination: NONE Technical
Reference:
Operator Information Manual, LOIT System Lesson Plans K&A:
Knowledge of the physical connections and/or cause-effect relationships between the RCPs and the following systems: CVCS.
Justification:
A. Correct - the 6.6 gpm from the seal injection portion of CVCS joins the 16 gpm from the RCS journal bearing lower cavity at the jet pump. From there it is goes to the HPSC where it is divided with 18.6 gpm is routed down through the aux impeller and the remainder moving upward to cool the seal face eventually goes to the VCT via seal bleed off.
B. Incorrect - Flow rates are inversed. CVCS is 6.6 vice 16.0 for RCS.
C. Incorrect - Flow rates for seal flow are correct, seal bleed off max spec is 9.5.
D. Incorrect - Flow rates are inversed. CVCS is 6.6 vice 16.0 for RCS, seal bleed off max spec is 9.5.
Rev 0 Question 28 Table-Item Links Q - 10CFR Sections 55.41 (3) Mechanical components and design features of the reactor primary system.
Q - Reactor Operator K & A Importance classification 2.6 Q - Cognitive Level Memory Q - Question Source New O - RO Tier Group Designations Tier 2 Group 1 Q - RO Exam 2010 RO NRC Associated objective(s):
Describe how the Reactor Coolant System is supported by the following systems:
CVCS.
Rev 0 29 ID: Q22665 Points: 1.00 Given the following conditions:
Unit 1 is operating at 100% power.
Regenerative HX outlet Temperature is 410 0F and rising at 5 0F/min.
Assuming NO operator action, which ONE of the following describes the impact on PZR Level and pressure over the next two minutes?
PZR level will ___(1)___ and PZR pressure will ___(2)___.
A. (1) rise (2) lower.
B. (1) lower (2) lower.
C. (1) rise (2) remain constant or rise slightly.
D. (1) lower (2) remain constant or rise slightly.
Answer: C
Rev 0 Question 29 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 3.00 System ID: 22665 User-Defined ID: Q22665 Cross Reference Number:
Topic: Q22665 PZR LEVEL/PRESS Response to CVCS malfunction.
Num Field 1: 3.8 Num Field 2: 4.1 Text Field: 3.3 004 K3.07 Comments: Proposed reference to be provided to applicant during examination:
NONE Technical
Reference:
40AL-9RK4A, LD SYS TRBL Alarm Response Procedure K&A:
Knowledge of the effect that a loss or malfunction of the CVCS will have on the following: PZR level and pressure.
Justification:
A. Incorrect - Letdown will isolate at 413 0F which will cause PZR level to rise, letdown has no effect on pressure.
B. Incorrect - Letdown will isolate at 413 0F which will cause PZR level to rise, letdown has no effect on pressure. The setpoint was recently changed from 4500F to 413 0F.
C. Correct - Letdown will isolate at 413 0F which will cause PZR level to rise, as level goes up the bubble will get compressed so pressure may rise slightly.
D. Incorrect - Letdown will isolate at 413 0F which will cause PZR level to rise, as level goes up the bubble will get compressed so pressure may rise slightly. The setpoint was recently changed from 4500F to 413 0F.
Rev 0 Question 29 Table-Item Links Q - 10CFR Sections 55.41 (7) Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
Q - Reactor Operator K & A Importance classification 3.8 Q - Cognitive Level Comprehension / Anal Q - Question Source New O - RO Tier Group Designations Tier 2 Group 1 Q - RO Exam 2010 RO NRC Associated objective(s):
Explain the operation of the following Letdown Isolation Valves, including their Control Room controls, under normal operating conditions:
- Letdown Line to Regen Heat Exch Vlv (CHB-UV-515)
- Letdown Line to Regen Heat Exch Vlv Containment Isolation Vlv (CHA-UV-516)
- Regen Heat Exch to Letdown Heat Exch Containment Isolation Vlv (CHB-UV-523)
Rev 0 30 ID: Q22671 Points: 1.00 Given the following conditions:
Unit 1 is in Mode 4 at the beginning of a refueling outage.
Shutdown cooling (SDC) is in service with the A train.
SIA-HV-306 (SDCHX A Bypass valve) Control Switch develops a short causing the valve to CLOSE.
Which ONE of the following describes the effect on the shutdown cooling system?
RCS Tcold will ____(1)____ and SDC flow will ____(2)____.
A. (1) rise (2) rise B. (1) rise (2) lower C. (1) lower (2) rise D. (1) lower (2) lower Answer: D
Rev 0 Question 30 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 4.00 System ID: 22671 User-Defined ID: Q22671 Cross Reference Number:
Topic: Q22671 SDC MALF SDCHX Num Field 1: 2.5 Num Field 2: 2.6 Text Field: 3.4 005 K6.03 Comments: Proposed reference to be provided to applicant during examination:
NONE Technical
Reference:
40OP-9SI01, SDC Initiation K&A:
Knowledge of the effect of a loss or malfunction on the following will have on the RHRS: RHR heat exchanger.
Justification:
A. Incorrect - Candidate must understand the physical arrangement of the SDCHX and the normal Mode 4 operating lineup with decay heat still a factor.
B. Incorrect - Candidate must understand the physical arrangement of the SDCHX and the normal Mode 4 operating lineup with decay heat still a factor.
C. Incorrect - Candidate must understand the physical arrangement of the SDCHX and the normal Mode 4 operating lineup with decay heat still a factor.
D. Correct - SDC flow is maintained by throttling both the bypass and the outlet valves of the SDCHX. In this case the bypass failing closed will force all the flow through the HX which will lower Tcold. At the same time, the outlet valve is throttled in relationship to the bypass valve, therefore SDC flow will lower.
Rev 0 Question 30 Table-Item Links Q - 10CFR Sections 55.41 (7) Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
Q - Reactor Operator K & A Importance classification 2.5 Q - Cognitive Level Comprehension / Anal Q - Question Source New O - RO Tier Group Designations Tier 2 Group 1 Q - RO Exam 2010 RO NRC Associated objective(s):
Describe the flowpath/s associated with the SDC system to include these major components:
- Safety Injection Pumps* Refueling Water Tank
- Shutdown Cooling heat exchangers *injection valves
Rev 0 31 ID: Q22673 Points: 1.00 Which ONE of the following describes the power supply to SIA-P01 (LPSI Pump A)?
A. PBA-S03.
B. PGA-L31.
C. NBN-S01.
D. NAN-S03.
Answer: A Question 31 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 1 Difficulty: 2.00 System ID: 22673 User-Defined ID: Q22673 Cross Reference Number:
Topic: Q22673 LPSI PUMP A POWER SUPPLY Num Field 1: 3.0 Num Field 2: 3.2 Text Field: 3.4 005 K2.01 Comments: Proposed reference to be provided to applicant during examination: NONE Technical
Reference:
Power Supply Table K&A:
Knowledge of bus power supplies to the following: RHR pumps.
Justification:
A. Correct - PBA-S03 is the correct 4160V power supply.
B. Incorrect - PGA-L31 is a class 480V load center supplied by PBA-S03.
C. Incorrect - NBN-S01 is the non class 4160V bus that supplies PBA-S03.
D. Incorrect - NAN-S03 is the non class 13.8kV bus that supplies NBN-S01.
Rev 0 Question 31 Table-Item Links Q - 10CFR Sections 55.41 (7) Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
Q - Reactor Operator K & A Importance classification 3.0 Q - Cognitive Level Memory Q - Question Source New O - RO Tier Group Designations Tier 2 Group 1 Q - RO Exam 2010 RO NRC Associated objective(s):
Identify the power supplies to SI related equipment.
Rev 0 32 ID: Q22674 Points: 1.00 Given the following conditions:
Unit 1 is operating at 100% power.
An inadvertent CSAS has occurred on Train "A".
The Operating Crew has stopped all Safety Injection pumps that started per 40AO-9ZZ17 (Inadvertent PPS ESFAS Actuations).
The CSAS will not reset.
As I&C is troubleshooting the CSAS problem, a LOCA occurs resulting in a SIAS.
Which ONE of the following describes the status of the 'A' CS Pump?
The 'A' CS pump is A. operating. No action required.
B. anti-pumped. The control switch can be taken to START two times in order to start it.
C. overridden off. The control switch can first be taken to STOP and then back to NORMAL to start it.
D. anti-pumped. It can NOT be operated from the control room, DC control power must be cycled locally to start it.
Answer: B
Rev 0 Question 32 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 3.00 System ID: 22674 User-Defined ID: Q22674 Cross Reference Number:
Topic: Q22674 INADVERTANT CS RESPONSE TO SIAS Num Field 1: 4.5 Num Field 2: 4.8 Text Field: 3.2 006 A2.12 Comments: Proposed reference to be provided to applicant during examination: NONE Technical
Reference:
40AO-9ZZ17, Inadvertent ESFAS Actuations K&A:
Ability to (a) predict the impacts of the following malfunctions or operations on the ECCS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Conditions requiring actuation of ECCS.
Justification:
A. Incorrect - The CS pump has been anti-pumped per the AOP, therefore the SIAS signal will not start the CS pump.
B. Correct - The CS pump has been anti-pumped per the AOP, therefore to reset the pump relays you must override the SIAS with the first positioning to start, then take it to start again to start the pump.
C. Incorrect - The CS pump has been anti-pumped per the AOP, therefore taking the switch to stop just re-enforces the anti pump condition. The positioning of the switch to the start next will bring in the override light but the pump will remain off.
D. Incorrect - The CS pump has been anti-pumped per the AOP, cycling the DC control power will start the pump, but the pump may still be operated from the control room. Local operation is not the only method to start the pump.
Rev 0 Question 32 Table-Item Links Q - 10CFR Sections 55.41 (5) Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.
Q - Reactor Operator K & A Importance classification 4.5 Q - Cognitive Level Comprehension / Anal Q - Question Source New O - RO Tier Group Designations Tier 2 Group 1 Q - RO Exam 2010 RO NRC Associated objective(s):
Given conditions of an inadvertent CSAS actuation describe the conditions necessary to restart a CS pump that has been stopped in accordance with 40AO-9ZZ17.
Rev 0 33 ID: Q22675 Points: 1.00 Given the following conditions:
Unit 1 is operating at 100% power.
An inadvertent SIAS results in the 'A' CS pump running.
Which ONE of the following describes how long CS 'A' is procedurally limited to operate in this condition?
A. 30 minutes.
B. 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
C. 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
D. Indefinitely.
Answer: B
Rev 0 Question 33 Info Question Multiple Choice Type:
Status: Active Always select No on test?
Authorized for No practice?
Points: 1.00 Time to 2
Complete:
Difficulty: 2.00 System ID: 22675 User-Defined Q22675 ID:
Cross Reference Number:
Topic: Q22675 ECCS MINI FLOW REQUIREMENTS Num Field 1: 2.7 Num Field 2: 3.1 Text Field: 3.2 006 K4.06 Comments: Proposed reference to be provided to applicant during examination: NONE Technical
Reference:
40OP-9SI01, SDC Initiation K&A:
Knowledge of ECCS design feature(s) and/or interlock(s) which provide for the following: Recirculation of minimum flow through pumps.
Justification:
A. Incorrect - 30 minutes is not a time limit associated with CS pumps.
B. Correct - CS pumps can operate for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> when flows are between 150 and 2100 gpm.
C. Incorrect - Recirculation Flow is approximately 158 - 228 gpm, therefore the 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> limit applies.
D. Incorrect - Indefinite operation is allowed at flows > 2100 gpm.
Rev 0 Question 33 Table-Item Links Q - 10CFR Sections 55.41 (7) Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
Q - Reactor Operator K & A Importance classification 2.7 Q - Cognitive Level Memory Q - Question Source PV Bank Not Modified O - RO Tier Group Designations Tier 2 Group 1 Q - RO Exam 2010 RO NRC Associated objective(s):
Describe the design characteristics of the LPSI and CS pumps.
Rev 0 34 ID: Q22676 Points: 1.00 Given the following conditions:
You are the RO in Unit 2.
You have been directed to pump the Reactor Drain Tank (RDT) to the CVCS HUT.
Initial RDT level is 72%
You commenced pumping the RDT at time 0100.
Level is lowering at 2% / minute.
Which ONE of the following describes the latest time the evolution can be stopped and still maintain the required quench volume in the RDT?
Stop the evolution at time A. 0107.
B. 0109.
C. 0112.
D. 0114.
Answer: B
Rev 0 Question 34 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 3.00 System ID: 22676 User-Defined ID: Q22676 Cross Reference Number:
Topic: Q22676 RDT Minimum Quench Volume Num Field 1: 2.9 Num Field 2: 3.1 Text Field: 3.5 007 A1.01 Comments: Proposed reference to be provided to applicant during examination:
NONE Technical
Reference:
40OP-9CH01, CVCS Normal Operations.
K&A:
Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the PRTS controls including: Maintaining quench tank water level within limits.
Justification:
A. Incorrect - (7 min)(2%/min) = 14 %. 72% -14% =58%. This is > 52%
minimum level requirement but higher than the answer of 54%.
B. Correct - (9 min)(2%/min) = 18 %. 72% -18% =54%. This is > 52%
minimum level requirement.
C. Incorrect - (12 min)(2%/min) = 24 %. 72% -24% =48%. This is < 52%
minimum level requirement.
D. Incorrect - (14 min)(2%/min) = 28%. 72% -28% =44%. This is < 52%
minimum level requirement.
Rev 0 Question 34 Table-Item Links Q - 10CFR Sections 55.41 (5) Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.
Q - Reactor Operator K & A Importance classification 2.9 Q - Cognitive Level Comprehension / Anal Q - Question Source New O - RO Tier Group Designations Tier 2 Group 1 Q - RO Exam 2010 RO NRC Associated objective(s):
Explain the operation of the Reactor Drain Tank (CHN-X02) under normal operating conditions.
Rev 0 35 ID: Q43775 Points: 1.00 Given the following conditions:
Initial Conditions:
Unit 1 is operating at 100% power.
Both Nuclear Cooling Water (NC) Pumps have faulted.
EW 'A' has been cross tied to supply NC.
Subsequently:
RU-2 (EW 'A') alarms in HIGH.
Which ONE of the following describes the cause and impact on the system(s)?
A. RCP HPSC leak. NC ONLY is contaminated.
B. Letdown HX leak. NC ONLY is contaminated.
C. RCP HPSC leak. NC and EW are contaminated.
D. Letdown HX leak. NC and EW are contaminated.
Answer: C
Rev 0 Question 35 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 2.00 System ID: 43775 User-Defined ID: Q43775 Cross Reference Number:
Topic: Q43775 NC RMS Alarms Num Field 1: 3.3 Num Field 2: 3.5 Text Field: 3.8 008 A2.04 Comments: Proposed reference to be provided to applicant during examination: NONE Technical
Reference:
74RM-9EF41, RMS Alarm Response Procedure, System Diagrams.
K&A:
Ability to (a) predict the impacts of the following malfunctions or operations on the CCWS, and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: PRMS alarm.
Justification:
A. Incorrect - When NC and EW are X-connected RCPs are an essential load. NC and EW are contaminated by the leak.
B. Incorrect - Letdown is isolated when NCV-99 is closed. NC and EW are contaminated by the leak.
C. Correct -When NC and EW are X-connected RCPs are an essential load, only RU-2 is in the system. NC piping is still used and NC is contaminated, RU-6 the NC RU monitor is isolated by NCV-99 D. Incorrect - Letdown is isolated when NCV-99 is closed.
Rev 0 Question 35 Table-Item Links Q - 10CFR Sections 55.41 (5) Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.
Q - Reactor Operator K & A Importance classification 3.3 Q - Cognitive Level Comprehension / Anal Q - Question Source New O - RO Tier Group Designations Tier 2 Group 1 Q - RO Exam 2010 RO NRC Associated objective(s):
Explain the operation of the NC Radiation Monitoring under normal operating conditions.
Rev 0 36 ID: Q22677 Points: 1.00 Given the following conditions:
Unit 1 is operating at 100% power.
TCN-P01A (TC Pump 'A') is under permit for performance of a bearing lubrication PM.
TCN-P01B (TC Pump 'B') experiences an electrical fault.
Which ONE of the following components will experience a temperature rise?
A. Gas Stripper.
B. Air Removal Pumps.
C. Letdown Heat Exchanger.
D. Instrument Air Compressors.
Answer: D Question 36 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 2.00 System ID: 22677 User-Defined ID: Q22677 Cross Reference Number:
Topic: Q22677 TC LOADS Response to a loss of TCW Num Field 1: 3.4 Num Field 2: 3.5 Text Field: 3.8 008 K3.01 Comments: Proposed reference to be provided to applicant during examination:
NONE Technical
Reference:
LOIT System Lesson Plans K&A:
Knowledge of the effect that a loss or malfunction of the CCWS will have on the following: Loads cooled by CCWS.
Justification:
A. Incorrect - Gas Stripper is cooled by Nuclear Cooling Water.
B. Incorrect - Air Removal Pumps are cooled by Plant Cooling Water.
C. Incorrect - Letdown HX is cooled by Nuclear Cooling Water.
D. Correct - IA compressors are cooled by Turbine Cooling Water.
Rev 0 Question 36 Table-Item Links Q - 10CFR Sections 55.41 (4) Secondary coolant and auxiliary systems that affect the facility.
Q - Reactor Operator K & A Importance classification 3.4 Q - Cognitive Level Memory Q - Question Source New O - RO Tier Group Designations Tier 2 Group 1 Q - RO Exam 2010 RO NRC Associated objective(s):
Describe the response to a failure of the Turbine Cooling Water System.
Rev 0 37 ID: Q22678 Points: 1.00 Given the following conditions:
Unit 1 is operating at 100% power.
Pressurizer Pressure Control System (PPCS) is in automatic.
PZR Pressure is 2250 psia and rising slowly.
Which ONE of the following describes the expected response of the PPCS as pressure continues to rise?
A. All PZR heaters will de-energize at 2285 psia.
B. PPCS will send a signal to the Plant Protection System to initiate a Reactor Trip at 2383 psia.
C. PZR back-up heaters will de-energize at 2285 psia and proportional heaters will de-energize at 2300 psia.
D. PZR proportional heaters will de-energize at 2285 psia and back-up heaters will de-energize at 2300 psia.
Answer: A
Rev 0 Question 37 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 2.00 System ID: 22678 User-Defined ID: Q22678 Cross Reference Number:
Topic: Q22678 PPCS Overpressure Protection Num Field 1: 3.8 Num Field 2: 4.1 Text Field: 3.3 010 K4.03 Comments: Proposed reference to be provided to applicant during examination:
NONE Technical
Reference:
Operator Information Manual, PPCS Lesson Plan K&A:
Knowledge of PZR PCS design feature(s) and/or interlock(s) which provide for the following: Over pressure control.
Justification:
A. Correct - High pressure bistable will de-energize all heaters at 2285 psia.
B. Incorrect - PPS receives its pressure input directly from the PZR Pressure Transmitters.
C. Incorrect - High pressure bistable will de-energize all heaters, not just the back up heaters, at 2285 psia.
D. Incorrect - High pressure bistable will de-energize all heaters, not just the proportional heaters, at 2285 psia.
Rev 0 Question 37 Table-Item Links Q - 10CFR Sections 55.41 (7) Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
Q - Reactor Operator K & A Importance classification 3.8 Q - Cognitive Level Memory Q - Question Source New O - RO Tier Group Designations Tier 2 Group 1 Q - RO Exam 2010 RO NRC Associated objective(s):
Describe the automatic features associated with the Pressurizer Pressure Control System Bistables.
Rev 0 38 ID: Q10462 Points: 1.00 Given the following conditions:
Group 5 CEAs are inserted from 140" to 125" withdrawn.
Reactor Power is maintained constant at 100% by dilution.
As CEAs insert, the CPCs Radial Peaking Factors (RPFs) will ...
A. rise and Local Power Density (LPD) margin to trip rises.
B. lower and Local Power Density (LPD) margin to trip rises.
C. rise and Local Power Density (LPD) margin to trip lowers.
D. lower and Local Power Density (LPD) margin to trip lowers.
Answer: C Question 38 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 3.00 System ID: 10462 User-Defined ID: Q10462 Cross Reference Number:
Q10462 SYS CEAs inserting, CPCs respond by RPF increase & LPD Topic:
margin decrease Num Field 1: 3.1 Num Field 2: 3.3 Text Field: 3.7 012 K5.02 Comments: Proposed reference to be provided to applicant during examination:
NONE Technical
Reference:
LOIT Lesson Plan K&A:
Knowledge of the operational implications of the following concepts as they apply to the RPS: Power density.
Justification:
A. Incorrect - As RPFs rise, this will cause the margin to become smaller.
B. Incorrect - RPFs rise on CEA insertions.
C. Correct - For a given Reactor Power, CEA insertion will cause power density to rise. This is accounted for with Radial Peaking Factors that are applied to DNBR and LPD calculations.
D. Incorrect - RPFs rise on CEA insertions.
Rev 0 Question 38 Table-Item Links Q - 10CFR Sections 55.41 (5) Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.
Q - Reactor Operator K & A Importance classification 3.1 Q - Cognitive Level Comprehension / Anal Q - Question Source PV Bank Not Modified O - RO Tier Group Designations Tier 2 Group 1 Q - RO Exam 2010 RO NRC Associated objective(s):
Explain the operation of the LPD Calculation in the Core Protection Calculators.
Rev 0 39 ID: Q43763 Points: 1.00 Given the following conditions:
Unit 2 is operating at 100% power.
RCS Pressure is 2230 psia.
The RO performs the following actions on the Control Room Handswitch for a group of non-class pressurizer backup heaters (RCN-HS-100-9) on B04.
Turns the handswitch to OFF.
Turns the handswitch to ON.
Which ONE of the following describes the status of this group of pressurizer backup heaters?
The pressurizer backup heaters controlled by RCN-HS-100-9 are ____(1)____, and it's SETPOINT OVERRIDE light is ____(2)____.
A. (1) ON (2) ON B. (1) ON (2) OFF C. (1) OFF (2) ON D. (1) OFF (2) OFF Answer: A
Rev 0 Question 39 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 2.00 System ID: 43763 User-Defined ID: Q43763 Cross Reference Number:
Topic: Q43763 PZR BACKUP HEATER SETPOINT OVERRIDE Num Field 1: 3.6 Num Field 2: 3.4 Text Field: 3.3 010 A4.02 Comments: Proposed reference to be provided to applicant during examination: NONE Technical
Reference:
LOIT PPCS Lesson Plan K&A:
Ability to manually operate and/or monitor in the control room: PZR heaters.
Justification:
A. Correct - Setpoint override is enabled when PZR pressure is >
2200 psia therefore by taking the handswitch to OFF it will enable the setpoint override circuit therefore the light will be lit. Taking the handswitch to ON after picking up the setpoint override circuit will energize the heaters ON until the high pressure bistable is met or the handswitch is taken to OFF. This explanation will prove all distracters Incorrect.
B. Incorrect - The heaters and the light will both be ON. Candidate may not know the override setpoint is 2200 psia. May also confuse the Setpoint Override with the SIAS Override.
C. Incorrect - The heaters and the light will both be ON. Candidate may not know the override setpoint is 2200 psia. May also confuse the Setpoint Override with the SIAS Override.
D. Incorrect - The heaters and the light will both be ON. Candidate may not know the override setpoint is 2200 psia. May also confuse the Setpoint Override with the SIAS Override.
Rev 0 Question 39 Table-Item Links Q - 10CFR Sections 55.41 (7) Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
Q - Reactor Operator K & A Importance classification 3.6 Q - Cognitive Level Comprehension / Anal Q - Question Source PV Bank Not Modified O - RO Tier Group Designations Tier 2 Group 1 Q - RO Exam 2010 RO NRC Associated objective(s):
Describe the Control Room controls associated with the Pressurizer Backup Heaters including indications.
Rev 0 40 ID: Q1409 Points: 1.00 Given the following conditions:
Unit 1 is operating at 100% power.
An inadvertent CSAS occurs.
The CRS has directed the restoration of NCW to containment, overriding valves as necessary.
Which ONE of the following describes the steps (in correct sequence) that must be taken to restore NCW to the containment?
A. Override and open the supply valve (NCB-UV-401), then override and open the return valves (NCA-UV-402 and NCB-UV-403).
B. Override and open the return valves (NCA-UV-402 and NCB-UV-403), then override and open the supply valve (NCB-UV-401).
C. Open NC CTMT Return Hdr Isol Vlv (NCN-UV-99), then Override and open the return valves (NCA-UV-402 and NCB-UV-403), then override and open the supply valve (NCB-UV-401).
D. Override and open the instrument air to containment isolation valve (IAA-UV-2) then Override and open the supply valve (NCB-UV-401), then override and open the return valves (NCA-UV-402 and NCB-UV-403).
Answer: B
Rev 0 Question 40 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 2.00 System ID: 1409 User-Defined ID: Q1409 Cross Reference Number:
Topic: Q1409 NC to containment after Inadvertant CSAS Num Field 1: 4.1 Num Field 2: 4.2 Text Field: 3.2 013 A3.02 Comments: Proposed reference to be provided to applicant during examination:
NONE Technical
Reference:
40AO-9ZZ17 Inadvertent PPS-ESFAS Actuation.
K&A:
Ability to monitor automatic operation of the ESFAS including: Operation of actuated equipment.
Justification:
A. Incorrect - Opening of the NC CTMT supply valve prior to the return valves may result in the lifting of the NC reliefs in CTMT B. Correct - Opening of the NC CTMT return valves prior to the supply valves minimizes the chance of lifting of the NC reliefs in CTMT.
C. Incorrect - NC-UV-99 does not close on a CSAS.
D. Incorrect - NC CTMT isolation valves are not air operated. The next step in the procedure directs restoring IA to CTMT and directs closing specific valves to prevent inadvertent opening.
Rev 0 Question 40 Table-Item Links Q - 10CFR Sections 55.41 (7) Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
Q - Reactor Operator K & A Importance classification 4.1 Q - Cognitive Level Memory Q - Question Source PV Bank Not Modified O - RO Tier Group Designations Tier 2 Group 1 Q - RO Exam 2010 RO NRC Associated objective(s):
Given conditions of an inadvertent CSAS actuation describe the method of restoring NC to containment, including the reasons why this is important in accordance with 40AO-9ZZ17.
Rev 0 41 ID: Q4649 Points: 1.00 Which ONE of the following describes the BOP-ESFAS power supply arrangement?
Each cabinet receives power from its associated train of...
A. PK and PN, which are auctioneered so that a loss of either supply to the cabinet will NOT impact the operation of the system.
B. PK ONLY, a loss of this power supply to the cabinet will cause an actuation of the components controlled by that train of BOP-ESFAS.
C. PN ONLY, a loss of this power supply to the cabinet will cause an actuation of the components controlled by that train of BOP-ESFAS.
D. PK and PN, a loss of either power supply to the cabinet will prevent those components controlled by the affected BOP-ESFAS cabinet from actuating when needed.
Answer: A
Rev 0 Question 41 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 3.00 System ID: 4649 User-Defined ID: Q4649 Cross Reference Number:
Topic: Q4649 BOP ESFAS Power Supply arrangement Num Field 1: 3.6 Num Field 2: 3.8 Text Field: 3.2 013 K2.01 Comments: Proposed reference to be provided to applicant during examination:
NONE Technical
Reference:
Power Supply Table.
K&A:
Knowledge of bus power supplies to the following: ESFAS/safeguards equipment control.
Justification:
A. Correct - Cabinets are powered by PK and PN and are auctioneered, therefore on a loss on one of the power supplies the other will supply power..
B. Incorrect - PK is only one of the two power supplies, therefore due to the auctioneered power supply arrangement a loss of one does not impact the system.
C. Incorrect - PN is only one of the two power supplies, therefore due to the auctioneered power supply arrangement a loss of one does not impact the system.
D. Incorrect - A loss of either will not impact the system.
Rev 0 Question 41 Table-Item Links Q - 10CFR Sections 55.41 (7) Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
Q - Reactor Operator K & A Importance classification 3.6 Q - Cognitive Level Memory Q - Question Source PV Bank Not Modified O - RO Tier Group Designations Tier 2 Group 1 Q - RO Exam 2010 RO NRC Associated objective(s):
Explain the operation of the BOP ESFAS Cabinets under normal operating conditions.
Rev 0 42 ID: Q43776 Points: 1.00 Given the following conditions:
Unit 1 is operating at 100% power.
A break in the Normal Chilled Water supply header has caused a loss of normal chilled water.
Which ONE of the following temperatures will be directly affected?
A. CEDMs.
B. Fuel Building.
C. Generator H2 Cooler.
D. Containment Atmosphere.
Answer: D Question 42 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 2.00 System ID: 43776 User-Defined ID: Q43776 Cross Reference Number:
Topic: Q43776 WC cools CTMT Normal ACUs Num Field 1: 3.2 Num Field 2: 3.3 Text Field: 3.5 022 A1.04 Comments: Proposed reference to be provided to applicant during examination:
NONE Technical
Reference:
LOIT System Lesson Plan K&A:
Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the CCS controls including: Cooling water flow.
Justification:
A. Incorrect - CEDM ACUs are cooled by Nuclear Cooling Water (NC)
B. Incorrect - Fuel Building AHUs do not use a cooling medium. They are an evaporative cooler that uses an air washer supplied by Domestic Water.
C. Incorrect - WC cools the Generator Collector Housing not the H2 coolers.
D. Correct - Containment Normal ACUs are cooled by Normal Chill Water (WC).
Rev 0 Question 42 Table-Item Links Q - 10CFR Sections 55.41 (5) Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.
Q - Reactor Operator K & A Importance classification 3.2 Q - Cognitive Level Memory Q - Question Source New O - RO Tier Group Designations Tier 2 Group 1 Q - RO Exam 2010 RO NRC Associated objective(s):
Describe how the Containment Building HVAC System is supported by the following systems:
- Nuclear Cooling Water (NC)
- Normal Chilled Water (WC)
- Engineered Safety Features Actuation System (ESFAS)
- Balance of Plant-Engineered Safety Features Actuation System (BOP-ESFAS)
Rev 0 43 ID: Q43777 Points: 1.00 Given the following conditions:
Unit 1 has tripped from 100% power Pressurizer Pressure is 1100 psia and lowering slowly.
Pressurizer Level is 13% and lowering.
RCS subcooling is 5 0F and lowering.
CTMT Pressure is 8.9 psig and rising.
NBN-X03 (ESF Transformer A) has faulted.
EDG B has a Hi Priority TRBL Alarm (Jacket Water Temperature High)
SPA-P01 (Essential Spray Pond Pump A) has an 86 lockout.
EDG A has been emergency stopped.
5 minutes into the event, B02 pump ammeters are as follows:
HPSI B 115 amps.
LPSI B 75 amps.
CS B 35 amps.
Which ONE of the following is true?
A. SBOGs are available to meet MVAC.
B. EDG B is required to be emergency stopped.
C. Containment Spray System is below minimum flow.
D. High Pressure Safety Injection is below minimum flow.
Answer: C
Rev 0 Question 43 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 3.00 System ID: 43777 User-Defined ID: Q43777 Cross Reference Number:
Topic: Q43777 CS Pump Failure Num Field 1: 3.9 Num Field 2: 4.2 Text Field: 3.5 026 A2.04 Comments: Proposed reference to be provided to applicant during examination: NONE Technical
Reference:
40EP-9EO01, SPTAs, 41AL-1RK1A DG A HI Priority TRBL K&A:
Ability to (a) predict the impacts of the following malfunctions or operations on the CSS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Failure of spray pump.
Justification:
A. Incorrect - SBOGs must be brought on to PBA-S03 which is unavailable due to the ESF transformer fault.
B. Incorrect - Only Overspeed, Gen Differential and Low Lube Oil Press will trip the EDG in emergency mode. High Jacket Water Temp would result in a trip if the EDG was running in any mode other than emergency.
C. Correct - Normal Pump Amperage is 90 - 100 amps. Sheared shaft amps are 30 - 40 amps. Therefore with a degraded CS B pump and no power available to CS A pump, there is no flow which is less than the minimum 4350 gpm.
D. Incorrect - HPSI pump amps are normal and flow will meet the 1 HPSI pump curve. Appendix 2 figures has flow curves for both 1 and 2 pump configurations.
Rev 0 Question 43 Table-Item Links Q - 10CFR Sections 55.41 (5) Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.
Q - Reactor Operator K & A Importance classification 3.9 Q - Cognitive Level Comprehension / Anal Q - Question Source New O - RO Tier Group Designations Tier 2 Group 1 Q - RO Exam 2010 RO NRC Associated objective(s):
Given plant conditions following a reactor trip analyze whether the Containment Temperature, Pressure and Combustible Gas Control Safety Function is met and what contingency actions are required if it is not in accordance with 40EP-9EO01.
Rev 0 44 ID: Q43752 Points: 1.00 Given the following conditions:
Unit 1 is operating at 100% power.
SBCS valve 1007 drifts open.
Which ONE of the following describes the plant response?
A. TLI will drop below Tavg.
B. An AMI will be generated.
C. A Quick Open Block will be generated.
D. Main Condenser Backpressure will degrade.
Answer: A
Rev 0 Question 44 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 4 Difficulty: 4.00 System ID: 43752 User-Defined ID: Q43752 Cross Reference Number:
Topic: Q43752 SBCS Plant Response Num Field 1: 3.3 Num Field 2: 3.3 Text Field: 3.4 039 K1.02 Comments: Proposed reference to be provided to applicant during examination:
NONE Technical
Reference:
Operator Information Manual K&A:
Knowledge of the physical connections and/or cause-effect relationships between the MRSS and the following systems: Atmospheric relief dump valves.
Justification:
A. Correct - TLI is essentially Turbine 1st Stage Pressure, therefore as steam flow goes up as the SBCS valve to atmosphere opens. TLI will lower and Tavg will be > Tref.
B. Incorrect - AMI is automatic motion inhibit. It is used when there is a mismatch between the Reactor Regulating System inputs. In this case the change in where the steam goes (Condenser via Main Turbine or Atmosphere via SBCS 1007) will affect both RRS inputs the same so no AMI.
C. Incorrect - AWP is an automatic withdrawal prohibit, which is enabled when Tavg is greater than Tref to prevent CEAs from stepping out. In this case Tavg < Tref and CEAs are required to step out to raise Tavg.
D. Incorrect - Main Condenser back pressure will not degrade as the steam flow will be directed to atmosphere not the main condenser.
Rev 0 Question 44 Table-Item Links Q - 10CFR Sections 55.41 (7) Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
Q - Reactor Operator K & A Importance classification 3.3 Q - Cognitive Level Comprehension / Anal Q - Question Source New O - RO Tier Group Designations Tier 2 Group 1 Q - RO Exam 2010 RO NRC Associated objective(s):
Describe the consequences of NOT taking CEDMCS out of AUTO with 1007 and 1008 open
Rev 1 45 ID: Q43765 Points: 1.00 Given the following conditions:
Unit 1 is operating at 100% power and stable.
The partial stroke exercise test for MSIV SGE-UV-170 has just been performed per 73ST-9SG01 (Train 'A').
The following annunciator is alarming in the Control Room.
(6A07A)
Which ONE of the following is the cause of the alarming condition?
A. MSIV valve position is at 50% travel.
B. MSFIS Logic cabinet HIGH temperature.
C. SG 1 MSIV UV-170 Hydr Accum 'A' Press Lo.
D. SG 1 MSIV UV-170 Hydr Accum Oil Rsvr Lvl Hi.
Answer: C
Rev 1 Question 45 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 3.00 System ID: 43765 User-Defined ID: Q43765 Cross Reference Number:
Topic: Q43765 MS ALARMS Num Field 1: 4.2 Num Field 2: 4.2 Text Field: 2.4.46 Comments: Proposed reference to be provided to applicant during examination:
NONE Technical
Reference:
40AL-9RK6A, B06A ARP K&A:
Ability to verify that the alarms are consistent with the plant conditions. As it applies to Main and Reheat Steam system.
Justification:
A. Incorrect - MSIV valve position is not an input to this alarm.
B. Incorrect - MFSIS Cabinet Loss of Power or Door Open will cause this alarm. There are numerous Cabinets that have high temp alarms.
C. Correct - This will result in the SG ISOL VLV TRBL alarm.
D. Incorrect - A low oil reservoir level will result in the alarm.
Rev 1 Question 45 Table-Item Links Q - 10CFR Sections 55.41 (10) Administrative, normal, abnormal, and emergency operating procedures for the facility.
Q - Reactor Operator K & A Importance classification 4.2 Q - Cognitive Level Comprehension / Anal Q - Question Source New O - RO Tier Group Designations Tier 2 Group 1 Q - RO Exam 2010 RO NRC Associated objective(s):
Describe the conditions required to generate the following annunciators:
- ADV CONTROL NOT IN CR POS
Rev 0 46 ID: Q4713 Points: 1.00 Which ONE of the following describes the start permissive interlock associated with AFN-P01 (Non-Essential Motor Driven Aux Feed Pump)?
The AFN-P01 start permissive interlock at a minimum is met when CTA-HV-1(AFN-P01 Suction Valve)...
A. and CTA-HV-4 (AFN-P01 Suction Valve) are both fully open.
B. and CTA-HV-4 (AFN-P01 Suction Valve) are both > 80% open.
C. and CTA-HV-4 (AFN-P01 Suction Valve) are both off their closed seats.
D. is > 80% open and CTA-HV-4 (AFN-P01 Suction Valve) is off its closed seat.
Answer: B Question 46 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 2.00 System ID: 4713 User-Defined ID: Q4713 Cross Reference Number:
Topic: Q4713 AF AFN suction valve position for start Num Field 1: 4.0 Num Field 2: 4.2 Text Field: 3. 4 061 K4.06 Comments: Proposed reference to be provided to applicant during examination:
NONE Technical
Reference:
LOIT AFW Lesson Plan K&A:
Knowledge of AFW design feature(s) and/or interlock(s) which provide for the following: AFW startup permissives.
Justification:
Incorrect - The interlock is both valves being > 80% open. This would satisfy the interlock seeing as both valves being full open is > 80%
open. At a minimum both valves being > 80% open is the correct answer.
Correct - The interlock is both valves being > 80% open.
Incorrect - The interlock is not met until both valves are > 80% open.
Incorrect - The interlock is not met until both valves are > 80% open.
Rev 0 Question 46 Table-Item Links Q - 10CFR Sections 55.41 (7) Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
Q - Reactor Operator K & A Importance classification 4.0 Q - Cognitive Level Memory Q - Question Source PV Bank Not Modified O - RO Tier Group Designations Tier 2 Group 1 Q - RO Exam 2010 RO NRC Associated objective(s):
Explain the operation of the Non-Essential AFP Suction Valves (CTA-UV-1)
(CTA-UV-4) under normal operating conditions.
Rev 0 47 ID: Q43816 Points: 1.00 Given the following conditions:
Unit 1 has tripped from 100% power.
SPTAs are complete and 40EP-9EO06 (LOAF) has been entered.
Assuming the Feedwater to the Steam Generators is not recovered and no further operator actions are taken, which ONE of the following conditions below is most likely?
A. RCS depressurized with HPSI flow to the RCS.
B. Pressurizer safety valves open and core uncovery.
C. RCS depressurized with HPSI and LPSI flow to the RCS.
D. RCS pressure controlled by main spray flow between 2200 and 2300 psia.
Answer: B
Rev 0 Question 47 Info Question Multiple Choice Type:
Status: Active Always select on No test?
Authorized No for practice?
Points: 1.00 Time to 2
Complete:
Difficulty: 4.00 System ID: 43816 User-Q43816 Defined ID:
Cross Reference Number:
Q43816 RCS response (temperature and pressure) as Steam Generator Topic:
dryout occurs Num Field 1: 3.6 Num Field 2: 3.8 Text Field: 3.4 059 K3.04 Comments: Proposed reference to be provided to applicant during examination:
NONE Technical
Reference:
LOIT Lesson Plan, UFSAR K&A:
Knowledge of the effect that a loss or malfunction of the MFW will have on the following: RCS.
Justification:
A. Incorrect - Pressure will rise and pressure will remain greater that HPSI pump shutoff head.
B. Correct - Loss of Feed is a heatup event due to the loss of SG Heat removal. RCS pressure will rise until PZR safeties lift and RCS inventory is lost. As Inventory is lost with Pressure remaining >
shutoff head of the HPSI and LPSI pumps, therefore they will not inject water, resulting in core uncovery.
C. Incorrect - Pressure will rise and pressure will remain greater that HPSI and LPSI pump shutoff head.
D. Incorrect - Main Spray will not be effective in reducing RCS pressure and it will continue to rise until safety valves lift.
Rev 0 Question 47 Table-Item Links Q - 10CFR Sections 55.41 (7) Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
Q - Reactor Operator K & A Importance classification 3.6 Q - Cognitive Level Comprehension / Anal Q - Question Source New O - RO Tier Group Designations Tier 2 Group 1 Q - RO Exam 2010 RO NRC Associated objective(s):
determine the RCS response (temperature and pressure) as Steam Generator dryout occurs
Rev 0 48 ID: Q10344 Points: 1.00 The RO is preparing to unload EDG 'A' per 40OP-9DG01, section 7 (Unloading DG "A").
Section 7.4.4 contains a note and caution stating that the DG output breaker should be opened immediately once DG load is reduced below 0.5 MW.
Which ONE of the following describes the possible impact of continuing to lower load?
A...
A. diesel generator differential trip may be activated.
B. diesel generator undervoltage trip may be activated.
C. reverse power trip of the DG output breaker may be activated.
D. negative sequence trip of the DG output breaker may be activated.
Answer: C
Rev 0 Question 48 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 3.00 System ID: 10344 User-Defined ID: Q10344 Cross Reference Number:
Topic: Q10344 EDG Minimum Load for Unloading.
Num Field 1: 3.4 Num Field 2: 3.8 Text Field: 3.6 062 A1.01 Comments: Proposed reference to be provided to applicant during examination: NONE Technical
Reference:
40OP-9DG01, 40AL-9DG01 K&A: Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the ac distribution system controls including:
Significance of D/G load limits.
Justification:
A- This is not correct. A Generator Differential trip senses a difference in current on the three phases of the neutral side of the generator and the three phases on the bus side of the output breaker.
B- This is not correct. This alarm senses a voltage regulator failure. The stem refers to lowering load, not voltage.
C- Correct. 40OP-9DG01 (and DG02) state that a reverse power trip can occur in the 0.3 to 0.5 MW indicated range.
D- This is not correct. This trip is caused by current differential between phases.
Rev 0 Question 48 Table-Item Links Q - 10CFR Sections 55.41 (5) Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.
Q - Reactor Operator K & A Importance classification 3.4 Q - Cognitive Level Memory Q - Question Source PV Bank Not Modified O - RO Tier Group Designations Tier 2 Group 1 Q - RO Exam 2010 RO NRC Associated objective(s):
Describe the minimum load required for unloading the DG.
Rev 0 49 ID: Q43817 Points: 1.00 Given the following conditions:
The Unit 3 Control Room has been evacuated due to a FIRE.
The CRS is implementing 40AO-9ZZ19 (Control Room Fire)
You as the RO, have been directed to perform Appendix A (Primary Reactor Operator Actions).
Which ONE of the following describes an Appendix A task and the operational impacts of performing this action?
A. Ensure EDG A is running to support 'A' Train equipment.
B. Open the breakers for all charging pumps to prevent inadvertent dilution.
C. Ensure the 'B' NCW Pump is operating to maintain RCP cooling water flow.
D. Open breakers for the MSIV Logic Cabinet SGA-C01 at PKA-M41, to prevent un-intended operation of MSIVs.
Answer: D
Rev 0 Question 49 Info Question Multiple Choice Type:
Status: Active Always select No on test?
Authorized No for practice?
Points: 1.00 Time to 2
Complete:
Difficulty: 3.00 System ID: 43817 User-Defined Q43817 ID:
Cross Reference Number:
Topic: Q43817 RO actions on CR FIRE outside CR.
Num Field 1: 4.2 Num Field 2: 4.1 Text Field: 2.4.34 Comments: Proposed reference to be provided to applicant during examination: NONE Technical
Reference:
40AO-9ZZ19, Control Room Fire K&A:
Knowledge of RO tasks performed outside the main control room during an emergency and the resultant operational effects. DC Electrical Distribution.
Justification:
A. Incorrect - The action is to secure the EDG A if running. Ensure it is running is wrong.
B. Incorrect - CCP B is to remain in operation to ensure seal injection to the RCP seals.
C. Incorrect - Appendix B directs securing the NCW pumps.
D. Correct - This is accomplished by opening the DC supply breakers on PKA-M41 in the DC Equipment Room.
These appendices A and B are used to isolate train B equipment from the control room to prevent fire related operation while allowing for remote operation as well as disabling/de-energizing other equipment to prevent fire initiated operations.
Rev 0 Question 49 Table-Item Links Q - 10CFR Sections 55.41 (10) Administrative, normal, abnormal, and emergency operating procedures for the facility.
Q - Reactor Operator K & A Importance classification 4.2 Q - Cognitive Level Comprehension / Anal Q - Question Source New O - RO Tier Group Designations Tier 2 Group 1 Q - RO Exam 2010 RO NRC Associated objective(s):
describe what is accomplished by performing the Primary and Secondary Reactor Operator Actions
Rev 0 50 ID: Q4831 Points: 1.00 Given the following conditions:
Unit 1 is operating at 15.5% power.
Power is being raised to 20%.
Steam Generator #1 downcomer valve position is 80% and is opening slowly.
Steam Generator #2 downcomer valve position is 70% and is opening slowly.
Concerning the DFWCS, which (if any) of the following automatic actions will occur?
A. Only Steam Generator #1 will go through swapover.
B. No DFWCS automatic action will occur until 16.5% power is reached.
C. Both Steam Generator #1 and Steam Generator #2 will go through swapover.
D. No DFWCS automatic action will occur until both Steam Generator downcomer valves are 80% open.
Answer: C
Rev 0 Question 50 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 3.00 System ID: 4831 User-Defined ID: Q4831 Cross Reference Number:
Topic: Q4831 Determine if swapover will occur Num Field 1: 2.7 Num Field 2: 2.9 Text Field: 3. 4 059 A1.03 Comments: Proposed reference to be provided to applicant during examination:
NONE Technical
Reference:
Operator Information Manual, 40OP-9ZZ04(Plant Startup Mode 2 to 1)
K&A: Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the MFW controls including: Power level restrictions for operation of MFW pumps and valves.
Justification:
A. Incorrect - Both SG#1 and SG#2 will go through swapover due to power greater that 15.5% and one downcomer valve > 80% open.
B. Incorrect - Both SG#1 and SG#2 will go through swapover due to power greater that 15.5% and one downcomer valve > 80% open.
C. Correct - 40OP-9ZZ04(Plant Startup Mode 2 to 1), page 46 states that swapover will occur if power is greater than 16.5% OR between 15 and 16.5% AND one downcomer valve reaches 80%. The simplified drawings show that if one DFWCS meets these criteria, the other will swapover as well. This explanation renders distracters A, B, and D as incorrect.
D. Incorrect - Only one of the Downcomer valves must be > 80% open.
Rev 0 Question 50 Table-Item Links Q - 10CFR Sections 55.41 (5) Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.
Q - Reactor Operator K & A Importance classification 2.7 Q - Cognitive Level Comprehension / Anal Q - Question Source PV Bank Not Modified O - RO Tier Group Designations Tier 2 Group 1 Q - RO Exam RO Exam Question 2007 2010 RO NRC Associated objective(s):
Describe the response of the DFWCS to an increase in reactor power to include the following:
- Swapover
- Single - three element control change
- Downcomer reopening
Rev 0 51 ID: Q22680 Points: 1.00 Given the following conditions:
EDG 'B' is operating at full rated load.
The associated fuel oil transfer pump has faulted.
EDG 'B' Fuel Oil Day Tank is at it's Technical Specification Limit of (2.75ft/550 gallons)
Which ONE of the following describes the approximate amount of time EDG 'B' will be able to operate in this condition?
A. 0.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />.
B. 1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />.
C. 2.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />.
D. 5.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />.
Answer: B
Rev 0 Question 51 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 3.00 System ID: 22680 User-Defined ID: Q22680 Cross Reference Number:
Topic: Q22680 DFO DAY TANK BASIS Num Field 1: 3.2 Num Field 2: 3.3 Text Field: 3.6 064 K6.08 Comments: Proposed reference to be provided to applicant during examination:
NONE Technical
Reference:
LOIT Lesson Plan, Tech Spec 3.8.1 and basis.
K&A:
Knowledge of the effect of a loss or malfunction of the following will have on the ED/G system: Fuel oil storage tanks.
Justification:
A. Incorrect - EDG uses 6.4 gpm of fuel oil at rated load, therefore 550 gallons is approximately 1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> of use.
B. Correct - EDG uses 6.4 gpm of fuel oil at rated load, therefore 550 gallons is approximately 1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> of use. Basis of the Limit is 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> of operation plus 10%.
C. Incorrect - Tank capacity is 1100 gallons and sized for 2.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> of fuel oil.
D. Incorrect - Each EDG has its own tank and each has a capacity of 1100 gallons and sized for 2.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> of fuel oil. Therefore if using both tanks you could have approximately 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> of fuel oil.
Rev 0 Question 51 Table-Item Links Q - 10CFR Sections 55.41 (7) Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
Q - Reactor Operator K & A Importance classification 3.2 Q - Cognitive Level Comprehension / Anal Q - Question Source New O - RO Tier Group Designations Tier 2 Group 1 Q - RO Exam 2010 RO NRC Associated objective(s):
Explain the operation of the Diesel Fuel Oil sub-system under normal operating conditions.
Rev 0 52 ID: Q43756 Points: 1.00 Given the following conditions:
EW A is supplying NC.
A leak develops in the 1A RCP HPSC.
You are directing the Area 3 operator to adjust EWA-HCV-53 (SDCH 'A' Bypass Valve).
Which ONE of the following describes the radiation intensity and exposure limits for the Area 3 operator?
Radiation Intensity in the SDCHX A room is ____(1)____ and the Federal exposure limit is ____(2)____.
A. (1) rising (2) 2.5 Rem/year.
B. (1) rising (2) 5.0 Rem/year.
C. (1) constant (2) 2.5 Rem/year.
D. (1) constant (2) 5.0 Rem/year.
Answer: B
Rev 0 Question 52 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 3.00 System ID: 43756 User-Defined ID: Q43756 Cross Reference Number:
Topic: Q43756 Radiation Intensity and Exp Limits Num Field 1: 2.9 Num Field 2: 3.4 Text Field: 3.7 073 K5.03 Comments: Proposed reference to be provided to applicant during examination:
NONE Technical
Reference:
Radworker Training.
K&A:
Knowledge of the operational implications as they apply to concepts as they apply to the PRM system: Relationship between radiation intensity and exposure limits.
Justification:
A. Incorrect - As the HPSC leak introduces RCS in the NC and EW systems, RU-2 will show a rising trend in activity and radiation levels.
The Federal exposure limit is 5 Rem/year. PVNGS has administrative hold points that the candidate may confuse for the Federal limit.
B. Correct - As the HPSC leak introduces RCS in the NC and EW systems, RU-2 will show a rising trend in activity and radiation levels.
The Federal exposure limit is 5 Rem/year.
C. Incorrect - Candidate may not understand the relationship between the HPSC leak and the NC and EW systems, RU-2 will show a rising trend in activity and radiation levels. The Federal exposure limit is 5 Rem/year. PVNGS has administrative hold points that the candidate may confuse for the Federal limit.
D. Incorrect - Candidate may not understand the relationship between the HPSC leak and the NC and EW systems, RU-2 will show a rising trend in activity and radiation levels.
Rev 0 Question 52 Table-Item Links Q - 10CFR Sections 55.41 (5) Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.
Q - Reactor Operator K & A Importance classification 2.9 Q - Cognitive Level Comprehension / Anal Q - Question Source New O - RO Tier Group Designations Tier 2 Group 1 Q - RO Exam 2010 RO NRC Associated objective(s):
Radiological Questions, Radworker Training.
Rev 0 53 ID: Q9362 Points: 1.00 Given the following conditions:
Initial conditions:
Unit 2 is operating at 100% power.
Plant Cooling Water pumps PWN-P01A is running with PWN-P01B in standby.
Subsequently:
The following indications are locked in on B07:
PCW PCW DISCH SYS PRESS TRBL HI/LO PWN-P01A has brighter than normal green indication.
PWN-P01B has both amber and red lights.
Which ONE of the following describes the current status of the Plant Cooling Water Pumps?
A. A multiplexer failure has caused a trip of both pumps.
B. Both pumps have locked out and can be reset (if the condition has cleared) from B07.
C. Both pumps have locked out and must be reset (if the condition has cleared) at the switchgear.
D. PWN-P01A has tripped and locked out. PWN-P01B is running but has degraded discharge pressure.
Answer: D
Rev 0 Question 53 Info Question Multiple Choice Type:
Status: Active Always select Yes on test?
Authorized for No practice?
Points: 1.00 Time to 3
Complete:
Difficulty: 2.00 System ID: 9362 User-Defined Q9362 ID:
Cross Reference Number:
Topic: Q9362 PW Controls associated with the PW Pumps Num Field 1: 2.9 Num Field 2: 2.9 Text Field: 3.4 062 A4.01 Comments: Proposed reference to be provided to applicant during examination: NONE Technical
Reference:
LOIT PW Lesson Plan K&A:
Ability to manually operate and/or monitor in the control room: SWS pumps.
Justification:
A. Incorrect - A multiplexer failure applies to the circulating water system and would result in no indications on B07.
B. Incorrect - Only P01A has tripped due to the brighter than normal green on B07, the 86 lockout can only be reset from the switchgear on non class pumps.
C. Incorrect - Only P01A has tripped due to the brighter than normal green on B07, the 86 lockout can only be reset from the switchgear on non class pumps.
D. Correct -The brighter than normal green shows that P01A has tripped, the P01B indications of amber means started in standby, red means running and the PCW low discharge pressure alarm means that the pump is not pumping properly.
Rev 0 Question 53 Table-Item Links Q - 10CFR Sections 55.41 (7) Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
Q - Reactor Operator K & A Importance classification 2.9 Q - Cognitive Level Comprehension / Anal Q - Question Source PV Bank Not Modified O - RO Tier Group Designations Tier 2 Group 1 Q - RO Exam 2010 RO NRC Associated objective(s):
Describe the Control Room controls associated with the PW Pumps, including their indications.
Rev 0 54 ID: Q8375 Points: 1.00 Which ONE of the following describes the order in which the Instrument Air (IA) system will respond to a lowering IA header pressure?
A. IA Header low pressure alarm, Standby compressor starts, Nitrogen backup supply opens.
B. Standby compressor starts, IA Header low pressure alarm, Nitrogen backup supply opens.
C. Standby compressor starts, Nitrogen backup supply opens, IA Header low pressure alarm.
D. IA Header low pressure alarm, Nitrogen backup supply opens, Standby compressor starts.
Answer: B Question 54 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 3.00 System ID: 8375 User-Defined ID: Q8375 Cross Reference Number:
Topic: Q8375 SYS Automatic functions IA Num Field 1: 3.1 Num Field 2: 3.2 Text Field: 3.8 078 A3.01 Comments: Proposed reference to be provided to applicant during examination:
NONE Technical
Reference:
40OP-9IA0, IA normal ops, 40AL-9RK7B, alarm response.
K&A: Ability to monitor automatic operation of the IAS, including: Air pressure Justification:
There is only one correct sequence for this scenario. Candidate may not fully understand the IA system/setpoints and choose any one of the choices for the same reasons.
A. Incorrect - Only one correct sequence for this scenario.
B. Correct - Normally one compressor is running maintaining 115 psig.
The first standby compressor will start at 109 psig and the second standby compressor starts at 105 psig (40OP-9IA01). The instrument air low pressure alarm is at 95 psig (40AL-9RK7B) and the N2 backup opens at 85.
C. Incorrect - Only one correct sequence for this scenario.
D. Incorrect - Only one correct sequence for this scenario.
Rev 0 Question 54 Table-Item Links Q - 10CFR Sections 55.41 (7) Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
Q - Reactor Operator K & A Importance classification 3.1 Q - Cognitive Level Memory Q - Question Source PV Bank Not Modified O - RO Tier Group Designations Tier 2 Group 1 Q - RO Exam RO Exam Question 2007 2010 RO NRC Associated objective(s):
Describe the automatic functions associated with the Instrument Air System.
Rev 0 55 ID: Q43767 Points: 1.00 Given the following conditions:
Unit 1 has tripped from 100% power.
A LOCA has occurred inside containment.
Pressurizer pressure is 1475 psia and lowering.
Containment pressure is 9.5 psia and rising.
Which ONE of the following describes the signal(s) that closed the Normal Chill Water (WC) Containment Isolation Valves?
A. SIAS Only.
B. CIAS Only.
C. CSAS Only.
D. SIAS and CIAS Only.
Answer: B Question 55 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 3.00 System ID: 43767 User-Defined ID: Q43767 Cross Reference Number:
Topic: Q43767 WC Containment Isolation Valves Num Field 1: 3.9 Num Field 2: 4.1 Text Field: 3.5 103 K1.02 Comments: Proposed reference to be provided to applicant during examination:
NONE Technical
Reference:
WC Containment Isolation Valve Electrical Print.
K&A:
Knowledge of the physical connections and/or cause-effect relationships between the containment system and the following systems: Containment isolation/containment integrity.
Justification:
A. Incorrect - SIAS closes the SG Blowdown containment isolation valves.
B. Correct - CIAS is the only ESFAS signal that will close the WC containment isolation valves.
C. Incorrect - CSAS closes the IA header containment isolation valves.
D. Incorrect - SIAS does not close the WC containment isolation valves.
Rev 0 Question 55 Table-Item Links Q - 10CFR Sections 55.41 (9) Shielding, isolation, and containment design features, including access limitations Q - Reactor Operator K & A Importance classification 3.9 Q - Cognitive Level Comprehension / Anal Q - Question Source PV Bank Not Modified O - RO Tier Group Designations Tier 2 Group 1 Q - RO Exam 2010 RO NRC Associated objective(s):
Explain the operation of the WC Containment Isolation Valves under normal operating conditions.
Rev 0 56 ID: Q22681 Points: 1.00 Given the following conditions:
Unit 1 has tripped from 100% power.
This trip was due to a sheared shaft on the 1A RCP.
SG #1 level is noticed to be higher than SG #2 with the same feed flow.
Which ONE of the following describes the SG level difference?
SG #1 level is higher than SG #2 due to...
A. SG #2 is generating more steam due to the greater RCS flow.
B. the lower Tcold condition in the loop with only 1 RCP operating.
C. SG #1 having significantly less pump heat to transfer out of the RCS.
D. RCS flow reversing in the loop with 1 RCP operating and two pumps operating in the opposite loop.
Answer: A
Rev 0 Question 56 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 4.00 System ID: 22681 User-Defined ID: Q22681 Cross Reference Number:
Topic: Q22681 UNBALANCED RCS FLOW EFFECTS Num Field 1: 3.6 Num Field 2: 3.8 Text Field: 3.2 002 K6.02 Comments: Proposed reference to be provided to applicant during examination:
NONE Technical
Reference:
LOIT Lesson Plan K&A:
Knowledge of the effect of a loss or malfunction on the following RCS components: RCP.
Justification:
A. Correct - With unbalanced RCS flow (one pump secured), the SG with two RCPs will transfer more heat than the other one. Based on this, its level will drop more quickly with equal feed flow to both SGs.
B. Incorrect - Tcold is approximately the same due to SBCS maintaining SG pressure.
C. Incorrect - SG #1 will have less pump heat to transfer, but it will not have the effect that the decay heat being pushed through SG #2 will have on level.
D. Incorrect - Unbalanced RCS loop flows will not cause flow reversals.
Rev 0 Question 56 Table-Item Links Q - 10CFR Sections 55.41 (7) Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
Q - Reactor Operator K & A Importance classification 3.6 Q - Cognitive Level Comprehension / Anal Q - Question Source New O - RO Tier Group Designations Tier 2 Group 2 Q - RO Exam 2010 RO NRC Associated objective(s):
Given conditions of a reactor trip describe SG response when unbalanced RCS flow is established in accordance with 40EP-9EO01.
Rev 0 57 ID: Q43769 Points: 1.00 Given the following conditions:
Gaseous Radwaste (GR) is lined up to receive gas from surge header.
Oxygen content in the Volume Control Tank (VCT) is 1.8% and rising.
Oxygen content in the surge header is 1.8% and rising.
Which ONE of the following describes the action(s) that will occur to reduce the risk or prevent flammability as oxygen concentration continues to rise?
A. VCT to GR header ONLY isolates automatically.
B. All inputs to GR header will isolate automatically.
C. GRN-AV-169 (Auto N2 Dilution) automatically opens.
D. Waste gas compressors will start and cycle on surge tank pressure.
Answer: C Question 57 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 3.00 System ID: 43769 User-Defined ID: Q43769 Cross Reference Number:
Topic: Q43769 GR--Nitrogen Dilution Num Field 1: 2.5 Num Field 2: 3.1 Text Field: 3.9 071 K5.04 Comments: Proposed reference to be provided to applicant during examination: NONE Technical
Reference:
40AL-9ZR1A, Alarm Response Procedure K&A:
Knowledge of the operational implication of the following concepts as they apply to the Waste Gas Disposal System: Relationship of hydrogen/oxygen concentrations to flammability.
Justification:
A. Incorrect - VCT to GR header isolation is a manual valve that has no automatic functions. GR-UV-34A/B auto close on high discharge flow, high rad levels or high pressure.
B. Incorrect - RDT, EDT, VCT and Gas Stripper all have manual valves that have on automatic functions.
C. Correct - At 3.75% O2 the Auto N2 Dilution Valve will open and the Waste Gas Compressor(s) will stop.
D. Incorrect - WG compressors stop on high-high 02 concentrations and will not start until reset manually.
Rev 0 Question 57 Table-Item Links Q - 10CFR Sections 55.41 (5) Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.
Q - Reactor Operator K & A Importance classification 2.5 Q - Cognitive Level Comprehension / Anal Q - Question Source PV Bank Not Modified O - RO Tier Group Designations Tier 2 Group 2 Q - RO Exam 2010 RO NRC Associated objective(s):
Describe the automatic functions / interlocks associated with the following:
Gaseous Discharge Isolation Valves, Gas Compressors (GRN-C01A, C01B),
Gaseous Discharge Header Isolation Valves (HS-34A and 34B), Radiation Monitor (RU-12).
Rev 0 58 ID: Q22683 Points: 1.00 Given the following conditions:
A reactor startup is in progress The HI LOG PWR LVL BYP PERM (5A15B) annunciator alarms.
Which ONE of the following describes the current conditions and appropriate response?
A. Rx power is greater than 10-4%, manually bypass the Hi Log Power trips.
B. Count rate is greater than 2000 cps, manually bypass the Hi Log Power trips.
C. Rx power is greater than 10-4%, verify that the Hi Log Power trips have been automatically bypassed.
D. Count rate is greater than 2000 cps, verify that the Hi Log Power trips have been automatically bypassed.
Answer: A Question 58 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 2.00 System ID: 22683 User-Defined ID: Q22683 Cross Reference Number:
Topic: Q22683 NI TRIP BYPASSES Num Field 1: 3.8 Num Field 2: 3.9 Text Field: 3.7 015 A4.03 Comments: Proposed reference to be provided to applicant during examination:
NONE Technical
Reference:
40OP-9ZZ03, Reactor Startup K&A:
Ability to manually operate and/or monitor in the control room: Trip bypasses.
Justification:
A. Correct - Alarm is generated at 10 -4% on start ups to remind the operator to bypass the Hi Log Power trip that is set at 10-2%.
B. Incorrect - Start Up channels will be de-energized at 2000 cps when the S/U Channel TRBL alarm is energized.
C. Incorrect - Hi Log Power trip will automatically become active below 10-4%.
D. Incorrect - Start Up channels will be de-energized at 2000 cps therefore when the S/U Channel TRBL alarm is energized. Log Power trip will automatically become active below 10-4%.
Rev 0 Question 58 Table-Item Links Q - 10CFR Sections 55.41 (7) Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
Q - Reactor Operator K & A Importance classification 3.8 Q - Cognitive Level Comprehension / Anal Q - Question Source New O - RO Tier Group Designations Tier 2 Group 2 Q - RO Exam 2010 RO NRC Associated objective(s):
Describe the Control Room controls associated with the Safety Channels, including their indications.
Rev 0 59 ID: Q43753 Points: 1.00 Core Exit Thermocouples (CETs) provide a DIRECT input to which ONE of the following?
A. COLSS.
B. QSPDS.
C. ERFDADS.
D. B02 Post Accident Meters.
Answer: B Question 59 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 3.00 System ID: 43753 User-Defined ID: Q43753 Cross Reference Number:
Topic: Q43753 ITM CET QSPDS Num Field 1: 3.2 Num Field 2: 3.2 Text Field: 3.7 017 K1.01 Comments: Proposed reference to be provided to applicant during examination:
NONE Technical
Reference:
LOIT Lesson Plan K&A:
Knowledge of the physical connections and/or cause effect relationships between the ITM system and the following systems: Plant computer.
Justification:
A. Correct - CET detectors are connected to the QSPDS cabinet by a chromel aluminum lead which removes the need for a temperature controlled environment junction box.
B. Incorrect - COLSS receives inputs from the Incore detectors which are on the same instrument string as the CETs.
C. Incorrect - ERFDADS receives CET data from QSPDS.
D. Incorrect - B02 Post Accident Monitors receive data from QSPDS to display Core Exit Temps and Saturation Margins.
Rev 0 Question 59 Table-Item Links Q - 10CFR Sections 55.41 (7) Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
Q - Reactor Operator K & A Importance classification 3.2 Q - Cognitive Level Memory Q - Question Source New O - RO Tier Group Designations Tier 2 Group 2 Q - RO Exam 2010 RO NRC Associated objective(s):
Explain the operation of the Core Exit Thermocouples (CETs) associated with the Incore Instrumentation System.
Rev 0 60 ID: Q22685 Points: 1.00 Which ONE of the following describes the power supply to the B Hydrogen Recombiner?
A. NBN-S02.
B. PBB-S04.
C. PHB-M34.
D. PGB-L36.
Answer: C Question 60 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 3.00 System ID: 22685 User-Defined ID: Q22685 Cross Reference Number:
Topic: Q22685 H2 RECOMB POWER SUPPLY Num Field 1: 2.5 Num Field 2: 2.8 Text Field: 3.5 028 K2.01 Comments: Proposed reference to be provided to applicant during examination:
NONE Technical
Reference:
40AO-9ZZ12, Degraded Electrical PHB M34 Loads K&A:
Knowledge of bus power supplies to the following: Hydrogen Recombiners.
Justification:
A. Incorrect - 4160V Non Class power supply.
B. Incorrect - 4160V Class power supply.
C. Correct - 480V Class MCC is the power supply to the H2 Recombiners.
D. Incorrect - 480V Class Load Center.
Rev 0 Question 60 Table-Item Links Q - 10CFR Sections 55.41 (7) Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
Q - Reactor Operator K & A Importance classification 2.5 Q - Cognitive Level Memory Q - Question Source New O - RO Tier Group Designations Tier 2 Group 2 Q - RO Exam 2010 RO NRC Associated objective(s):
Explain the operation of the Hydrogen Recombiner under normal operating conditions.
Rev 0 61 ID: Q43732 Points: 1.00 Given the following conditions:
Unit 1 has just experienced an uncomplicated reactor trip.
Tc is 565°F.
Which ONE of the following describes how the DFWCS is controlling steam generator levels?
A. RTO controller output modulating the Economizer control valve.
B. RTO controller output modulating the Downcomer control valve.
C. Master controller output in 3 element control, modulating the Downcomer control valve.
D. Master controller output in single element control, modulating the Downcomer control valve.
Answer: B
Rev 0 Question 61 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 3.00 System ID: 43732 User-Defined ID: Q43732 Cross Reference Number:
Topic: Q43732 DFWCS to a Reactor trip condition.
Num Field 1: 4.0 Num Field 2: 3.9 Text Field: 3.4 035 A3.01 Comments: Proposed reference to be provided to applicant during examination:
NONE Technical
Reference:
LOIT DFWCS Lesson Plan K&A:
Ability to monitor automatic operation of the S/G including: S/G water level control.
Justification:
A. Incorrect - RTO controller will only control the downcomer control valve.
B. Correct - RTO will control the downcomer control valve post trip until SGWL reaches 51.9% NR.
C. Incorrect - Master Controller is disabled until SGWL reaches 51.9%
NR. 3E control is the normal DFWCS control mode.
D. Incorrect - Master Controller is disabled until SGWL reaches 51.9%
NR. 1E control is the normal low power DFWCS control mode.
Rev 0 Question 61 Table-Item Links Q - 10CFR Sections 55.41 (7) Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
Q - Reactor Operator K & A Importance classification 4.0 Q - Cognitive Level Comprehension / Anal Q - Question Source New O - RO Tier Group Designations Tier 2 Group 2 Q - RO Exam 2010 RO NRC Associated objective(s):
Describe the response of the DFWCS to a Reactor Trip condition.
Rev 0 62 ID: Q10487 Points: 1.00 Given the following conditions:
Initial Conditions:
Unit 1 is operating at 99%.
40AO-9ZZ25 (ECC Directed Turbine Unloading) is in progress.
Turbine load has been lowered approximately 11%.
SBCS Master Controller SGN-PIC-1010 is in Local Auto with its setpoint lowered to 1075 psia.
CEDMCS is in manual sequential.
Reg Group 5 is 120 inches withdrawn.
SBCV #1001 is 70% open in automatic with a manual permissive.
SBCV #1004 is 30% open in manual with a manual permissive.
Subsequently:
SBCV #1004 fails 100% open.
Tavg-Tref Hi-Lo window is in alarm (4A08B).
COLSS CMC (5B01C) & PC (5B01D) windows are in alarm.
CEA insertion demand (green lights) are present on B05.
SBCV #1001 will ____(1)____. The operating crew should ____(2)____ to correct this condition.
A. (1) fast close. (2) withdraw CEAs to clear the Tavg - Tref alarm window.
B. (1) modulate closed. (2) reduce power to less than the licensed power limit.
C. (1) not close until the manual permissive is removed. (2) reduce power to less than the licensed power limit.
D. (1) only modulate closed if SGN-PIC-1010 is returned to Auto/Auto control. (2) withdraw CEAs to clear the Tavg - Tref alarm window.
Answer: B
Rev 0 Question 62 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 5 Difficulty: 4.00 System ID: 10487 User-Defined ID: Q10487 Cross Reference Number:
Topic: Q10487 SBCV stuck open affect on SBCS Num Field 1: 3.6 Num Field 2: 3.9 Text Field: 34041A2.02 Comments: Proposed reference to be provided to applicant during examination: None Technical
Reference:
40AL-9RK5B, ARPs / 40AO-9ZZ25, ECC Directed Turbine Unloading K&A: Ability to (a) predict the impacts of the following malfunctions or operations on the SDS; and (b) based on those predictions or mitigate the consequences of those malfunctions or operations: Steam valve stuck open.
Justification:
A. Incorrect - Tavg/Tref is a low alarm due to 1004 coming open. CEA wd demand would be expected for turbine power > reactor power.
B. Correct - 1001 will modulate closed. Manual permissive and Local/Auto will not effect valve operation. COLSS alarms need to be cleared.
C. Incorrect - SBCV 1001 will modulate closed as 1004 comes open.
The manual permissive allows for the valves to remain open. If the manual permissive is removed the valves will quick close, but is not the only way to close the valves. The permissive is only one of the criteria needed for the valve to open.
D. Incorrect - SBCV 1001 will modulate closed as 1004 comes open.
Auto/Auto will cause the setpoint to change back to 1170 psia, therefore the valves would modulate closed in auto/auto as well as the condition in the answer.
Rev 0 Question 62 Table-Item Links Q - 10CFR Sections 55.41 (5) Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.
Q - Reactor Operator K & A Importance classification 3.6 Q - Cognitive Level Comprehension / Anal Q - Question Source PV Bank Not Modified O - RO Tier Group Designations Tier 2 Group 2 Q - RO Exam 2009 RO Exam 2010 RO NRC Associated objective(s):
Describe the Control Room controls associated with the individual SBCS valves including:
- Emergency Off Reset
- Valve Mode Select
Rev 0 63 ID: Q43770 Points: 1.00 Given the following conditions:
PCA-P01 (Spent Fuel Pool Cooling Pump A) is operating on the Spent Fuel Pool.
A large break occurs in the discharge of the pump.
Which ONE of the following will prevent Spent Fuel Pool level from lowering to the point of uncovering fuel assemblies?
A. Excessive flow check valves will automatically close.
B. PCA-P01 (Spent Fuel Pool Cooling Pump A) will trip on high discharge flow.
C. PCA-P01 (Spent Fuel Pool Cooling Pump A) will trip on low suction pressure.
D. Anti-siphon holes in PC cooling pump suction piping become uncovered by the lowering spent fuel pool level.
Answer: D
Rev 0 Question 63 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 2.00 System ID: 43770 User-Defined ID: Q43770 Cross Reference Number:
Topic: Q43770 PC ANTI SIPHON Num Field 1: 2.7 Num Field 2: 3.3 Text Field: 3.8 033 A1.01 Comments: Proposed reference to be provided to applicant during examination:
NONE Technical
Reference:
LOIT PC Lesson Plan K&A:
Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with Spent Fuel Pool Cooling System operating the controls including: Spent fuel pool water level.
Justification:
A. Incorrect - The check valves installed on the discharge of the pump are not excessive flow check valves as used in other locations in the plant.
B. Incorrect - PC Cooling Pumps are not provided with a high discharge flow trip. Some pumps at PVNGS do have a high discharge pressure trip.
C. Incorrect - PC Cooling Pumps are not provided with a low suction pressure trip. Some pumps installed at PVNGS do have a low suction trip.
D. Correct - Anti Siphon holes are drilled into the PC Cooling Pump suction piping so that a break external to the SFP will not drain the pool.
Rev 0 Question 63 Table-Item Links Q - 10CFR Sections 55.41 (5) Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.
Q - Reactor Operator K & A Importance classification 2.7 Q - Cognitive Level Memory Q - Question Source New O - RO Tier Group Designations Tier 2 Group 2 Q - RO Exam 2010 RO NRC Associated objective(s):
Explain the operation of the Spent Fuel Pool under normal operating conditions.
Rev 0 64 ID: Q43771 Points: 1.00 Given the following conditions:
Unit 1 is operating at 100% power.
Main Condenser shell A vacuum is 2.6 in HgA and degrading.
Which ONE of the following conditions would be the cause of this condition?
A. Ambient relative humidity lowering.
B. Circulating Water temperature lowering.
C. 86 lockout on ARN-P01 (Air Removal Pump A).
D. Vacuum breaker make-up isolation valve leaking by.
Answer: C Question 64 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 1 Difficulty: 2.00 System ID: 43771 User-Defined ID: Q43771 Cross Reference Number:
Topic: Q43771 AR Pump Malf effect on MC Num Field 1: 2.5 Num Field 2: 2.7 Text Field: 3.4 055 K3.01 Comments: Proposed reference to be provided to applicant during examination:
NONE Technical
Reference:
LOIT Lesson Plan, Fundamental Knowledge.
K&A:
Knowledge of the effect that a loss or malfunction of the CARS will have on the following: Main condenser.
Justification:
A. Incorrect - lowering humidity will raise the efficiency of the main condenser therefore resulting in improved vacuum.
B. Incorrect - CW temperature lowering will raise the efficiency of the main condenser therefore resulting in improved vacuum.
C. Correct - AR pumps remove air and non condensable gases from the main condenser. 3 of 4 pumps are normally operating. A 4th standby pump will start when vacuum degrades further to excess of 4.5 in HgA.
D. Incorrect - Vacuum breaker makeup valve will have no impact due to supplying more water to the vacuum breaker. If the drain valve were to leak by the resultant air in-leakage would result in degrading vacuum.
Rev 0 Question 64 Table-Item Links Q - 10CFR Sections 55.41 (5) Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.
Q - Reactor Operator K & A Importance classification 2.7 Q - Cognitive Level Memory Q - Question Source New O - RO Tier Group Designations Tier 2 Group 2 Q - RO Exam 2010 RO NRC Associated objective(s):
Discuss the purpose and conditions under which the Air Removal System is designed to function.
Rev 0 65 ID: Q5534 Points: 1.00 Given the following conditions:
An attempt to start a Circulating Water pump has been made and the pump failed to start.
Which ONE of the following describes the interlock that is preventing the pump from starting?
The Circulating Water...
A. intake canal level is too low.
B. pump's discharge valve is closed.
C. loop's condenser outlet valve is closed.
D. loop's cross-tie valve (CW-HV-11) is closed.
Answer: C Question 65 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 1 Difficulty: 2.00 System ID: 5534 User-Defined ID: Q5534 Cross Reference Number:
Topic: Q5534 CW pump interlock with condenser valve Num Field 1: 2.5 Num Field 2: 2.8 Text Field: 3.8 075 K4.01 Comments: Proposed reference to be provided to applicant during examination:
NONE Technical
Reference:
40OP-9CW01, Operating the CW system K&A:
Knowledge of circulating water system design feature(s) and interlock(s) which provide for the following: Heat Sink.
Justification:
A. Incorrect - There is no interlock associated with intake canal level.
Low level could cause low suction pressure and low flow but will not prevent a pump start.
B. Incorrect - The CW pump discharge valve must be closed to start the pump.
C. Correct - This ensures that water hammer does not damage the condenser water box when a CW pump is started.
D. Incorrect - This valve will close when there is unbalanced flow between CW loops. It will not prevent a pump from starting.
Rev 0 Question 65 Table-Item Links Q - 10CFR Sections 55.41 (7) Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
Q - Reactor Operator K & A Importance classification 2.5 Q - Cognitive Level Memory Q - Question Source PV Bank Not Modified O - RO Tier Group Designations Tier 2 Group 2 Q - RO Exam 2010 RO NRC Associated objective(s):
Describe the automatic interlocks associated with the Circulating Water Condenser Outlet Valves.
Rev 0 66 ID: Q43733 Points: 1.00 Given the following conditions:
You are scheduled to begin your shift as the RO.
This is your first day of nights following two days off.
Which ONE of the following describes when the Night Orders shall be reviewed?
A. After turnover (On-Coming).
B. Prior to turnover (Off-Going).
C. Prior to turnover (On-Coming).
D. Not required to be reviewed since last shift was less than 3 days ago.
Answer: A Question 66 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 3.00 System ID: 43733 User-Defined ID: Q43733 Cross Reference Number:
Topic: Q43733 REVIEW OF NIGHT ORDERS Num Field 1: 2.7 Num Field 2: 3.4 Text Field: 2.1.15 Comments: Proposed reference to be provided to applicant during examination:
NONE Technical
Reference:
40DP-9OP33, Shift Turnover K&A:
Knowledge of administrative requirements for temporary management directives, such as standing orders, night orders, Operations memos, etc.
Justification:
A. Correct - PER 40DP-9OP33 SECT 5.4.
B. Incorrect - RO must ensure PVARs have been issued prior to turnover off-going.
C. Incorrect - RO must review numerous logs and data prior to taking the shift.
D. Incorrect - RO must review Unit Logs back to the last shift worked or last 3 days, whichever is shorter.
Rev 0 Question 66 Table-Item Links Q - 10CFR Sections 55.41 (10) Administrative, normal, abnormal, and emergency operating procedures for the facility.
Q - Reactor Operator K & A Importance classification 2.7 Q - Cognitive Level Memory Q - Question Source New O - RO Tier Group Designations Tier 3 Q - RO Exam 2010 RO NRC Associated objective(s):
Given the conditions associated with a Control Room relief describe the required review of operating logs prior to this relief in accordance with 40DP-9OP33.
Rev 0 67 ID: Q43778 Points: 1.00 Given the following conditions:
A Loss of Class DC control power has occurred.
A Class pump needs to be started.
The Class pump was in a normal standby line up prior to the Loss of DC control power.
Which ONE of the following describes how this would be accomplished?
A. Start the pump using the control switch on B02.
B. Locally start the pump using the local control switch on the breaker.
C. Locally start the pump using the manual close pushbutton located on the breaker.
D. You must manually charge the closing springs and then start the pump using the manual close pushbutton located on the breaker.
Answer: C
Rev 0 Question 67 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 1 Difficulty: 2.00 System ID: 43778 User-Defined ID: Q43778 Cross Reference Number:
Topic: Q43778 Loss of Control Power Close Breaker Num Field 1: 4.4 Num Field 2: 4.0 Text Field: 2.1.30 Comments: Proposed reference to be provided to applicant during examination:
NONE Technical
Reference:
LOIT Lesson Plan PB K&A:
Ability to locate and operate components, including local controls.
Justification:
A. Incorrect - Loss of DC control power to the breaker will prevent controlling the pump from the control room.
B. Incorrect - Loss of DC control power to the breaker will prevent controlling the pump from the local hand switch on the breaker cubicle.
C. Correct - You must open the breaker cubicle and press the pushbutton.
D. Incorrect - Closing springs get closed when the breaker is opened. If this was not the first time to close the breaker following the loss of DC control power, you would have to charge the closing springs manually and then depress the pushbutton.
Rev 0 Question 67 Table-Item Links Q - 10CFR Sections 55.41 (7) Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
Q - Reactor Operator K & A Importance classification 4.4 Q - Cognitive Level Comprehension / Anal Q - Question Source New O - RO Tier Group Designations Tier 3 Q - RO Exam 2010 RO NRC Associated objective(s):
Explain the operation of the 4.16 kV Circuit Breakers under normal operating conditions.
Rev 0 68 ID: Q21189 Points: 1.00 A reactor startup is in progress after a two week mid-cycle outage. The ECC has the following data:
ECRP -500 pcm......... Group 3 at 90 inches ECRP......................... Group 4 at 74 inches ECRP +500 pcm........ All Rods Out Criticality is achieved and verified with Group 3 at 50 inches withdrawn.
Which ONE of the following describes the required operator response?
A. Trip the reactor and emergency borate.
B. Insert regulating group rods to their lower group stop ONLY.
C. Insert CEAs as necessary to hold current power level and take critical data.
D. Insert regulating group rods to their lower group stop and emergency borate.
Answer: A
Rev 0 Question 68 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 3.00 System ID: 21189 User-Defined ID: Q21189 Cross Reference Number:
Q21189 Required actions if criticality is achieved below the ECRP - 500/-
Topic:
750 pcm position but Num Field 1: 4.5 Num Field 2: 4.4 Text Field: 2.2.1 Comments: Proposed reference to be provided to applicant during examination:
NONE Technical
Reference:
40OP-9ZZ03 Reactor Startup K&A:
Ability to perform pre-startup procedures for the facility, including operating those controls associated with plant equipment that could affect reactivity.
Justification:
A. Correct - This is the correct action per App C of 40OP-9ZZ03 (Reactor Startup) for achieving criticality below the PDIL (Grp 3 at 60 inches)
B. Incorrect - This is the correct action when critical below the ECRP -
500 but above the PDIL.
C. Incorrect - This is performed if the Reactor goes critical at or near the ECRP.
D. Incorrect - This is a combination of the correct responses to early criticality.
Rev 0 Question 68 Table-Item Links Q - 10CFR Sections 55.41 (10) Administrative, normal, abnormal, and emergency operating procedures for the facility.
Q - Reactor Operator K & A Importance classification 4.5 Q - Cognitive Level Comprehensive / Anal Q - Question Source Bank O - RO Tier Group Designations Tier 3 Q - RO Exam 2010 RO NRC Associated objective(s):
describe the required actions if criticality is achieved below the ECRP - 500/-750 pcm position but above the PDILs describe the required actions if the Reactor goes critical below the PDILs determine whether or not the Rx startup in progress must be aborted
Rev 0 69 ID: Q43734 Points: 1.00 Which ONE of the following describes the Reactor Core Safety Limits for DNBR and Peak Fuel Centerline temperature?
DNBR shall be maintained at ____(1)____ and Peak Fuel Centerline temperature shall be maintained < _____(2)____.
A. (1) 1.34 (2) 5080 F B. (1) 1.40 (2) 5000 F C. (1) 1.28 (2) 5150 F D. (1) 1.30 (2) 4950 F Answer: A Question 69 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 2.00 System ID: 43734 User-Defined ID: Q43734 Cross Reference Number:
Topic: Q43734 RX CORE SLs Num Field 1: 4.0 Num Field 2: 4.7 Text Field: 2.2.22 Comments: Proposed reference to be provided to applicant during examination:
NONE Technical
Reference:
Technical Specifications K&A:
Knowledge of limiting conditions for operations and safety limits.
Justification:
A. Correct - These are the Safety Limits.
B. Incorrect -The safety limit used to be 21 KW/FT but it is no longer used as a safety limit. It is still calculated and will result in a CPC trip.
C. Incorrect - The safety limit used to be 21 KW/FT but it is no longer used as a safety limit. It is still calculated and will result in a CPC trip.
D. Incorrect - The safety limit used to be 21 KW/FT but it is no longer used as a safety limit. It is still calculated and will result in a CPC trip.
Rev 0 Question 69 Table-Item Links Q - 10CFR Sections 55.41 (5) Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.
Q - Reactor Operator K & A Importance classification 4.0 Q - Cognitive Level Memory Q - Question Source New O - RO Tier Group Designations Tier 3 Q - RO Exam 2010 RO NRC Associated objective(s):
Concerning Technical Specifications state the Safety Limits and describe the bases behind these limits in accordance with the Tech Specs.
Rev 0 70 ID: Q43779 Points: 1.00 Which ONE of the following would require entry into a ONE hour or less LCO condition while in Mode 1, steady state conditions?
A. One LPSI Train Inoperable.
B. One Automatic RPS trip channel Inoperable.
C. Containment average air temperature not within limits.
D. One Safety Injection Tank (SIT) Pressure stable at 560 psig.
Answer: C Question 70 Info Question Multiple Choice Type:
Status: Active Always select on No test?
Authorized No for practice?
Points: 1.00 Time to 3
Complete:
Difficulty: 3.00 System ID: 43779 User-Defined Q43779 ID:
Cross Reference Number:
Topic: Q43779 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> actions for systems Num Field 1: 3.9 Num Field 2: 4.5 Text Field: 2.2.39 Comments: Proposed reference to be provided to applicant during examination:
NONE Technical
Reference:
Tech Specs K&A: Knowledge of less than or equal to one hour Technical Specification action statements for systems.
Justification:
A. Incorrect - This is a 7 day action per TS 3.1.1.
B. Correct - This is a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> action per TS 3.3.1.
C. Incorrect - This is a 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> action per TS 3.6.5.
D. Incorrect - This is a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> action per TS 3.5.1.
Rev 0 Question 70 Table-Item Links Q - 10CFR Sections 55.41 (10) Administrative, normal, abnormal, and emergency operating procedures for the facility.
Q - Reactor Operator K & A Importance classification 3.9 Q - Cognitive Level Memory Q - Question Source New O - RO Tier Group Designations Tier 3 Q - RO Exam 2010 RO NRC Associated objective(s):
apply the one hour or less actions statements of T.S. 3.3
Rev 0 71 ID: Q43737 Points: 1.00 Which ONE of the following describes the use of a hand held frisker?
Frisk the surface of the object at a distance of ____(1)____, at a rate of ____(2)____
inches per second.
A. (1) 1/2 inch (2) 2 B. (1) 1 inch (2) 4 C. (1) 1 inch (2) 2 D. (1) 1/2 inch (2) 4 Answer: A Question 71 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 2.00 System ID: 43737 User-Defined ID: Q43737 Cross Reference Number:
Topic: Q43737 FRISKER USE Num Field 1: 2.9 Num Field 2: 2.9 Text Field: 2.3.5 Comments: Proposed reference to be provided to applicant during examination:
NONE Technical
Reference:
Radworker Training Handout K&A:
Ability to use radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc.
Justification:
A. Correct - This is the correct distance and rate of frisk.
B. Incorrect - Dose rates are measured at 1 ft. Candidate may think that 1 inch for dose rates is the distance for frisking.
C. Incorrect - 4 in2 is the size of a swipe, candidate may think that the swipe size is the same as the frisk speed.
D. Incorrect - Dose rates are measured at 1 ft. Candidate may think that 1 inch for dose rates is the distance for frisking. 4 in2 is the size of a swipe, candidate may think that the swipe size is the same as the frisk speed.
Rev 0 Question 71 Table-Item Links Q - 10CFR Sections 55.41 (11) Purpose and operation of radiation monitoring systems, including alarms and survey equipment.
Q - Reactor Operator K & A Importance classification 2.9 Q - Cognitive Level Memory Q - Question Source New O - RO Tier Group Designations Tier 3 Q - RO Exam 2010 RO NRC Associated objective(s):
Explain the operation of field units under normal operating conditions.
Rev 0 72 ID: Q43738 Points: 1.00 The following information is posted on a survey map outside a room you are about to enter.
General Area Radiation levels of 500 mrem/hr.
Swipes of 150,000 dpm/100cm2.
The posting of this room should be a ...
A. Radiation Area (RA) and Contamination Area (CA).
B. High Radiation Area (HRA) and Contamination Area (CA).
C. High Radiation Area (HRA) and High Contamination Area (HCA).
D. Locked High Radiation Area (LHRA) and High Contamination Area (HCA).
Answer: C Question 72 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 3.00 System ID: 43738 User-Defined ID: Q43738 Cross Reference Number:
Topic: Q43738 RADIATION AREAS Num Field 1: 3.4 Num Field 2: 3.8 Text Field: 2.3.14 Comments: Proposed reference to be provided to applicant during examination:
NONE Technical
Reference:
Radwoker Training Handout K&A:
Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or emergency conditions or activities.
Justification:
A. Incorrect - RA > 5 mrem/hr, CA > 1000 DPM/100cm2 B. Incorrect - HRA > 100 mrem/hr, CA > 1000 DPM/100cm2 C. Correct - HRA > 100 mrem/hr, HCA > 100,000 DPM/100cm2 D. Incorrect - LHRA > 1000 mrem/hr, HCA > 100,000 DPM/100cm2
Rev 0 Question 72 Table-Item Links Q - 10CFR Sections 55.41 (12) Radiological safety principles and procedures.
Q - Reactor Operator K & A Importance classification 3.4 Q - Cognitive Level Memory Q - Question Source Modified PV Bank O - RO Tier Group Designations Tier 3 Q - RO Exam 2010 RO NRC Associated objective(s):
Radiological Questions. Radworker Training.
1 ID: Q22574 Points: 1.00 The following information is posted on a survey map outside a room you are about to enter.
General Area Radiation levels of 75 mrem/hr.
Swipes of 75,000 dpm/100cm2.
The posting of this room should be a ...
A. Radiation Area only.
B. Radiation Area and Contamination Area.
C. Radiation Area and High Contamination Area.
D. High Radiation Area and High Contamination Area.
Answer: B Question 1 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 3.00 System ID: 22574 User-Defined ID: Q22574 Cross Reference Number:
Topic: Q22574 define the radiological area Num Field 1: 3.4 Num Field 2: 3.8 Text Field: 2.3.14 Comments: Proposed reference to be provided to applicant during examination: NONE Technical
Reference:
Rad Worker Training K&A: Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or emergency conditions or activities.
Justification:
Radiation areas are 5 to 100 mrem/hr 2
Contamination areas are 1000 to 100,000 dpm/100cm A is Wrong -
B is Correct -
C is Wrong -
D is Wrong -
OPTRNG_EXAM Page: 1 of 2 09 June 2010
Question 1 Table-Item Links Q - 10CFR Sections 55.41 (12) Radiological safety principles and procedures.
Q - Cognitive Level Memory Q - Question Source New O - RO Tier Group Designations Tier 3 Q - RO Exam 2009 RO Exam Associated objective(s):
Radiological questions OPTRNG_EXAM Page: 2 of 2 09 June 2010
Rev 0 73 ID: Q43739 Points: 1.00 Given the following conditions:
Unit 1 tripped from 100% power.
Main turbine tripped.
Steam Bypass "Quick Open" failed to actuate.
Pressurizer safeties lifted and RCE-PSV-200 did not reseat.
Containment pressure is 16 psig and rising.
Containment temperature is 195°F and rising.
Which ONE of the following describes the Control Room instruments that are rated for use under these conditions?
A. ERFDADS only.
B. Channel 'B' instruments only.
C. White placard instruments only.
D. Channel 'A' instruments located on B02 only.
Answer: C
Rev 0 Question 73 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 2.00 System ID: 43739 User-Defined ID: Q43739 Cross Reference Number:
Topic: Q43739 Post Accident Instruments Identification Num Field 1: 3.7 Num Field 2: 3.9 Text Field: 2.4.3 Comments: Proposed reference to be provided to applicant during examination:
NONE Technical
Reference:
40DP-9AP16 K&A:
Ability to identify post-accident instrumentation.
Justification:
A. Incorrect - ERFDADS is not rated for LOCA.
B. Incorrect - some B instruments are post accident qualified but not exclusive, examine may confuse fire rated with post LOCA C. Correct - Per 40DP-9AP16 sect 23.1.
D. Incorrect - these are PAM instrument found in Tech Specs but not all the LOCA rated instruments in the CR.
Rev 0 Question 73 Table-Item Links Q - 10CFR Sections 55.41 (6) Design, components, and functions of reactivity control mechanisms and instrumentation.
Q - Reactor Operator K & A Importance classification 3.7 Q - Cognitive Level Comprehension / Anal Q - Question Source New Q - RO Exam 2010 RO NRC Associated objective(s):
Given that an ORP is being implemented describe how a parameter's value should be obtained in accordance with 40DP-9AP16.
Rev 0 74 ID: Q43780 Points: 1.00 Given the following conditions AND COLR Figure 3.1.5-1:
Unit 1 is operating at 76% power.
A Four (4) Fingered CEA slips into the core 15 inches at time 1500.
Which ONE of the following will meet the requirements of the COLR?
The Reactor is at ____(1)____ % power at time ____(2)____.
A. (1) 70 (2) 1535 B. (1) 67 (2) 1538 C. (1) 66 (2) 1547 D. (1) 63 (2) 1555 Answer: B Question 74 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 4.00 System ID: 43780 User-Defined ID: Q43780 Cross Reference Number:
Topic: Q43780 GEN COLR SLIP CEA Num Field 1: 3.9 Num Field 2: 4.2 Text Field: 2.1.25 Comments: Proposed reference to be provided to applicant during examination: COLR Figure 3.1.5-1 Unit 1 Rev 21.
Technical
Reference:
COLR Figure 3.1.5-1 Unit 1 Rev 21.
K&A:
Ability to interpret reference materials, such as graphs, curves, tables, etc Justification:
A. Incorrect - 6 % change must be completed within 30 minutes.
B. Correct - 9 % change must be completed within 40 minutes.
C. Incorrect - 10 % change must be completed within 45 minutes.
D. Incorrect - 13 % change must be completed within 52 minutes.
Rev 0 Question 74 Table-Item Links Q - 10CFR Sections 55.41 (10) Administrative, normal, abnormal, and emergency operating procedures for the facility.
Q - Reactor Operator K & A Importance classification 3.9 Q - Cognitive Level Comprehension / Anal Q - Question Source New O - RO Tier Group Designations Tier 3 Q - RO Exam 2010 RO NRC Associated objective(s):
Describe how the plant is stabilized to counter the effect of the dropped/slipped CEA.
Rev 0 75 ID: Q7166 Points: 1.00 Given the following conditions:
Unit 1 is operating at 50% power coming out of a refueling outage.
Unit 2 is operating at 100% power.
Unit 3 is in Mode 3 to repair a Hydrogen Leak on the Main Turbine.
Severe weather conditions are occurring that require implementation of EPIP-01 (STSC Actions).
Which ONE of the following describes the person responsible for initially classifying the emergency and assuming the position of On-shift Emergency Coordinator?
A. Unit 1 Shift Manager.
B. Unit 2 Shift Manager.
C. Unit 3 Shift Manager.
D. Unit 3 Work Control SRO.
Answer: A
Rev 0 Question 75 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 2.00 System ID: 7166 User-Defined ID: Q7166 Cross Reference Number:
Topic: Q7166 EC for Multiple Unit Classification Num Field 1: 3.0 Num Field 2: 4.1 Text Field: 2.4.37 Comments: Proposed reference to be provided to applicant during examination: NONE Technical
Reference:
EPIP-01, STSC ACTIONS K&A:
Knowledge of the lines of authority during an emergency.
Justification:
A. Correct - Unit 1 SM is the EC for all multi unit events unless another Unit was already in an EAL and a EC was established.
B. Incorrect - Unit 2 is the only unit not performing an evolution.
Candidate may think that Unit 2 is the choice due to being at a steady state condition.
C. Incorrect - Unit 3 is in an outage. Candidate may think that Unit 3 is the choice due to being in an outage with additional personnel to support.
D. Incorrect - Unit 3 is in an outage. Candidate may think that Unit 3 WC SRO is the choice due to being available to become EC.
Rev 0 Question 75 Table-Item Links Q - 10CFR Sections 55.41 (10) Administrative, normal, abnormal, and emergency operating procedures for the facility.
Q - Reactor Operator K & A Importance classification 3.0 Q - Cognitive Level Memory Q - Question Source PV Bank Not Modified O - RO Tier Group Designations Tier 3 Q - RO Exam 2010 RO NRC Associated objective(s):
Identify the on-shift responsibilities for emergency plan implementation including who in the on shift organization is designated as the emergency coordinator.
Rev 2 1 ID: Q22629 Points: 1.00 Given the following conditions:
Initial Conditions Unit 1 tripped due to a Loss of Offsite Power (LOOP).
SPTAs are in progress.
Subsequently:
RCS pressure is 1800 psia and trending down slowly.
Containment temperature is 150°F and rising slowly.
Containment pressure is 2.7 psig and rising slowly.
RU-1 (CTMT ATMOSPHERE) is in ALERT and rising slowly.
Pressurizer level is 8% and stable.
RCS subcooling is 40°F and stable.
AFB-P01 (Essential Motor Driven Aux Feed Pump) is feeding both Steam Generators.
Steam Generator #1 is 50% WR and rising.
Steam Generator #2 is 45% WR and rising.
Which ONE of the following describes the appropriate response?
The CRS will direct entering A. 40EP-9EO03 (LOCA) and re-initiate HPSI flow to restore PZR level.
B. 40EP-9EO05 (ESD) and isolate #2 SG to prevent CTMT overpressure.
C. 40EP-9EO04 (SGTR) and feed #1SG 1360 - 1600 gpm to restore SG level.
D. 40EP-9EO07 (LOOP/LOFC) and isolate RCP seal bleed off to protect the RCP seals.
Answer: A
Rev 2 Question 1 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 2.00 System ID: 22629 User-Defined ID: Q22629 Cross Reference Number:
Q22629 SRO EOP GEN analyze whether it is permissible to throttle Topic:
HPSI flow.
Num Field 1: 4.2 Num Field 2: 4.4 Text Field: 2.2.44 Comments: Proposed reference to be provided to applicant during examination: NONE Technical
Reference:
40EP-9EO10, Appendix 2 figures, 40EP-9EO03, LOCA.
K&A:
Ability to interpret control room indications to verify the status and operation of a system, and understand how operator actions and directives affect plant and system conditions.
Justification:
A. Correct - Pzr Level criteria is not met per standard appendix 2 page 3 of 3. Parameters are consistent with a LOCA inside CTMT.
CTMT conditions are not harsh.
B. Incorrect - ESD is differentiated by a higher subcooling and rising trend, isolating the faulted SG is the correct mitigation strategy per ESD EOP. ESD is plausible due to the energy being released to CTMT is an indication of ESD.
C. Incorrect - CTMT conditions do not support a SGTR. RCS leak is evident, during a SGTR the LOCA will be contained in the secondary system, therefore the CTMT environment would not change. The indicated level mismatch is easily explained due to normal plant response post trip.
D. Incorrect - LOOP has occurred which makes this a plausible distracter, only the LOOP EOP does not address the RC Inventory Control Safety Function therefore LOOP is not the appropriate EOP.
Rev 2 Question 1 Table-Item Links Q - 10CFR Sections 55.43 (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
Q - Senior Reactor Operator K & A Importance classification 4.4 Q - Cognitive Level Comprehension / Anal Q - Question Source New Q - SRO Exam SRO Exam 2010 O - SRO Tier Group Designation Tier 1 Group 1 Associated objective(s):
analyze whether it is permissible to throttle HPSI flow.
Rev 1 2 ID: Q43750 Points: 1.00 Given the following conditions:
Unit 2 has been operating at 100% power for 7 days following a refueling outage.
A small leak has been identified on the discharge piping of CHA-P01, charging pump A.
Charging has been secured until the leak can be isolated.
Letdown has been secured due to the charging pump leak.
Pressurizer level is 50% and lowering slowly.
NCN-P01A, Nuclear Cooling Water pump A is under clearance and unavailable.
Subsequently:
While taking down scaffolding the supply breaker for 4.16 KV bus NBN-S02 is inadvertently tripped.
RCP LO NC FLOW alarms are LIT on B04.
Which ONE of the following actions is required?
A. All RCPs must be secured within 3 minutes unless cooling water flow is restored per 40A0-9ZZ03 (Loss of Cooling Water).
B. If charging flow is restored within 3 minutes then the RCPs can run indefinitely per 40A0-9ZZ04 (RCP Motor Emergencies).
C. If charging flow is restored within 10 minutes then the RCPs can run indefinitely per 40A0-9ZZ04 (RCP Motor Emergencies).
D. All RCPs must be secured within 10 minutes unless cooling water flow is restored per 40A0-9ZZ03 (Loss of Cooling Water).
Answer: A
Rev 1 Question 2 Info Question Type: Multiple Choice Status: Active Always select No on test?
Authorized for No practice?
Points: 1.00 Time to 3
Complete:
Difficulty: 3.00 System ID: 43750 User-Defined Q43750 ID:
Cross Reference Number:
Topic: Q43750 SRO when to secure RCPs Num Field 1: 3.7 Num Field 2: 3.7 Text Field: 4.2 015/017 AA2.10 Comments: Proposed reference to be provided to applicant during examination:
NONE Technical
Reference:
40AO-9ZZ03, Loss of Cooling Water AOP.
K&A: Ability to determine and interpret the following as they apply to the Reactor Coolant Pump Malfunctions (Loss of RC Flow): When to secure RCPs on loss of cooling or seal injection Justification:
A. Correct - Candidate must determine that the RCPs are with out seal injection. This is the correct action for a Loss of NC flow with no seal injection available. Loss of Cooling Water AOP will direct this action. Loss of NBN-S02 results in a Loss of Power to NCN-P01B (NCW Pump B). This causes a loss on NC Flow.
B. Incorrect - RCP operation is allowed for 10 minutes with seal injection flow available without NC flow. RCP Emergencies directs you to the Loss of Cooling Water AOP.
C. Incorrect - RCP operation is allowed for 10 minutes with seal injection flow available without NC flow. RCP Emergencies directs you to the Loss of Cooling Water AOP.
D. Incorrect - RCP operation is allowed for 10 minutes with seal injection flow available without NC flow. Loss of Cooling Water AOP would correct but the time limit is wrong.
Rev 1 Question 2 Table-Item Links Q - 10CFR Sections 55.43 (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
Q - Senior Reactor Operator K & A Importance classification 3.7 Q - Cognitive Level Comprehension / Anal Q - Question Source New Q - SRO Exam SRO Exam 2010 O - SRO Tier Group Designation Tier 1 Group 1 Associated objective(s):
describe the limitations on RCP operation without NC
Rev 1 3 ID: Q10275 Points: 1.00 Given the following conditions:
SPTAs are in progress.
525 kV switchyard voltmeters East and West indicate 000 volts.
Both EDGs have powered their respective class buses.
CDN-PI-47 indicates that Main Condenser pressure is 12 inches HgA and continuing to rise.
SG levels are being restored by AFB-P01 (Ess Motor Driven Aux Feed Pump).
0 RCS Tcold is being controlled at 562 F with ADVs.
Ess Spray Pond pump "B" has tripped due to an 86 lockout.
RO attempted to start AFN-P01 (Non-Ess Motor Driven Aux Feed Pump), but all attempts to open CTA-HV-1 (AFN-P01 Suction Valve) have failed.
AFA-P01 (Ess Turbine Driven Aux Feed Pump) is running with a maximum discharge pressure of 900 psia.
Which ONE of the following describes the appropriate success path?
The CRS should enter...
A. 40EP-9EO02 (Reactor Trip) and feed using AFB-P01.
B. 40EP-9EO09 (Functional Recovery) and feed using AFB-P01.
C. 40EP-9EO06 (Loss of all Feedwater) and feed using condensate pumps.
D. 40EP-9EO07 (Loss of Offsite Power/Loss of Forced Circulation) and feed using AFA-P01.
Answer: B
Rev 1 Question 3 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 4 Difficulty: 3.00 System ID: 10275 User-Defined ID: Q10275 Cross Reference Number:
Topic: Q10275 - analyze whether or not entry into the LOAF EOP is appropriate Num Field 1: 4.2 Num Field 2: 4.4 Text Field: 2.2.44 Comments: Proposed reference to be provided to applicant during examination:
Steam Tables Technical
Reference:
EOP Operations Expectations & 40EP-9EO01, Standard Post Trip Actions K&A:
Ability to interpret control room indications to verify the status and operation of l a system, and understand how operator actions and directives affect plant and l system conditions. Loss of Offsite Power Justification:
A. Incorrect - AFB-P01 is powered from PBB-S04. The Loss of Offsite Power indicated by the 000 volts has caused both EDGs to start and power their respective electrical bus. EDG B must be emergency stopped due to the 86 lockout on Ess Spray Pond Pump B.
Therefore PBB-S04 is de-energized resulting in AFB-P01 losing power. AFN-P01 is powered from PBA-S03, which is still energized but will not start due to its suction valve being failed closed.
Therefore the Rx trip EOP entry conditions are no longer met.
B. Correct - CRS should realize that only available feed source is AF pump "B" and it must be powered by the "A" DG which requires entry into the FRP.
C. Incorrect - Loss of offsite power precludes the use of Condensate pumps D. Incorrect - AFA-P01 discharge pressure is too low for a T-cold of 0
562 F. SG Pressures are 1150 psia.
Rev 1 Question 3 Table-Item Links Q - 10CFR Sections 55.43 (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
Q - Senior Reactor Operator K & A Importance classification 4.4 Q - Cognitive Level Comprehension / Anal Q - Question Source PV Bank Not Modified Q - SRO Exam SRO Exam 2007 SRO Exam 2010 O - SRO Tier Group Designation Tier 1 Group 1 Associated objective(s):
analyze whether or not entry into the LOAF EOP is appropriate
Rev 2 4 ID: Q22645 Points: 1.00 Given the following conditions:
Initial Conditions:
Unit 1 is operating at 100% power.
The quarterly Surveillance Test on EDG 'A' is in progress.
EDG 'A' is paralleled to Off-Site power, PBA-S03B (EDG A Output Breaker) is closed.
EDG 'A' is carrying 5000 KW.
Subsequently:
PKA-M41 faults and is de-energized.
Which ONE of the following describes the appropriate response?
A. Enter 40AO-9ZZ12 (Degraded Electrical) to restore PBA-S03.
B. Enter 40AO-9ZZ13 (Loss of Class Inst and Cntrl Power) to restore PBA-S03.
C. Enter 40AO-9ZZ12 (Degraded Electrical) and trip PBA-S03B due to the EDG 'A' tripping and PBA-S03B remaining closed.
D. Enter 40AO-9ZZ13 (Loss of Class Inst and Cntrl Power) and trip PBA-S03B due to the EDG 'A' tripping and PBA-S03B remaining closed.
Answer: D
Rev 2 Question 4 Info Question Multiple Choice Type:
Status: Active Always select No on test?
Authorized for No practice?
Points: 1.00 Time to 3
Complete:
Difficulty: 2.00 System ID: 22645 User-Defined Q22645 ID:
Cross Reference Number:
Topic: Q22645 SRO APE LOSS of DC Num Field 1: 4.6 Num Field 2: 4.6 Text Field: 2.1.20 Comments: Proposed reference to be provided to applicant during examination: NONE Technical
Reference:
40AO-9ZZ13, Loss of Class Instrument and Control Power.
K&A:
Ability to interpret and execute procedure steps. Loss of DC.
Justification:
A. Incorrect - PBA-S03 is still supplied by offsite power. Therefore entry conditions for Degraded Electrical AOP are not met.
Operator may think Loss of Class DC control power will cause EDG to trip. Similar to Loss of Non Class DC bus NKN-M45.
B. Incorrect - Loss of Class instrument and control power will not address direct restoration of PBA-S03.
C. Incorrect - Degraded Electrical AOP does not provide direction to trip the EDG.
D. Correct - EDG A is now being back fed from the grid due to the Loss of tripping power resulting in the motoring of the EDG A, must trip the EDG A output breaker to prevent damage to the EDG per Sect 3.0 step 2 of 40AO-9ZZ13.
Rev 2 Question 4 Table-Item Links Q - 10CFR Sections 55.43 (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
Q - Senior Reactor Operator K & A Importance classification 4.6 Q - Cognitive Level Comprehension / Anal Q - Question Source New Q - SRO Exam SRO Exam 2010 O - SRO Tier Group Designation Tier 1 Group 1 Associated objective(s):
determine if the Loss of Class Instrument or Control Power AOP should be executed
Rev 2 5 ID: Q9469 Points: 1.00 Given the following conditions:
Unit 1 has tripped from 100% power.
SPTAs are in progress.
RCS pressure 2150 psia and rising.
Tcold 560°F and rising.
DFWCS is in Reactor Trip Override with 0 gpm to each SG.
13.8 Kv non class bus NAN-S02 has a fault and is de-energized.
DG "B" is out of service for scheduled maintenance.
Class 125 VDC Battery Bus PKC-M43 is de-energized due to an unrelated fault.
Which ONE of the following recovery procedures is designed specifically to mitigate this set of symptoms?
A. 40EP-9EO06 (LOAF).
B. 40EP-9EO02 (Reactor Trip).
C. 40EP-9EO07 (LOOP/LOFC).
D. 40EP-9EO09 (FRP - MVDC).
Answer: B
Rev 2 Question 5 Info Question Type: Multiple Choice Status: Active Always select No on test?
Authorized for No practice?
Points: 1.00 Time to 2
Complete:
Difficulty: 4.00 System ID: 9469 User-Defined Q9469 ID:
Cross Reference Number:
Q9469 SRO CE EPE whether or not entry into the Reactor Trip EOP is Topic:
appropriate Num Field 1: 2.7 Num Field 2: 3.7 Text Field: 4.4 E02 EA2.1 Comments: Proposed reference to be provided to applicant during examination:
NONE Technical
Reference:
40EP-9EO02, Reactor Trip procedure / 40EP-9EO01, SPTAs K&A:
Ability to determine and interpret the following as they apply to the (Reactor Trip Recovery): Facility conditions and selection of appropriate procedures during abnormal and emergency operations.
Justification:
A. Incorrect - Feedrate is correct for current plant conditions. RTO 0
does not feed with Tavg < 564 F. Operator may see 0 feedrate and choose LOAF EOP.
B. Correct - The indications are within the Rx trip Entry conditions and Safety Functions, therefore entering Rx Trip EOP is correct.
C. Incorrect - Indications of a Partial Loss of Offsite power, NAN-S01 is not listed as de-energized therefore LOOP is not the correct EOP.
D. Incorrect - PBB-S04 is still energized from offsite power, the candidate may interpret the electric plant status not to meet the MVDC safety function in Rx Trip. MVDC safety function directs restoration of a vital DC bus. If examinee believes that PBB-S04 was de-energized due to the loss of PKA-M43, this would appear to be correct.
Rev 2 Question 5 Table-Item Links Q - 10CFR Sections 55.43 (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
Q - Senior Reactor Operator K & A Importance classification 3.7 Q - Cognitive Level Comprehension / Anal Q - Question Source PV Bank Not Modified Q - SRO Exam SRO Exam 2009 SRO Exam 2010 O - SRO Tier Group Designation Tier 1 Group 1 Associated objective(s):
analyze whether or not entry into the Reactor Trip EOP is appropriate
Rev 0 6 ID: Q10273 Points: 1.00 Given the following conditions:
Unit 1 tripped from 100% power.
SIAS, CIAS, MSIS, and AFAS-1 have initiated.
The RCS is in natural circulation due to a Loss of Offsite power (LOOP).
Steam Generator #1 pressure is 750 psia and dropping rapidly.
Steam Generator #1 level is 8% WR and dropping rapidly Steam Generator #2 pressure is 940 psia and slowly lowering.
Steam Generator #2 level is 37% WR and slowly lowering.
Auxiliary feedwater flow is > 2000 gpm to steam generator #1.
Auxiliary feedwater flow is 0 gpm to steam generator #2.
HPSI pump "B" tripped due to a ground fault.
AFN-P01 tripped due SIAS initiation.
SPTAs have just been completed.
Which ONE of the following EOPs should the CRS enter?
A. 40EP-9EO05 (Excessive Steam Demand).
B. 40EP-9EO07 (Loss of Offsite Power/Loss of Forced Circulation).
C. 40EP-9EO09 (Functional Recovery Procedure) - HR-2 is jeopardized.
D. 40EP-9EO09 (Functional Recovery Procedure) - MVAC-2 is jeopardized.
Answer: A
Rev 0 Question 6 Info Question Multiple Choice Type:
Status: Active Always select on No test?
Authorized No for practice?
Points: 1.00 Time to 3
Complete:
Difficulty: 3.00 System ID: 10273 User-Defined Q10273 ID:
Cross Reference Number:
Topic: Q10273 EOP Select the ESD procedure Num Field 1: 2.7 Num Field 2: 4.0 Text Field: 4.4 E05 EA2.1 Comments: Proposed reference to be provided to applicant during examination:
NONE Technical
Reference:
EOP Operations Expectations & 40EP-9EO01, Standard Post Trip Actions K&A:
Facility conditions and selection of appropriate procedures during abnormal and emergency operations. Steam Line Rupture - Excessive Heat Transfer Justification:
A. Correct - SIAS, CIAS, MSIS and AFAS-1 are consistent with an inside CTMT ESD on SG 1. There should be no AFW flow to SG 1 due to a high D/P lockout at 185 psia, but it has failed.
B. Incorrect - Examinee may choose LOOP/LOFC due to the Loss of Offsite Power.
C. Incorrect - Examinee may choose HR-2 due to no AFW flow to SG 2 and the tripping of AFN-P01. Excessive AFW flow to #1 SG should be corrected by the board operator once it is known. This is not expected.
D. Incorrect - Examinee may choose MVAC-2 due to the tripping of AFN and HPSI pump B.
Rev 0 Question 6 Table-Item Links Q - 10CFR Sections 55.43 (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
Q - Senior Reactor Operator K & A Importance classification 4.0 Q - Cognitive Level Comprehension / Anal Q - Question Source PV Bank Not Modified Q - SRO Exam SRO Exam 2010 O - SRO Tier Group Designation Tier 1 Group 1 Associated objective(s):
describe the mitigating strategy outlined in the ESD EOP
Rev 0 7 ID: Q10381 Points: 1.00 Given the following conditions:
Unit 1 tripped on Low SG pressure caused by a steam line break.
SG 1 pressure is currently 1170 psia being controlled by ADVs.
SG 2 pressure is currently 1180 psia being controlled by ADVs.
4160 bus PBA-S03 has faulted and is de-energized.
Both SG levels are 20% WR and slowly lowering.
AFA-P01 has tripped on overspeed.
AFB-P01 has tripped due to an 86 lock-out.
Feed rate to both SGs is 0 gpm.
All appropriate ESFAS actuations have initiated.
Which ONE of the following is appropriate to restore the RCS Heat Removal Safety Function?
A. Establish feed using a Main Feedwater pump after 40EP-9EO06 (LOAF) is entered.
B. Reset AFA-P01 overspeed trip and establish feed prior to exiting 40EP-9EO01 (SPTAs).
C. Depressurize a SG and establish feed with a Condensate pump after 40EP-9EO09 (FRP) is entered D. Override the Downcomer isolation valves and establish feed using AFN-P01 prior to exiting 40EP-9EO01 (SPTAs).
Answer: C
Rev 0 Question 7 Info Question Multiple Choice Type:
Status: Active Always select on No test?
Authorized for No practice?
Points: 1.00 Time to 3
Complete:
Difficulty: 4.00 System ID: 10381 User-Q10381 Defined ID:
Cross Reference Number:
Topic: Q10381 recover the RCS heat sink Num Field 4.3 1:
Num Field 4.6 2:
Text Field: 4.1 074 EA2.02 Comments: Proposed reference to be provided to applicant during examination:
NONE Technical
Reference:
EOP Operations Expectations, 40EP-9EO01 (SPTAs). Relay Resetting (40DP-9OP02)
K&A: Ability to determine or interpret the following as they apply to a Inadequate Core Cooling: Availability of main or auxiliary feedwater Justification:
A. Incorrect - MFW is not available due to the SLB. Candidate may believe that only one steam line is affected and choose to utilize a MFP in LOAF.
B. Incorrect - Resetting AFA in SPTAs is not allowed per procedure.
This is a Standard Appendix to the EOPs that will be performed once the SPTAs are exited.
C. Correct - Due to the lack of Aux Feed available and the MSIS resulting in a loss of MFW and condenser vacuum, the FRP will direct depressurizing and utilizing CD pumps to restore SG levels.
D. Incorrect - AFN is supplied by PBA-S03 therefore the pump does not have electrical power to operate. This action would be performed during the performance of SPTAs if possible.
Rev 0 Question 7 Table-Item Links Q - 10CFR Sections 55.43 (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
Q - Senior Reactor Operator K & A Importance classification 4.6 Q - Cognitive Level Comprehension / Anal Q - Question Source PV Bank Not Modified Q - SRO Exam SRO Exam 2007 SRO Exam 2010 O - SRO Tier Group Designation Tier 1 Group 2 Associated objective(s):
determine the major mitigating strategies contained in 40EP-9EO06
Rev 2 8 ID: Q22640 Points: 1.00 Given the following conditions:
Unit 1 is operating at 100% power.
Pressurizer Level Master Controller is in REMOTE and AUTO..
Pressurizer Level Control selector switch is in the Y position.
RRS is aligned to monitor the average Tavg.
RC N- TT-111Y (Tcold) input into the RRS fails low.
Tavg/Tref HI-LO (4A08B) Alarm is in.
Which ONE of the following describes how the Pressurizer Level Control System (PLCS) will respond, and the appropriate actions to be taken?
PLCS setpoint goes to...
A. Maximum, place the PLCS Master Controller in Manual per 40AO-9ZZ16 (RRS Malfunctions).
B. Minimum, place the PLCS Master Controller in Local Automatic per 40AO-9ZZ16 (RRS Malfunctions).
C. Minimum, shift the PLCS selector switch to the X position per 40AL-9RK4A (B04A Alarm Response).
D. Maximum, shift the PLCS selector switch to the X position per 40AL-9RK4A (B04A Alarm Response).
Answer: B
Rev 2 Question 8 Info Question Multiple Choice Type:
Status: Active Always select No on test?
Authorized No for practice?
Points: 1.00 Time to 3
Complete:
Difficulty: 4.00 System ID: 22640 User-Defined Q22640 ID:
Cross Reference Number:
Topic: Q22640 SRO Tcold Failing Low effect on PLCS Num Field 1: 3.3 Num Field 2: 3.4 Text Field: 4.2 028 AA2.10 Comments: Proposed reference to be provided to applicant during examination:
NONE Technical
Reference:
40AO-9ZZ16, RRS Malfunction, Operator Information Manual K&A:
Ability to determine and interpret the following as they apply to the Pressurizer Level Control Malfunctions: Whether the automatic mode for PZR level control is functioning improperly, necessity of shift to manual modes.
Justification:
A. Incorrect - Plausible since the setpoint will go to minimum output.
Action is correct.
B. Correct - The setpoint will go to minimum and RRS Malfunction AOP directs placing the master controller in local auto or manual control mode.
C. Incorrect - Plausible since the alarm response directs this operator action for a transmitter failure.
D. Incorrect - Plausible since the setpoint will go to minimum and the alarm response directs this operator action for a transmitter failure.
Rev 2 Question 8 Table-Item Links Q - 10CFR Sections 55.43 (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
Q - Senior Reactor Operator K & A Importance classification 3.4 Q - Cognitive Level Comprehension / Anal Q - Question Source New Q - SRO Exam SRO Exam 2010 O - SRO Tier Group Designation Tier 1 Group 2 Associated objective(s):
describe the impact that a temperature instrument failing LOW would have on the various control systems that rely on this indication
Rev 0 9 ID: Q43783 Points: 1.00 Given the following conditions:
Initially:
Unit 1 was operating at 100% power prior to the trip.
A RCS Leak has developed into the RCP 2A HPSC.
SPTAs are complete and the CRS has entered 40EP-9EO03 (LOCA).
Subsequently:
A Loss of Offsite Power (LOOP) has occurred.
EDG B has tripped on Overspeed.
SIAS, CIAS, MSIS and CSAS have initiated.
SESS Panel has the following alarms:
BLUE SEAS and WHITE SEIS Light:
o NCB-UV-401 (NC Supply External CTMT Isolation Valve).
o NCB-UV-403 (NC Return Internal CTMT Isolation Valve).
BLUE SEAS Light :
o NCA-UV-402 (NC Return External CTMT Isolation Valve).
Which ONE of the following describes the appropriate response?
A. Stay in 40EP-9EO03 (LOCA), close NCB-UV-401/403 (NCW CTMT Isolation valves).
B. Stay in 40EP-9EO03 (LOCA), close RCN-HV-448/452 (RCP 2A HPSC Isolation Valves).
C. Go to 40EP-9EO09 (FRP), cross tie EDG A to PBB-S04 and close NCB-UV-401/403 (NCW CTMT Isolation valves).
D. Go to 40EP-9EO09 (FRP), cross tie EDG A to PBB-S04 and close RCN-HV-448/452 (RCP 2A HPSC Isolation Valves).
Answer: C
Rev 0 Question 9 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 4.00 System ID: 43783 User-Defined ID: Q43783 Cross Reference Number:
Topic: Q43783 CTMT INTEGRITY NCW ALARMS Num Field 1: 4.2 Num Field 2: 4.2 Text Field: 2.4.46 Comments: Proposed reference to be provided to applicant during examination:
NONE Technical
Reference:
K&A:
Ability to verify that the alarms are consistent with the plant conditions:
Loss of CTMT Integrity.
Justification:
A. Incorrect - NCB-UV-401/403 do not have electrical power, and LOCA procedure does not provide for restoring bus power to the required electrical bus.
B. Incorrect - Closing RCN-HV-448/452 will isolate the leak but will not meet the Containment Isolation safety function.
C. Correct - This will provide the electrical power required to restore power to NCB-UV-401/403 to allow them to be closed to Meet Containment Isolation.
D. Incorrect - This will provide the electrical power required to restore power to NCB-UV-401/403 but the action of closing RCN-HV-448/452 will isolate the leak but will not meet the Containment Isolation safety function.
Rev 0 Question 9 Table-Item Links Q - 10CFR Sections 55.43 (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
Q - Senior Reactor Operator K & A Importance classification 4.2 Q - Cognitive Level Comprehension / Anal Q - Question Source New Q - SRO Exam SRO Exam 2010 O - SRO Tier Group Designation Tier 1 Group 2 Associated objective(s):
Given plant conditions determine whether or not entry into the FRP is appropriate in accordance with 40EP-9EO09.
Rev 1 10 ID: Q43809 Points: 1.00 Which ONE of the following events REQUIRES the Nuclear Regulatory Commission be notified?
A. Manual reactor trip during a shutdown entering a refueling outage.
B. Steam Generator Tube Leak of approximately 275 gallons per day (GPD).
C. Reactor Protection System instrument placed in BYPASS within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> of inoperability.
D. Stroking SGE-UV-170 instead of SGE-UV-180 during 73ST-9SG01 (MSIVs -
INSERVICE TEST).
Answer: B
Rev 1 Question 10 Info Question Multiple Choice Type:
Status: Active Always select on No test?
Authorized for No practice?
Points: 1.00 Time to 3
Complete:
Difficulty: 4.00 System ID: 43809 User-Q43809 Defined ID:
Cross Reference Number:
Topic: Q43809 Event requiring reporting Num Field 2.7 1:
Num Field 4.1 2:
Text Field: 2.4.30 Comments: Proposed reference to be provided to applicant during examination:
NONE Technical
Reference:
NUREG 1022, Event Reporting Manual.
K&A:
Knowledge of which events related to system operations/status that must be reported to internal organizations or external agencies, such as the State, the NRC, or the transmission system operator. Steam Generator Tube Leak.
Justification:
A. Incorrect - Planned Manual Reactor trips IAW plant procedures is not a reportable event.
B. Correct - This event is a Tech Spec required shutdown condition which requires NRC notification.
C. Incorrect - Tech Spec 3.3.1 Condition A is met, and does not require a shutdown.
D. Incorrect - Specifically described in NUREG 1022 that a notification is not required.
Rev 1 Question 10 Table-Item Links Q - 10CFR Sections 55.43 (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
Q - Senior Reactor Operator K & A Importance classification 4.1 Q - Cognitive Level Comprehension / Anal Q - Question Source New Q - SRO Exam SRO Exam 2010 O - SRO Tier Group Designation Tier 1 Group 2 Associated objective(s):
Describe the event reporting process
Rev 1 11 ID: Q22527 Points: 1.00 Given the following conditions:
Initial Conditions:
Unit 1 is operating at 100% power.
NAN-S02 (13.8 Kv non class bus) faults and is de-energized.
An automatic reactor trip occurs.
Standard Post Trip Actions (SPTAs) are completed.
Pressurizer level is 33% and stable.
Subsequently:
Start-up transformer NAN-X03 faults and is de-energized.
The CRS should now anticipate that Pressurizer level will ...
A. rise above the control band, level should be controlled by initiating a cooldown per 40EP-9EO07 (LOOP/LOFC).
B. rise above the control band, level should be controlled by securing all charging pumps per 40EP-9EO02 (Reactor Trip).
C. drop below the control band, level should be controlled by placing the always running charging pump in the AUTO-AFTER STOP position per 40EP-9EO02 (Reactor Trip).
D. drop below the control band, level should be restored by placing the always running charging pump in the AUTO-AFTER STOP position per 40EP-9EO07 (LOOP/LOFC).
Answer: A
Rev 1 Question 11 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 3.00 System ID: 22527 User-Defined ID: Q22527 Cross Reference Number:
Topic: Q22527 affect of a LOOP on the RCS Num Field 1: 4.1 Num Field 2: 4.3 Text Field: 3. 4 002 A2.03 Comments: Proposed reference to be provided to applicant during examination:
NONE Technical
Reference:
LOOP, 40EP-9EO07/ LOIT lesson plan K&A: Ability to (a) predict the impacts of the following malfunctions or operations on the RCS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Loss of forced circulation Justification:
A. Correct - Increasing Thot and loss of Letdown will cause Pzr level to go high. With no cooling flow (NC) to the RCPs charging flow must be maintained, level is controlled by initiating a cooldown. NAN-S02 is faulted which is the normal 13.8 kV bus for Unit 1. When NAN-X03 faults, this causes NAN-S01 to de-energize which results in no 13.8 kV power in Unit 1 Which is a Loss of Offsite Power (LOOP) and this results in a Loss of Forced Circulation (LOFC), due to no RCPs operating.
B. Incorrect - This is a LOOP event, stopping charging pumps is wrong.
C. Incorrect - level will go high. Examinee may not correctly associate which buses have de-energized or think that charging pumps have become anti-pumped due to the LOP signals. The Reactor Trip procedure allows for maintaining PZR level manually which would direct the operator to correct the Charging Pump configuration to maintain level.
D. Incorrect - level will go high. Examinee may not correctly associate which buses have de-energized or think that charging pumps have become anti-pumped due to the LOP signals. The LOOP/LOFC procedure will allow for maintaining PZR level by cooling down due to no NCW pumps and Charging Pumps required for seal injection.
Rev 1 Question 11 Table-Item Links Q - 10CFR Sections 55.43 (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
Q - Senior Reactor Operator K & A Importance classification 4.3 Q - Cognitive Level Comprehension / Anal Q - Question Source PV Bank Not Modified Q - SRO Exam SRO Exam 2009 SRO Exam 2010 O - SRO Tier Group Designation Tier 2 Group 2 Associated objective(s):
determine whether or not entry into the LOOP/LOFC EOP is appropriate
Rev 0 12 ID: Q43791 Points: 1.00 Given the following conditions:
Initial Conditions:
Unit 2 is operating at 85% power with both Circulating Water (CW) Loops in service.
A Condenser tube leak in the A-B Circulating Water Loop has been confirmed.
A normal downpower is in progress in order to remove the ruptured CW train from service.
The ability to stabilize power at 40% has been verified.
At the present rate 40% power will be achieved in approximately 50 minutes.
Subsequently:
Chemistry reports that Steam Generator sodium (Na) and chloride (Cl) are rising rapidly.
It is determined that Steam Generator sodium will be 1.0 ppm in approximately 20 minutes.
Based on this report, which ONE of the following actions should the CRS direct?
A. Trip the Reactor, Perform SPTAs, trip both Main Feedwater pumps.
B. Trip the Reactor, Perform SPTAs, align high rate blowdown to the Blowdown Flash Tank (BFT).
C. Initiate Reactor Power Cutback (Loss of Feedpump), align high rate blowdown to the Blowdown Flash Tank (BFT).
D. Initiate Reactor Power Cutback (Loss of Feedpump), place all available Condensate Demineralizer beds in service.
Answer: A
Rev 0 Question 12 Info Question Multiple Choice Type:
Status: Active Always select No on test?
Authorized for No practice?
Points: 1.00 Time to 2
Complete:
Difficulty: 4.00 System ID: 43791 User-Defined Q43791 ID:
Cross Reference 40AO-9ZZ10 Number:
Topic: Q43791 Rx trip based on chemistry trend Num Field 1: 2.1 Num Field 2: 2.5 Text Field: 3.4 056 A2.05 Comments: Proposed reference to be provided to applicant during examination: NONE Technical
Reference:
40AO-9ZZ10, Condenser Tube Rupture K&A: Ability to (a) predict the impacts of the following malfunctions or operations on the Condensate System; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Condenser tube leakage.
Justification:
A. Correct - Per step 3.5 of Cond Tube Rupture 40AO-Correct - Per step 3.5 of Cond Tube Rupture 40AO-9ZZ10 and App M.
B. Incorrect - Trip the Rx is correct, but HR BD is not.
Candidate may think that since Sect 4.0 describes the use of High Rate Blowdown that it is the correct action.
40OP-9SG03 is the procedure for performing High Rate Blowdown.
C. Incorrect - CRS may believe that a rapid downpower is warranted and initiate a RPCB per step 13. This is incorrect due to trending to a trip setpoint.
D. Incorrect - Placing all Demins in service is directed in step 3.7-10, which will help with the Sodium ingress to the SGs, but the RPCB is incorrect.
Rev 0 Question 12 Table-Item Links Q - 10CFR Sections 55.43 (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
Q - Senior Reactor Operator K & A Importance classification 2.5 Q - Cognitive Level Comprehension / Anal Q - Question Source PV Bank Not Modified Q - SRO Exam SRO Exam 2007 SRO Exam 2010 O - SRO Tier Group Designation Tier 3 Question Used NRC SRO 2007 Associated objective(s):
Determine the actions required if it is unlikely that the leak can be isolated using a normal downpower before SG samples reach 1.0 ppm
Rev 0 13 ID: Q43790 Points: 1.00 Given the following conditions:
Unit 1 has tripped from 100% due to an event.
RVLMS Upper Head level indicates 47%.
Both S/G levels are currently at 65% WR and slowly rising.
Both S/G pressures are at 1100 psia and slowly rising.
Containment pressure is 12.5 psig and Containment spray flow is adequate.
Reactor power is 1x10-6 and slowly dropping.
VALID RU-148 reading is 1.2 E+7 mrem/HR All three charging pumps are running.
Dose Assessment is in progress.
HPSI flow is adequate.
SIAS, CIAS, and CSAS have actuated.
Given a copy of the EAL Classification Hot and Cold charts and based on the above conditions.
Which ONE of the following is the appropriate EAL classification?
A. FU1 -- Notification of Unusual Event.
B. FA1 -- Alert.
C. FS1 -- Site Area Emergency.
D. FG1 -- General Emergency.
Answer: D
Rev 0 Question 13 Info Question Multiple Choice Type:
Status: Active Always select on No test?
Authorized for No practice?
Points: 1.00 Time to 2
Complete:
Difficulty: 2.00 System ID: 43790 User-Q43790 Defined ID:
Cross Reference Number:
Topic: Q43790 EP Classification LOCA (GE -- FG1)
Num Field 2.9 1:
Num Field 4.6 2:
Text Field: 2.4.41 Comments: Proposed reference to be provided to applicant during examination:
EAL HOT/COLD Chart.
Technical
Reference:
EPIP-01 STSC ACTIONS, EAL HOT/COLD Chart.
K&A:
Knowledge of the emergency action level thresholds and classifications:
Justification:
A. Incorrect - NUE is met but it is not the highest EAL classification of the event. Candidate may confuse any of the indications and not properly apply them to the EAL chart.
B. Incorrect - Alert is met but it is not the highest EAL classification of the event. Candidate may confuse any of the indications and not properly apply them to the EAL chart.
C. Incorrect - SAE is met but it is not the highest EAL classification of the event. Candidate may confuse any of the indications and not properly apply them to the EAL chart.
D. Correct - GE is met and is the highest EAL classification of the event. A loss of Fuel Clad is indicated by the RU-148 reading > 2.1 E+05 mR/hr. A potential loss of containment is indicated by the RU-148 reading > 2.1 E+05 mR/hr. And a potential loss of RCS barrier is indicated by the fact that all 3 charging pumps are running but we have lost level in the RVUH.
Rev 0 Question 13 Table-Item Links Q - 10CFR Sections 55.43 (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
Q - Senior Reactor Operator K & A Importance classification 4.6 Q - Cognitive Level Comprehension / Anal Q - Question Source PV Bank Not Modified Q - SRO Exam SRO Exam 2010 O - SRO Tier Group Designation Tier 2 Group 2 Associated objective(s):
determine the emergency plan classification.
Rev 1 14 ID: Q22686 Points: 1.00 Given the following conditions:
Initial Conditions:
Unit 1 is operating at 50% steady state power.
Subsequently:
Letdown flow as observed on (CHN-FI-202) is lowering.
All 4 channels of PPS instrumentation for RCS T-hot are lowering.
Which ONE of the following describes these conditions and appropriate response?
A. VCT level transmitter (CH-LT-227) has failed low, take corrective action per 40AO-9ZZ01 (Emergency Boration).
B. Loop 1 T-hot instrument (RCN-TT-111X) has failed high, take corrective action per 40AO-9ZZ16 (RRS Malfunctions).
C. VCT level transmitter (CH-LT-227) has failed low, take corrective actions per alarm window 3A08B (VCT LVL LO-LO).
D. Loop 1 T-hot instrument (RCN-TT-111X) has failed high, take corrective action per alarm window 4A08B (T-AVG\ T-REF HI-LO).
Answer: C
Rev 1 Question 14 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 3.00 System ID: 22686 User-Defined ID: Q22686 Cross Reference Number:
Topic: Q22686 INADVERTENT BORATION ON CVCS Num Field 1: 4.2 Num Field 2: 4.3 Text Field: 3.2 004 A2.06 Comments: Proposed reference to be provided to applicant during examination:
NONE Technical
Reference:
40AL-9RK3A , 3A08B VCT LVL LO-LO Alarm
Response
K&A:
Ability to (a) predict the impacts of the following malfunctions or operations on the CVCS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Inadvertent boration/dilution.
Justification:
A. Incorrect - Failure is correct, the procedure entry conditions are not met and does not contain direction to correct this failure.
B. Incorrect - RCN-TT-111X (Thot) failing high will cause the Reactor Regulating System to believe that the RCS is too hot and would respond by inserting CEAs except an AMI will prevent rod motion.
This would result in the Actual RCS Thot to lower. RRS Malfunction procedure would correct this issue if it were the cause of the indications. Letdown is lowering due to the RCS cooling down resulting in a shrink in the PZR, therefore letdown flow will lower to maintain PZR level. RRS sends Tavg to PZR Level Control and a failed high instrument will send a maximum level program to PLCS which would result in letdown flow lowering.
C. Correct - LT-227 failing low will cause the charging pump suction to shift to gravity feed from the RWT, therefore initiating a constant boration. Only the alarm response procedure will address this condition.
D. Incorrect - RCN-TT-111X (Thot) failing high will cause the Reactor Regulating System to believe that the RCS is too hot and would respond by inserting CEAs except an AMI will prevent rod motion.
This would result in the Actual RCS Thot to lower. T-AVG\ T-REF HI-LO will not correct this issue. Letdown is lowering due to the RCS cooling down resulting in a shrink in the PZR, therefore letdown flow will lower to maintain PZR level. RRS sends Tavg to PZR Level Control and a failed high instrument will send a maximum level program to PLCS which would result in letdown flow lowering.
Rev 1 Question 14 Table-Item Links Q - 10CFR Sections 55.43 (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
Q - Senior Reactor Operator K & A Importance classification 4.3 Q - Cognitive Level Comprehension / Anal Q - Question Source New Q - SRO Exam SRO Exam 2010 O - SRO Tier Group Designation Tier 2 Group 1 Associated objective(s):
Explain the operation of the Volume Control Tank under normal operating conditions.
Rev 0 15 ID: Q22646 Points: 1.00 Given the following conditions:
Unit 1 has tripped from 100% power.
Sub-Cooled Margin is 30°F and lowering slowly.
Containment Pressure is 2.1 psig and rising slowly.
Containment temperature is 160° F and rising slowly.
Pressurizer level is 20% and lowering rapidly.
Reactor vessel head is reading 100% in Outlet Plenum.
RCS Pressure is 1700 psia and lowering rapidly.
SG #1 level is 40% WR, SG #2 level is 35% WR.
SG#1 and SG #2 are being fed with AFN-P01 and rising slowly.
SG pressures are 1170 psia, maintained with SBCS in automatic.
SPTAs are in progress.
Which ONE of the following describes the expected response by ESFAS to the CURRENT conditions and the appropriate procedure selection?
A. SIAS, CIAS and MSIS; Enter 40EP-9EO05 (ESD) upon exiting SPTAs.
B. SIAS and CIAS ONLY; Enter 40EP-9EO04 (SGTR) upon exiting SPTAs.
C. SIAS and CIAS ONLY; Enter 40EP-9EO003 (LOCA) upon exiting SPTAs.
D. SIAS, CIAS and MSIS; Enter 40EP-9EO009 (Functional Recovery Procedure) upon exiting SPTAs.
Answer: C
Rev 0 Question 15 Info Question Multiple Choice Type:
Status: Active Always select on No test?
Authorized for No practice?
Points: 1.00 Time to 2
Complete:
Difficulty: 2.00 System ID: 22646 User-Q22646 Defined ID:
Cross Reference Number:
Topic: Q22646 SRO SYS ESFAS Num Field 4.4 1:
Num Field 4.7 2:
Text Field: 3.2 013 A2.03 Comments Proposed reference to be provided to applicant during examination:
- NONE Technical
Reference:
40EP-9EO03, LOCA K&A:
Ability to (a) predict the impacts of the following malfunctions or operations on the ESFAS; and (b) based Ability on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations; Rapid depressurization.
Justification:
A. Incorrect - ESD entry conditions not met, Candidate may confuse the indications of LOCA with ESD. MSIS setpoint of 3.0 psig in CTMT not exceeded, candidate may confuse the SIAS/CIAS setpoint with MSIS.
B. Incorrect - SGTR entry conditions not met, candidate may confuse the difference in SG levels as a SGTR. MSIS setpoint not exceeded, candidate may confuse the SIAS/CIAS setpoint with MSIS..
C. Correct - LOCA entry conditions are met and SIAS and CIAS initiated at 1837 psia in the RCS.
D. Incorrect - MSIS set point not exceeded, candidate may confuse the SIAS/CIAS setpoint with MSIS. FRP could be chosen but not the most appropriate recovery procedure.
Rev 0 Question 15 Table-Item Links Q - 10CFR Sections 55.43 (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
Q - Senior Reactor Operator K & A Importance classification 4.7 Q - Cognitive Level Comprehension / Anal Q - Question Source New Q - SRO Exam SRO Exam 2010 O - SRO Tier Group Designation Tier 2 Group 1 Associated objective(s):
determine the major mitigating strategies contained in 40EP-9EO03
Rev 1 16 ID: Q43743 Points: 1.00 Given the following conditions:
Initial Conditions:
Unit 1 is operating at 100% power.
Subsequently:
COLSS calculated primary calorimetric power (NKBDELT) is rising.
COLSS calculated turbine power (NKTFSP) is lowering.
T-cold is 555°F and lowering.
Pressurizer pressure is 2210 psia and lowering.
Pressurizer level is lowering.
Containment pressure is 0.3 psig and stable.
Containment temperature is 112°F and stable.
Which ONE of the following events and mitigating procedures is correct?
A. Partial load reject. Implement 40AO-9ZZ08 (Load Rejection).
B. Partial load reject. Trip the reactor, complete SPTAs and implement 40EP-9EO02 (Reactor Trip).
C. Steam leak outside containment. Trip the reactor, complete SPTAs and implement 40EP-9EO05 (ESD).
D. Steam leak outside containment. Maintain reactor power less than 100% per 40OP-9ZZ05 (Power Operations).
Answer: C
Rev 1 Question 16 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 3.00 System ID: 43743 User-Defined ID: Q43743 Cross Reference Number:
Topic: Q43743 analyze whether ESD or Load Reject is appropriate Num Field 1: 3.3 Num Field 2: 3.6 Text Field: 3.4 039 A2.05 Comments: Proposed reference to be provided to applicant during examination:
NONE Technical
Reference:
40EP-9EO05, ESD EOP K&A:
Ability to (a) predict the impacts of the following malfunctions or operations on the MRSS; and (b) based on predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Increasing steam demand, its relationship to increases in reactor power.
Justification:
A. Incorrect - Turbine power lowering is an indication of a load rejection. The primary power rising does not support this. The Load Rejection AOP would be appropriate response if it was a Partial Load Rejection.
B. Incorrect - Turbine power lowering is an indication of a load rejection. The primary power rising does not support this. Tripping the reactor would only be appropriate in a Load Rejection if RCS temperature was not able to be controlled by SBCS. LR AOP would be the appropriate procedure to address a partial load rejection.
C. Correct - NKBDELT rising is indicative of steam flow rising causing RX power to rise. Since Turbine power is lowering, the steam is not going to the turbine. CTMT conditions are stable; therefore the leak is not inside the CTMT. The correct action is to trip the RX and implement ESD.
D. Incorrect - NKBDELT rising is indicative of steam flow rising causing RX power to rise. Since Turbine power is lowering, the steam is not going to the turbine. CTMT conditions are stable; therefore the leak is not inside the CTMT. The correct action is to trip the RX and implement ESD. In the past, PVNGS has operated with small steam leaks in the Turbine building while repairs and management were contacted. The severity of the steam leaks did not have a significant effect on the RCS temperature and pressure. We would not expect a crew to operate with a steam leak large enough to lower RCS pressure and temperature.
Rev 1 Question 16 Table-Item Links Q - 10CFR Sections 55.43 (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
Q - Senior Reactor Operator K & A Importance classification 3.6 Q - Cognitive Level Comprehension / Anal Q - Question Source PV Bank Not Modified Q - SRO Exam SRO Exam 2010 O - SRO Tier Group Designation Tier 2 Group 1 Associated objective(s):
analyze whether or not entry into the ESD EOP is appropriate
Rev 0 17 ID: Q22647 Points: 1.00 Given the following conditions:
Initial Conditions:
Unit 1 is operating at 100% power.
AFA-P01 (Ess Turb Driven Aux Feedpump) is out of service.
Subsequently:
Unit 1 trips.
PBA-S03 faults.
Containment Pressure is 2.1 psig and rising.
0 Containment Temperature is 113 F and rising.
Containment Humidity is rising.
Containment sump levels are rising.
PZR Pressure is 2280 psia and rising.
PZR Level is 56% and rising.
Tcold is 567 0F and rising.
SG 1 and 2 levels are 30% WR and lowering.
SPTAs are complete.
Which ONE of the following describes the appropriate EOP to enter?
A. 40EP-9EO03 (LOCA).
B. 40EP-9EO05 (ESD).
C. 40EP-9EO07 (LOAF).
D. 40EP-9EO09 (FRP for Dual Event).
Answer: B
Rev 0 Question 17 Info Question Multiple Choice Type:
Status: Active Always select on No test?
Authorized for No practice?
Points: 1.00 Time to 3
Complete:
Difficulty: 3.00 System ID: 22647 User-Q22647 Defined ID:
Cross Reference Number:
Topic: Q22647 SRO SYS GEN FEED RUPTURE Num Field 4.5 1:
Num Field 4.7 2:
Text Field: 2.4.4 Comments: Proposed reference to be provided to applicant during examination:
NONE Technical
Reference:
40EP-9EO05, ESD K&A:
Ability to recognize abnormal indications for system operating parameters which are entry-level conditions for emergency and abnormal operating procedures. Main Feedwater.
Justification:
A. Incorrect - Rising PZR level and CTMT conditions are indicative of PZR steam space LOCA. Candidate may confuse the heat up event with the Steam Space LOCA.
B. Correct - Feed header rupture is a heat up event, ESD will mitigate this event by isolating the affected SG.
C. Incorrect - SG levels lowering are indicative of a LOAF. Candidate may not understand that AFB-P01 is still available for feed; therefore the LOAF entry conditions are not met.
D. Incorrect - Due to the multiple indications given that may be confused with any of the other events, candidate may think that 2 events are in progress and enter the FRP. Only one event is in progress that is what makes the FRP wrong.
Rev 0 Question 17 Table-Item Links Q - 10CFR Sections 55.43 (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
Q - Senior Reactor Operator K & A Importance classification 4.7 Q - Cognitive Level Comprehension / Anal Q - Question Source New Q - SRO Exam SRO Exam 2010 O - SRO Tier Group Designation Tier 2 Group 1 Associated objective(s):
Given conditions of an ESD analyze whether or not entry into the ESD EOP is appropriate in accordance with 40EP-9EO05.
Rev 0 18 ID: Q43792 Points: 1.00 Given the following conditions:
Unit 1 is operating in Mode 4 coming out of a refueling outage.
RU-1 (Containment Atmosphere) is declared inoperable.
Which ONE of the following describes the required action (if any) per Technical Specification 3.4.16 (RCS Leakage Detection Instrumentation)?
A. No action required in Mode 4.
B. Perform RCS water balance inventory once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
C. No action required since containment sump monitor is operable.
D. Analyze grab samples of the containment atmosphere once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
Answer: D
Rev 0 Question 18 Info Question Multiple Choice Type:
Status: Active Always select No on test?
Authorized No for practice?
Points: 1.00 Time to 3
Complete:
Difficulty: 3.00 System ID: 43792 User-Defined Q43792 ID:
Cross Reference Number:
Topic: Q43792 TS 3.4.16 (RCS Leak Detection Instrumentation) RU-1 INOP Num Field 1: 3.4 Num Field 2: 4.7 Text Field: 2.2.40 Comments: Proposed reference to be provided to applicant during examination: NONE Technical
Reference:
Technical Specifications.
K&A:
Ability to apply Technical Specifications for a system: Process Radiation Monitoring.
Justification:
A. Incorrect - This TS applies in Modes 1-4. Candidate may think that it only applies in Modes 1 and 2.
B. Incorrect - The RCS water balance inventory is a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> action statement.
C. Incorrect - Verifying the Containment Sump monitor operable per action D. If it is INOPERABLE you must enter 3.0.3. The CTMT Sump monitor and RU-1 must be operable for the LCO to be met.
D. Correct - TS 3.4.16 requires this action to be taken once every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
Rev 0 Question 18 Table-Item Links Q - 10CFR Sections 55.43 (2)Facility operating limitations in the technical specifications and their bases.
Q - Senior Reactor Operator K & A Importance classification 4.7 Q - Cognitive Level Memory Q - Question Source New Q - SRO Exam SRO Exam 2010 O - SRO Tier Group Designation Tier 2 Group 1 Associated objective(s):
apply Tech Spec Section 3.4.16 (RCS Leak Detection Instrumentation)
Rev 0 19 ID: Q22659 Points: 1.00 Given the following conditions:
The Site Control Rooms are presently staffed by the following personnel.
UNIT 1 UNIT 2 UNIT 3 1 Shift Manager 1 Shift Manager 1 Shift Manager 1 STA (Unlicensed) 1 STA (Unlicensed) 1 STA (Unlicensed) 1 CRS 1 CRS 1 CRS 2 ROs 2 ROs 3 ROs The Fire Team Advisor is being supplied by the Unit 3 third RO.
One of the Unit 1 ROs becomes ill and has to leave the site immediately.
It is three hours until the end of the shift.
Which ONE of the following describes the appropriate response?
The Operating Crew...
A. can continue to operate with the reduced manning for the remainder of the shift per ODP-2 (Operations Shift Coverage).
B. must commence a plant shutdown to be in MODE 3 within 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> per Technical Specification 5.2.2 (Unit Organization Staff).
C. can continue to operate with the reduced manning, provided that the FTA from Unit 3 assumes the 2nd RO in Unit 1 per Technical Specification 5.3.1 (Unit Staff Qualifications).
D. can continue to operate with the reduced manning for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, provided immediate action is taken to restore crew manning to the minimum requirements per 40DP-9OP02 (Conduct of Shift Operations).
Answer: D
Rev 0 Question 19 Info Question Multiple Choice Type:
Status: Active Always select on No test?
Authorized for No practice?
Points: 1.00 Time to 3
Complete:
Difficulty: 3.00 System ID: 22659 User-Q22659 Defined ID:
Cross Reference Number:
Topic: Q22659 SRO GEN STAFF MANNING Num Field 2.9 1:
Num Field 3.9 2:
Text Field: 2.1.5 Comments: Proposed reference to be provided to applicant during examination:
NONE Technical
Reference:
40DP-9OP02, Conduct of Shift Operations K&A:
Ability to use procedures related to shift staffing, such as minimum crew complement, overtime limitations, etc.
Justification:
A. Incorrect - The unit must take action to remedy the situation immediately with manning restored within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
B. Incorrect - The TS allows for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to meet the 10 CFR 50.54 (m) requirements as long as action is taken to restore shift manning.
C. Incorrect - While this would make the U1 meet the min manning requirements, the site would not meet the requirement to have a FTA on site for all 3 units.
D. Correct - This is correct and will meet all the requirements of site specific procedures and TS.
Rev 0 Question 19 Table-Item Links Q - 10CFR Sections 55.43 (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
Q - Senior Reactor Operator K & A Importance classification 3.9 Q - Cognitive Level Comprehension / Anal Q - Question Source New Q - SRO Exam SRO Exam 2010 O - SRO Tier Group Designation Tier 3 Associated objective(s):
Describe the required actions if the shift manning is less than the minimum required manning.
Rev 1 20 ID: Q43785 Points: 1.00 Which ONE of the following describes when a plant-wide announcement is required to be made?
A. Changing from Mode 3 to Mode 2.
B. Energizing PNA-D25 after a permit has been cleared.
C. Starting HCN-A01C (CTMT Normal ACU Fan) from the Control Room.
D. AFB-P01 (Essential Motor Driven Aux Feed Pump) started automatically on AFAS-1.
Answer: A
Rev 1 Question 20 Info Question Multiple Choice Type:
Status: Active Always select No on test?
Authorized No for practice?
Points: 1.00 Time to 2
Complete:
Difficulty: 2.00 System ID: 43785 User-Defined Q43785 ID:
Cross Reference Number:
Topic: Q43785 ODP-1 Plant Announcements Num Field 1: 3.1 Num Field 2: 3.1 Text Field: 2.1.14 Comments: Proposed reference to be provided to applicant during examination: NONE Technical
Reference:
ODP-1, Operations Department Principles and Standards K&A:
Knowledge of criteria or conditions that require plant-wide announcements, such as pump starts, reactor trips, mode changes etc.
Justification:
A. Correct - Plant-wide announcements shall be made when changing modes.
B. Incorrect - 120 Vac distribution panels are not required to be announced.
C. Incorrect - 480 Vac motor starts are not required to be announced.
D. Incorrect - Equipment that starts automatically is not required to be announced.
Rev 1 Question 20 Table-Item Links Q - 10CFR Sections 55.43 (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
Q - Senior Reactor Operator K & A Importance classification 3.1 Q - Cognitive Level Memory Q - Question Source New Q - SRO Exam SRO Exam 2010 O - SRO Tier Group Designation Tier 3 Associated objective(s):
ODP-1 Reactivity Management
Rev 2 21 ID: Q10303 Points: 1.00 Given the following conditions:
You are the Unit 1 CRS during a weekend nightshift.
The Effluent Tech has requested a procedure change to 74RM-9EF41, Radiation Monitoring System Alarm Response.
The change affects the Radiological Monitoring Group Response to an Alert/High alarm on RU-2/3 (EW 'A' and EW 'B').
You meet the requirements for a TAPA.
Which ONE of the following describes the TAPA process?
As a minimum this change requires a Temporary Approved Procedure Action (TAPA) be generated with ...
A. an approval by two SROs with valid licenses on the affected Unit.
B. an approval by two SROs, one of whom holds a valid SRO license on the affected Unit.
C. a completed 50.59 review and approval by two members of the plant supervisory staff each within the affected discipline.
D. a completed 50.59 review and approval by two members of the plant supervisory staff one of whom holds a valid SRO license on the affected Unit.
Answer: D
Rev 2 Question 21 Info Question Multiple Choice Type:
Status: Active Always select on No test?
Authorized for No practice?
Points: 1.00 Time to 3
Complete:
Difficulty: 4.00 System ID: 10303 User-Q10303 Defined ID:
Cross Reference 01DP-0AP01 Number:
Topic: Q10303 Describe the process for generating a TAPA.
Num Field 3.0 1:
Num Field 3.6 2:
Text Field: 2.2.6 Comments: Proposed reference to be provided to applicant during examination:
None Technical
Reference:
01DP-0AP01, Procedure Process K&A: Knowledge of the process for making changes to procedures.
Justification:
A. Incorrect - SRO does not mean plant supervisory staff, also it does not address the specific discipline. 50.59 reviews are required per 93DP-0LC07.
B. Incorrect - SRO does not mean plant supervisory staff, also does not address the specific discipline. 50.59 reviews are required per 93DP-0LC07.
C. Incorrect - requires 50.59 reviews and needs at least one SRO as supervisory staff.
D. Correct - meets requirements of 01DP-0AP01 section 8.3.5 - 6 and 50.59 reviews are required per 93DP-0LC07.
Rev 2 Question 21 Table-Item Links Q - 10CFR Sections Facility licensee procedures required to obtain authority for design and operating changes in the facility.
55.43 (3) Facility licensee procedures required to obtain authority for design and operating changes in the facility Q - Senior Reactor Operator K & A Importance classification 3.6 Q - Cognitive Level Memory Q - Question Source PV Bank Not Modified Q - SRO Exam SRO Exam 2007 SRO Exam 2010 O - SRO Tier Group Designation Tier 3 Associated objective(s):
Describe the process for generating a TAPA.
Rev 1 22 ID: Q5210 Points: 1.00 Given the following conditions:
Gaseous radwaste monitor RU-12 IS NOT in service.
A waste gas decay tank release is required.
Which ONE of the following must be performed by Operations prior to the release?
A. RU-12 must be restored.
B. The Shift Manager must invoke 50.54(x).
C. Two independent verifications must be performed of the discharge valve lineup.
D. Alternate sampling must be in place per 74RM-9EF60, RMS Sample Collection.
Answer: C
Rev 1 Question 22 Info Question Multiple Choice Type:
Status: Active Always select on No test?
Authorized for No practice?
Points: 1.00 Time to 2
Complete:
Difficulty: 3.00 System ID: 5210 User-Q5210 Defined ID:
Cross Reference Number:
Q5210 GEN SRO ADMIN Effl - gas decay tank release is Topic:
required Num Field 3.8 1:
Num Field 4.3 2:
Text Field: 2.3.11 Comments: Proposed reference to be provided to applicant during examination: NONE Technical
Reference:
74DP - 9CY08, The Radiological Monitoring Program K&A:
Ability to control radiation releases.
Justification:
A. Incorrect - No requirement to restore the RU monitor.
B. Incorrect - 50.54(x) is only used when the ODCM will be exceeded.
C. Incorrect - This only applies to RU141/143/145.
D. Correct - Directed per 74DP-9CY08 The Radiological Monitoring Program.
Rev 1 Question 22 Table-Item Links Q - 10CFR Sections 55.43 (4) Radiation hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions.
Q - Senior Reactor Operator K & A Importance classification 4.3 Q - Cognitive Level Memory Q - Question Source PV Bank Not Modified Q - SRO Exam SRO Exam 2010 O - SRO Tier Group Designation Tier 3 Associated objective(s):
Define Operations responsibilities
Rev 0 23 ID: Q43815 Points: 1.00 Which ONE of the following describes the EPA Guidance for Life Saving and which position must approve a PVNGS worker receiving Potassium Iodide (KI)?
____(1)____ TEDE is the EPA Guidance for Life Saving (Non Voluntary Basis) and the
____(2)____ must approve the PVNGS worker receiving Potassium Iodide (KI).
A. (1) 10 Rem per event (2) Emergency Coordinator.
B. (1) 25 Rem per event (2) Emergency Coordinator.
C. (1) 10 Rem per event (2) Radiation Protection Monitor.
D. (1) 25 Rem per event (2) Radiation Protection Monitor.
Answer: B Question 23 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 5 Difficulty: 0.00 System ID: 43815 User-Defined ID: Q43815 Cross Reference Number:
Topic: Q43815 SRO GEN EP Authorize Exposre Extension Num Field 1: 3.2 Num Field 2: 3.7 Text Field: 2.3.4 Comments: Proposed reference to be provided to applicant during examination: NONE Technical
Reference:
EPIP-01 STSC Actions Sect 10.12 K&A:
Knowledge of radiation exposure limits under normal or emergency conditions.
Justification:
Incorrect - 10 Rem is the limit for protecting valuable property.
Incorrect - Per step 10.12 of EPIP-01 and EPIP-99 App K.
Correct - 10 Rem is the limit for protecting valuable property. RPM is used to consult on such matters, but does not approve the dose.
Incorrect - Limit is correct. RPM is used to consult on such matters, but does not approve the dose.
Rev 0 Question 23 Table-Item Links Q - 10CFR Sections 55.43 (4) Radiation hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions.
Q - Senior Reactor Operator K & A Importance classification 3.7 Q - Cognitive Level Memory Q - Question Source New Q - SRO Exam SRO Exam 2010 O - SRO Tier Group Designation Tier 3 Associated objective(s):
Identify the Emergency Coordinator's responsibilities associated with Emergency Exposure.
Rev 2 24 ID: Q22667 Points: 1.00 Given the following conditions:
Initial Conditions:
Unit 1 is operating at 100% power.
A SGTL has been identified and 40AO-9ZZ02 (Excessive RCS Leakage) is in progress.
RU-4 (SG #1 Blowdown) is in Alert alarm and trending up.
Subsequently:
The Reactor trips due to an ESD.
Which ONE of the following describes the appropriate EOP mitigation strategy?
Perform SPTAs and enter...
A. 40EP-9EO09 (FRP).
B. 40EP-9EO05 (ESD).
C. 40EP-9EO04 (SGTR).
D. 40EP-9EO05 (ESD) and 40AO-9ZZ02 (Excessive RCS Leakage).
Answer: A
Rev 2 Question 24 Info Question Multiple Choice Type:
Status: Active Always select on No test?
Authorized No for practice?
Points: 1.00 Time to 3
Complete:
Difficulty: 2.00 System ID: 22667 User-Defined Q22667 ID:
Cross Reference Number:
Topic: Q22667 SRO GEN SGTR ESD EOP mitigation strategy Num Field 1: 3.7 Num Field 2: 4.7 Text Field: 2.4.6 Comments: Proposed reference to be provided to applicant during examination:
NONE Technical
Reference:
40EP-9EO09, FRP K&A:
Knowledge of EOP mitigation strategies.
Justification:
A. Correct - FRP will address the dual event. HR-2 Step 14 directs mitigating strategy of this instance that you have a tube rupture and are uncontrollably steaming to atmosphere.
B. Incorrect - ESD is the major trip initiating event here, CRS may determine that the ESD EOP will mitigate the event.
C. Incorrect: Due to the Dual Event must go to FRP, Note in App F of Excessive RCS Leakage states post trip enter SGTR, CRS may choose this based on that note.
D. Incorrect - Due to the Dual Event must go to FRP, Note in App F of Excessive RCS Leakage states post trip enter SGTR, CRS may not know this note and use ESD and AOP to mitigate event.
Rev 2 Question 24 Table-Item Links Q - 10CFR Sections 55.43 (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
Q - Senior Reactor Operator K & A Importance classification 4.7 Q - Cognitive Level Memory Q - Question Source PV Bank Not Modified Q - SRO Exam SRO Exam 2010 O - SRO Tier Group Designation Tier 3 Associated objective(s):
determine whether or not entry into the FRP is appropriate
Rev 1 25 ID: Q43787 Points: 1.00 Given the following conditions:
Unit 3 is operating at 100% power.
An Abnormal Operating Procedure (AOP) is in progress due to equipment failure.
Conditions requiring a reactor trip are approaching.
The CRS will continue to implement the AOP after the reactor has been tripped.
The RO manually trips the reactor from B05.
Which ONE of the following statements below describe the rules the CRS must follow to continue implementation of the AOP?
The CRS CANNOT continue implementing the AOP until after A. addressing all safety functions in the SPTAs.
B. all alarm response procedures have been addressed.
C. addressing the reactivity control safety function in SPTAs.
D. diagnosing the event and transitioning to the Optimal Recovery or Functional Recovery Procedures.
Answer: C
Rev 1 Question 25 Info Question Multiple Choice Type:
Status: Active Always select No on test?
Authorized for No practice?
Points: 1.00 Time to 3
Complete:
Difficulty: 3.00 System ID: 43787 User-Defined Q43787 ID:
Cross Reference Number:
Topic: Q43787 AOP in EOP USE Num Field 1: 3.8 Num Field 2: 4.5 Text Field: 2.4.8 Comments: Proposed reference to be provided to applicant during examination: NONE Technical
Reference:
40DP-9AP18, AOP Users Guide.
K&A:
Knowledge of how abnormal operating procedures are used in conjunction with EOPs.
Justification:
A. Incorrect - Some AOPs must be completed concurrently such as the "PERFORM" direction.
B. Incorrect - No actions are permitted until the Reactivity Safety Function is complete. ARPs must be addressed once plant conditions have stabilized.
C. Correct - This is correct per 40DP-9AP18 step 17 bullet 1. AOP Users Guide.
D. Incorrect - Some AOPs must be exited such as the "Go To" direction.
Rev 1 Question 25 Table-Item Links Q - 10CFR Sections 55.43 (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
Q - Senior Reactor Operator K & A Importance classification 4.5 Q - Cognitive Level Memory Q - Question Source New Q - SRO Exam SRO Exam 2010 O - SRO Tier Group Designation Tier 3 Associated objective(s):
define the required actions for the conditions given