ML050830162
| ML050830162 | |
| Person / Time | |
|---|---|
| Site: | Peach Bottom  |
| Issue date: | 02/18/2005 |
| From: | Popielarski J Exelon Generation Co |
| To: | Conte R NRC/RGN-I/DRS/OSB |
| Conte R | |
| References | |
| Download: ML050830162 (200) | |
Text
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Peach Bottom February 2005
- 1.
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Initial License Operator NRC Examination Unit 3 was operating at full power when the following transient occurred:
0 BOTH Recirculation Pumps tripped 0 The reactor continued to operate Specific procedural actions are required for this condition in accordance with OT-112, UnexpectedKJnexplained Change in Core Flow.
The actions of OT-1 12 are based on preventing the plant fiom exceeding which one of the following limits.
A. LHGR B. MCPR C. FLLLP D. APLHGR
NRC Question Data Sheet Level of Knowledge Difficulty Time Allowance (minutes)
LOW 2.25 3
Answer Kev SRO N/A Question ID# 001 Both RO/SRO Knowledge/Ability K/A Choice Correct:
295001 AK1.02 Importance:
RO / SRO Partial or Complete Loss of Forced Core Flow Circulation 3.5 / 3.6 Distractors:
B A
C D
Basis or Justification The power oscillations that are a result of this high power to low flow condition could result in exceeding the MCPR limit.
LHGR is primarily a concern during continuous high power conditions that could overpower a specific bundle of fuel.
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FLLLP is not the specific thermal limit of concern, but makes an excellent distractor due to its relationship with flow.
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APLHGR is primarily a concern following a Loss of Coolant Accident.
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Source:
Reference@):
Learning Objective:
Source Documentation New Exam Item 0 Modified Bank Item 0 ILT Exam Bank 0 Previous NRC Exam 0 Other Exam Bank OT-112 and Bases PBIG PLOT-1540.03 REQUIRED MATERIALS:
Notes and Comments:
Prepared By:
Philip E. Nielsen (717) 4564122 PBAPS Regulatory Exam Author Page 2 of 150
Peach Bottom February 2005 Initial License Operator NRC Examination
- 2.
The following conditions existed at Peach Bottom Atomic Power Station:
0 0
0 Unit 2 was operating at full power.
Both units had electrical loads aligned in the preferred line-up.
Unit 3 had been Shutdown due to a small steam leak resulting in 2.3 psig in the Drywell.
A trip of the SU-25 Breaker (452-02) occurred resulting in a partial loss of electrical power. Thirty seconds later, the PRO reviewed the 4KV bus and annunciator status with the following results:
0 E-13, E-22, E-33, E-42 were powered by their normal Emergency Auxiliary Bus 0 E-12, E-32, and E-43 had transferred to an alternate Emergency supply.
0 E-23 was deenergized with E23 BUS UNDERVOLTAGE (002 D-4) lit 0
All four Diesel Generators were running normally.
Diagnose the above conditions to determine the cause of the undervoltage condition on the E23 bus.
A. ONLY the E-223 breaker failed to close on the auto transfer.
B. ONLY the E-323 breaker failed to close on the auto transfer.
C. BOTH the E-223 breaker and then the E-23 Diesel Generator Output Breaker failed to close on the auto transfer.
D. BOTH the E-323 breaker and then the E-23 Diesel Generator Output Breaker failed to close on the auto transfer.
NRC Question Data Sheet Correct:
Answer Key D
Question ID# 002 Both ROISRO Level of Knowledge Difficulty Time Allowance (minutes)
HIGH 3.5 4
SRO NIA Distractors: 4 295003 AA2.01 Partial or Complete Loss of AC I
C Importance:
RO I SRO 3.4 13.7 Basis or Justification Correct - The candidate must recognize that SU-25 powers the 2 SUE transformer, which is the normal supply of power to E-23 through E-223.
With power lost to the 2 SUE transformer, the alternate supply breaker (E-323) should have closed in 0.25 seconds and with the diesel available, the diesel output breaker should have closed after 0.5 seconds if the bus was still deenerg ized.
The candidate must recognize that E-223 is the normal power supply to this bus and it has lost power due to the trip of the SU-25 breaker.
Although E-323 should have closed, the candidate must recognize that the Diesel Generator output breaker (E-23) has also failed to close.
Although the Diesel Generator Output breaker (E-23) has failed to close, the candidate must recognize that the alternate start up source breaker (E-323) has also failed to close.
Source: c Reference@):
Objective:
Source Documentation New Exam Item Modified Bank Item ILT Exam Bank 0
Previous NRC Exam Other Exam Bank E-8, Standby Diesel Generator & 41 60 Volt Emergency Power System PLOT-5054.6b (Description of K&A, from catalog)
Ability to determine and/or interpret the following as they apply to Partial or Complete Loss of AC:
Cause of partial or complete loss of AC power REQUIRED MATERIALS:
Notes and Comments Prepared By:
Philip E. Nielsen (717) 456-4122 PBAPS Regulatory Exam Author Page 4 of 150
Peach Bottom February 2005 Initial License Operator NRC Examination
- 3.
Both units were operating at full power when the following transient occurred:
0 A large number of annunciators including 2A DC POWER PANEL LO VOLTAGE (209 C-3) were received.
The PRO reported that the breaker indication lights for the E-12 and E-13 busses were NOT lit.
0 The STA determined that the alarms were due to the loss of voltage to 125 VDC Power Distribution Panel 20D2 1.
SE-13, Loss of a 125 or 250 VDC Safety Related Bus, Attachment 2, Part 1 directs that the E-12 and E-13 busses be taken out of service.
The reason for shedding loads and removing the E-12 and E-13 busses fiom service is to:
A. limit the possibility of emergency bus damage.
B. minimize the voltage drain on the associated DC batteries.
C. maintain personnel safety fiom the potential system grounds.
D. prevent inadvertent breaker operation while reenergizing 20D2 1.
NRC Question Data Sheet A
B Answer Key The E-I 2 and E-I 3 busses are required to be removed from service because power has been lost to their breaker control and protection circuits which would prevent them from automatically tripping as expected during a bus problem. This could result in bus damage.
The candidate must recognize with voltage lost to 20D21, not just degraded as in a ground situation, the batteries are not being drained.
Question ID# 003 RO/SRO I
D Choice The candidate must recognize that the breakers would not be expected to cycle on restoration of control and protection circuitry power.
I Basis or Justification Level of Knowledge D ifficu I ty Time Allowance (minutes)
LOW 3.0 3
Correct:
SRO N/A Distractors:
295004 AK3.01 Partial or Total Loss of DC Power Importance:
RO / SRO 2.6 13.1 The candidate must recognize that removing the emergency busses from service is not based on personnel safety.
Source:
Reference@):
Learning Objective:
Knowledge/Ability WA (Description of K&l Source Documentation New Exam Item 0
Modified Bank Item 0
ILT Exam Bank SE-13 Attachment 2 0 Previous NRC Exam 0
Other Exam Bank PLOT-1 555.9 Knowledge of the reasons for the following responses as they apply to a Partial or Total Loss of DC Power:
Load Shedding REQUIRED MATERIALS:
Notes and Comments Prepared By:
Philip E. Nielsen (71 7) 4 5 6 4 22 PBAPS Regulatory Exam Author Page 6 of 150
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Peach Bottom February 2005 Initial License Operator NRC Examination
- 4.
Unit 3 was operating at full power when the following transient occurred:
0 A loss of vacuum transient occurred with vacuum steadily getting worse.
0 The Reactor was scrammed at 25 inches of vacuum.
0 An ATWS condition occurred.
0 Attempts to restore instrument nitrogen have NOT yet been successful.
0 The Main Turbine tripped on low vacuum as it continued to drop rapidly until it was stabilized at 4 inches of vacuum.
Evaluate the above conditions and determine which of the following methods of manual reactor pressure control is currently AVAILABLE and REQUIRED to be used to control pressure.
In accordance with T-10 1, RPV Control, Reactor Pressure will be controlled MANUALLY using the:
B. Turbine Bypass Jack.
C. Reactor Feedwater Pump Turbines.
D. High Pressure Coolant Injection system.
NRC Question Data Sheet Level of Knowledge Difficulty Time Allowance (minutes)
HIGH 2.75 3
I Answer Kev SRO N/A
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Question ID# 004 Both ROISRO Main Turbine Generator Trip Correct:
3.6 13.7 r
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D I
Basis or Justification HPCl is available and required to be used under these conditions to control pressure.
SRVs are not to be used in the manual mode to control pressure under these conditions because their only source of nitrogen is the accumulators.
B C
The Turbine Bypass Jack would be directed by procedure, however it is not able to open the Bypass Valves due to the low condenser vacuum (less than the 7 vacuum) Bypass Valve Lockout.
The Reactor Feed Pump Turbines would be directed by procedure, however, they are not available due to the low condenser vacuum condition requiring the MSlVs to be closed.
Source:
Reference(s):
Learning 0 bjective:
Knowledge/Ability K/A Source Documentation New Exam Item Modified Bank Item 0
ILT Exam Bank 0 Previous NRC Exam 0 Other Exam Bank OT-106, Low Condenser Vacuum, Section 4.0 T-101, RPV Control, RC/P-13 PBI G PLOT-1540.05 295005 AK2.07 I Importance:
RO / SRO (Description of K&A, from catalog)
Knowledge of the interrelations between a Main Turbine Generator Trip and the following:
Reactor Pressure Control REQUIRED MATERIALS:
Notes and Comments Prepared By:
Philip E. Nielsen (71 7) 456-41 22 PBAPS Regulatory Exam Author Page 8 of 150
Peach Bottom February 2005 Initial License Operator NRC Examination
- 5.
Unit 2 was operating at full power when the "G" Safety Relief Valve (SRV) stuck open. The SRV cannot be reclosed.
In accordance with OT-114 Stuck Open Safety Relieve Valve, what is the MINIMUM Torus Temperature where the operators are REQUIRED to perform a manual scram using GP-4 and enter T-100 Scram?
A. 95°F B. 105°F C. 110°F D. 120°F
NRC Question Data Sheet Choice Answer Key Basis or Justification Correct:
B OT-110 requires the operator to perform a GP-4 Shutdown and enter T-100 when temperature cannot be maintained below 105°F.
Distractors:
A At 95OF, the operators are required to enter T-102 Primary Containment Control, but are not required to Scram.
C At 1 10°F, the operators are required to immediately place the Mode Switch in Shutdown rather than performing a GP-4 as indicated by the question.
D At 120°F, the reactor is required by Technical Specifications to be depressurized, but a scram should have already been initiated and a GP-4 Shutdown would be inappropriate.
REQUIRED MATERIALS:
Level of Knowledge Difficulty Time Allowance (minutes)
LOW 2.0 3
Notes and Comments SRO N/A Prepared By:
Philip E. Nielsen (717) 456-4122 PBAPS Regulatory Exam Author Source:
Reference@):
Learning 0 bjective:
Page 10 of 150 0
New Exam Item 0
Modified Bank Item ILT Exam Bank 0 Previous NRC Exam 0 Other Exam Bank OT-114 Inadvertent Opening of a Relief Valve T-1 00 Scram PBIGPLOT-1540.04 PLOT-1560.01
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Knowledge/Ability WA
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295006 G2.4.4 Importance:
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Peach Bottom February 2005 Initial License Operator NRC Examination
- 6.
The Peach Bottom Control Room has been abandoned in accordance with SE-IO, the Alternative Shutdown Procedure. The following conditions exist on Unit 2:
The Reactor Operator is controlling level using HPCI at the Unit 2 Alternative Shutdown Panel 0
Indicated Reactor Level (LI-2-2-3-112) is 10 inches 0
Reactor Pressure is 800 psig Interpret Figure 1 of SE-10, Attachment 9 (provided) to determine the actual reactor level and determine the HPCI response to an actual high level condition.
Actual level is:
A. greater than 40, HPCI will AUTOMATICALLY trip on a high level condition.
B. greater than 40, HPCI must be MANUALLY tripped on a high level condition.
C. between 0 and 40, HPCI will AUTOMATICALLY trip on a high level condition.
D. between 0 and 40, HPCI must be MANUALLY tripped on a high level condition.
NRC Question Data Sheet Level of Knowledge Difficulty Time Allowance (minutes)
HIGH 2.5 4
Answer Key SRO NIA I Question ID# 006 Both RC Choice Source:
Reference(s):
Objective:
Learning I
Correct:
New Exam Item 0
Modified Bank Item 0 ILT Exam Bank 0
Previous NRC Exam 0
Other Exam Bank SE-10, Attachment 9, Section 4.0 Cautions and Figure 1 PLOT-1555.12 Distractors:
Knowledge/Ability KIA D
29501 6 AA2.02 Importance:
RO / SRO Control Room Abandonment 4.2 14.3 A
B C
SRO I
Basis or Justification Plotting 1 0 and 800 psig on Figure 1 indicates level is between 0" and 40" and according to the procedure cautions all HPCl trips are bypassed.
Plotting 10" and 800 psig on Figure 1 shows that level is NOT greater than 40" and according to the procedure cautions, all HPCl trips are bypassed.
Plotting 10" and 800 psig on Figure 1 shows that level is NOT greater than 4 0 and according to the procedure cautions all HPCl trips are bypassed.
Plotting 10" and 800 psig on Figure 1 indicates level is between 0" and 40",
however according to procedure cautions, all HPCl trips are bypassed.
I REQUIRED MATERIALS:
SE-10 Attachment 9, Figure 1 Notes and Comments Prepared By:
Philip E. Nielsen (71 7) 4 5 6 4 22 PBAPS Regulatory Exam Author Page 12 of 150
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Peach Bottom February 2005 Initial License Operator NRC Examination
- 7.
Unit 3 was operating at full power when it experienced the following transient:
A RBCCW pump tripped due to an overcurrent condition.
B RBCCW pump started automatically but at a reduced discharge pressure.
RBCCW system temperatures are rising steadily.
Under these conditions, ON-1 13 Loss of RBCCW directs the RWCU pumps to be tripped and the system to be isolated.
According to the ON-1 13 Bases, the reason for these steps are to:
A. Isolate a likely primary-to-secondary leak in the RBCCW heat exchangers.
B. Allow more time to diagnose and correct the cause of the RBCCW problem.
C. Prevent RWCU pump cavitation due to high reactor water inlet temperatures.
D. Reduce the required RBCCW system flow rate thereby preventing pump runout.
NRC Question Data Sheet Level of Knowledge Difficulty Time Allowance (minutes)
LOW 2.25 3
Answer Key Question ID# 007 Both RO/SRO SRO N/A Choice Correct:
Source:
Reference(s):
0 bjective:
Learning Distractors:
New Exam Item 0
Modified Bank Item 0
ILT Exam Bank 0 Previous NRC Exam 0 Other Exam Bank OT-113 Loss of RBCCW Bases, Step 2.2 PLOT-1550.03 B
Knowledge/Ability K/A A
29501 8 AK3.03 Importance:
RO I SRO Partial or Total Loss of CCW 3.9 13.8 C
D Basis or Justification I
Removing RWCU from service removes a significant heat load which will greatly slow the heatup of the RBCCW system thereby providing more time to correct the problem.
A primary-to-secondary leak in the RBCCW heat exchanger is plausible, but is not what the isolation of the system is based on according to the procedure.
Although the RWCU pump inlet temperatures could rise, these steps are not based on preventing this condition.
Although isolating RWCU will reduce the required heat input to the system, it does not reduce the flow rate in the system because the isolation occurs on the Reactor Water side of the system.
REQUIRED MATERIALS:
Notes and Comments Prepared By:
Philip E. Nielsen (717) 456-4122 PBAPS Regulatory Exam Author Page 14 of 150
Peach Bottom February 2005
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Initial License Operator NRC Examination
- 8.
Unit 2 was operating at full power with all Instrument Air and Instrument Nitrogen systems aligned normally when it experienced the following:
0 0
Annunciator NITROGEN COMPRESSOR A OR B TROUBLE (228 E-2) alarms After investigation, the EO reports:
+ The A and B Instrument Nitrogen Compressors are tripped.
+ The A and B Instrument Nitrogen Receiver pressures are at 80 psig.
With no operator action, under these conditions, pressure will AUTOMATICALLY be maintained to the A and B Instrument Nitrogen Headers by the:
A. Nitrogen Bottles aligned by the auto opening of SV-8 130 AB, AB Supply.
B. Containment Atmosphere Dilution System aligned by the auto opening of PCV-765 1 A/B, SGIG Pressure Control Valve.
C. Truck Connection aligned by the auto opening of PCV-8917 A/B, AB Nitrogen Pressure Control Valve for Backup Supply.
D. Instrument Air System aligned by the auto opening of AO-4230 AB, AB Instrument Air Backup to Instrument Nitrogen.
NRC Question Data Sheet D
A B
Answer Key Basis or Justification Instrument air will automatically backup the Instrument Nitrogen System when Instrument Nitrogen Receiver pressure drops below 85 psig.
Although the SV-8130 valves have an open/auto position, the valves are normally in the closed position. If left in open/auto, the valves would be open unless they were isolated. Pressure would only be aligned to the ADS Valves.
Alignment of the CAD system through the SGlG system to supply the Instrument Nitrogen System requires manual valve alignments.
Question ID# 008 Both RO/SRO Level of Knowledge HIGH Choice Difficulty Time Allowance (minutes)
SRO 3.25 4
N/A Correct:
Source:
Reference( s):
Learning 0 bjective:
Distractors:
New Exam Item 0
Modified Bank Item 0
ILT Exam Bank 0
Previous NRC Exam
[7 Other Exam Bank ARC 228 E-2, Nitrogen Compressor A or B Trouble PLOT-1536.04a
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Knowledge/Ability 29501 9 AA1.02 WA Partial or Total Loss of Instrument Air The truck connection is available to be used, but is not aligned for automatic operation. Pressure would only be supplied to the ADS valves.
Importance: RO / SRO 3.3 / 3.1 Psychometrics REQUIRED MATERIALS:
Notes and Comments Prepared By:
Philip E. Nielsen (71 7) 456-41 22 PBAPS Regulatory Exam Author Page 16 of 150
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Peach Bottom February 2005 Initial License Operator NRC Examination
- 9.
Unit 3 was in Day 5 of a refueling outage with the following conditions present:
0 Both Recirculation Pumps were secured.
0 RWCU was in service vessel to vessel.
0 Normal Shutdown Cooling was in service using the D RHR Pump.
The unit then experienced a loss of shutdown cooling due to an inadvertent isolation signal. The cause of the signal is unknown and shutdown cooling cannot be immediately returned to service.
In accordance with ST-0-80-500-3, Recording and Monitoring Reactor Vessel Temperatures and Pressure, select the RPV Coolant Temperature indication that is the most useful in detecting possible RPV Coolant temperature stratification.
A. RPV Skin Temperature B. Vessel Drain Pipe Temperature C. Recirc Pump Suction Temperature D. RHR Heat Exchanger Inlet Temperature
NRC Question Data Sheet A
B C
D Answer Key ON-125, Loss or Unavailability of Shutdown Cooling directs the use of ST-0-80-500-3, Recording and Monitoring Reactor Vessel Temperatures and Pressure. A NOTE in the ST (after step 6.1.l
- 1) states that while in natural circulation, RPV skin thermocouples are the best indication of RPV coolant temperature and may provide an indication of RPV coolant temperature stratification.
The candidate must recognize that although RWCU is in service, this temperature is not an accurate indication of RPV coolant temperature unless forced circulation of the RPV is present.
The candidate must recognize that with the Recirculation Pumps out of service, this indication is not the best indication of RPV coolant temperature.
The candidate must recognize that with the Shutdown Cooling Pumps out of service, this indication is not the best indication of RPV coolant temperature.
Question ID# 009 Both RO/SRO Level of Knowledge Difficulty Time Allowance (minutes)
LOW 3.0 3
Choice Correct:
SRO N/A Distractors:
295021 AA2.04 Loss of Shutdown Cooling I
Basis or Justification Importance:
Reference(s):
Learning Objective:
Knowledge/Ability WA Source Documentation New Exam Item 0
Modified Bank Item 0
ILT Exam Bank 0
Previous NRC Exam 0 Other Exam Bank ON-125, Loss or Unavailability of Shutdown Cooling ST-0-80-500-3, Recording and Monitoring of Reactor Vessel Temperatures and Pressure GP-12, Core Cooling PB IG PLOT-1 550.28a (Description of K&A, from catalog)
Ability to determine and/or interpret the following as they apply to a Loss of Shutdown Cooling:
Reactor water temperature EQUIRED MATERIALS:
Notes and Comments Prepared By:
Philip E. Nielsen (717) 456-4122 PBAPS Regulatory Exam Author Page 18 of 150
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Peach Bottom February 2005 Initial License Operator NRC Examination
- 10.
Refueling movements are in progress on Unit 2.
Evaluate the following unanticipated conditions to determine which one is symptomatic of a refueling event requiring action in accordance with ON-124, Fuel Floor and Fuel Handling Problems.
A. Refuel Slot 234 Elevation Area Radiation Monitor alarms.
B. Refueling Floor Vent Exhaust Hi Radiation (2 18 A-1) alarms.
C. An irradiated LPRM Detector is dropped in the ISFSI Cask Handling Area.
D. Count Rate doubles as the fifth (Sth) fuel assembly is loaded near a WRNM.
NRC Question Data Sheet A
Answer Kev
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ON-124 requires entry and action for any Fuel Floor ARM alarm.
Question ID# 010 Both ROiSRO Level of Knowledge LOW Choice I
Basis or Justification Difficulty Time Allowance (minutes)
SRO 3.25 3
N/A Correct:
295023 AK2.03 Refueling Accident Importance:
RO / SRO 3.4 i 3.6 Distractors:
Although this condition obviously requires action, it is not an entry condition into ON-I 24.
ON-124 entry is required for a fuel assembly or single fuel rod dropped or damaged, but not for an LPRM detector.
ON-124 entry would only be required if the count rate had doubled twice I D l between CCTAS steps.
Source:
Reference(s):
Learning Objective:
Knowledge/Ability WA Source Documentation New Exam Item 0
Modified Bank Item 0
ILT Exam Bank 0 Previous NRC Exam Other Exam Bank ON-I 24, Fuel Floor and Fuel Handling Problems PBIGPLOT-1550.2 (Description of K U, from catalog)
Knowledge of the interrelations between a Refueling Accident and the following:
Radiation Monitoring Equipment EQUIRED MATERIALS:
Notes and Comments Prepared By:
Philip E. Nielsen (717) 456-4122 PBAPS Regulatory Exam Author Page 20 of 150
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Peach Bottom February 2005 Initial License Operator NRC Examination 1 1.
Unit 2 was operating at full power when the following transient occurred:
A small break Loss of Coolant Accident (LOCA) occurred.
The unit was shutdown when it was determined that Drywell Pressure could NOT be maintained less than 1.2 psig.
0 The Ctmt. Spray Override 2/3 Core Coverage Keylock Switch (S18B) was placed in Manual Overrd and Drywell sprays were initiated using the B Loop of RHR as directed by T-102, Primary Containment Control.
Drywell pressure is currently 6 psig and dropping steadily.
With no additional operator action, how will the.Drywel1 Sprays respond to the continued Drywell Pressure drop due to the Drywell Spray?
Containment Spray System logic will cause Drywell Sprays to:
A. automatically isolate just as Drywell Pressure drops below 1 psig.
B. automatically isolate just as Drywell Pressure drops below 2 psig.
C. remain in service, since once initiated, they will only secure manually.
D. remain in service, regardless of Drywell Pressure, since S 18B is in Manual Overrd.
NRC Question Data Sheet Level of Knowledge HIGH Answer Key Difficulty Time Allowance (minutes)
SRO 3.0 4
NIA Question ID# 01 1 Both RC Choice Source:
Reference(s):
Objective:
Learning Correct:
New Exam Item 0
Modified Bank Item 0 ILT Exam Bank 0 Previous NRC Exam 0
Other Exam Bank T-204, Initiation of Drywell Sprays using RHR PLOT-5010.01L Distractors:
KnowledgelAbility KIA A
295024 EK2.11 Importance:
RO I SRO High Drywell Pressure 4.2 14.2 B
C D
SRO Basis or Justification The containment spray logic requires a LOCA signal, 2/13 Core Coverage, Drywell pressure >1 psig, and the S17 switch taken momentarily to manual.
Dropping 4 psig breaks this logic and will cause the valves to close.
The Ctmt. Spray 2/3 Core Coverage Override Keylock Switch bypasses the requirement for a LOCA signal (which the 2 psig signal contributes to).
Dropping below 2 psig will not result in isolating the Drywell Spray alianment.
The Drywell Spray Initiation Curve is for initiation purposes only, if parameters drop outside of the curve after initiation, drywell sprays are left in service.
The Ctmt. Spray 2/3 Core Coverage Override Switch bypasses the LOCA signal and 2/3 Core Coverage, but it does not bypass the required 1 psig Drywell pressure input.
REQUIRED MATERIALS:
Notes and Comments Prepared By:
Philip E. Nielsen (717) 456-4122 PBAPS Regulatory Exam Author Page 22 of 150
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Peach Bottom February 2005 Initial License Operator NRC Examination
- 12.
Unit 2 was operating at full power when it experienced the following transient:
A small break Loss of Coolant Accident (LOCA) occurred.
0 The unit was shutdown when it was determined that Drywell Pressure could not be maintained less than 1.2 psig.
The Drywell Bulk Average Temperature Indication has failed.
The crew attempted to perform a manual calculation of Drywell Bulk Average Temperature using RT-0-40C-530-2, but was unsuccessful due to temperature point (Point 139) in Zone 4 being out of service.
Torus Pressure has exceeded 9 psig and the crew is attempting to determine if it is safe to spray the Drywell based on the Drywell Spray Initiation Limit (DWSIL)
Curve.
Evaluate these conditions to determine the appropriate action related to Drywell Spray based on interpretation of the available Drywell Temperature indications.
A. Do NOT spray the Drywell since the safe side of the DWSIL curve cannot be verified.
B. Spray the Drywell after verifying the safe side of the DWSIL Curve using TI-2501, Point 136.
C. Spray the Drywell after verifying the safe side of the DWSIL Curve using TI-2501, Point 136 PLUS 10'F.
D. Spray the Drywell after verifying the safe side of the DWSIL Curve using the hottest temperature indicated on TI-2501, Points 119-127.
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NRC Question Data Sheet Distractors:
Answer Key B
C I Question ID# 012 Both RO/SRO Knowledge/Ability KIA 295024 EA2.02 Importance:
RO I SRO High Drywell Pressure 3.9 14.0 Basis or Justification RT-0-40C-530-2, Temperature Monitoring, states that if the calculation of Drywell Bulk Average Temperature is invalid, the safe side of the DWSIL curve cannot be verified. DO NOT SPRAY THE DRWVELL.
TI-2501, Point 136 is indicated on the Data sheet as an alternate temperature indication for entering ON-120 or T-102, however, reading the body of the procedure indicates that 10°F must be added even to use it in this manner.
TI-2501, Point 136 PLUS 10°F is an acceptable indication to use for entry into ON-120 (High DW Temp) or T-102 (Secondary Containment Control),
but not for use on the DWSIL curve.
Using the hottest temperature point on TI-2501 Points 119-127 is an acceptable value for determining if a RPV Blowdown should be initiated, but is not acceptable for use on the DWSIL curve.
Psychometrics Level of Knowledge I Difficulty I Time Allowance (minutes) I SRO I
Source:
Reference(s):
Learning Objective:
Source Documentation New Exam Item 0 Modified Bank Item 0 ILT Exam Bank 0
Previous NRC Exam 0
Other Exam Bank T-102, Primary Containment Note #27 RT-0-40C-530-2, Temperature Monitoring PLOT-1560.11 Notes and Comments Prepared By:
Philip E. Nielsen (717) 456-4122 PBAPS Regulatory Exam Author Page 24 of 150
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Peach Bottom February 2005 Initial License Operator NRC Examination
- 13.
Unit 3 was operating at hll power when it experienced the following transient:
A Digital Feedwater Control System malfunction causes reactor level to rise rapidly.
The reactor was manually scrammed just before reaching the high-level turbine trip setpoint.
Level continued to rise to 77 on LI-3-2-3-86.
Reactor Pressure is 1070 and rising.
The Main Steam Isolation Valves (MSIVs) have been manually closed.
All URO and PRO Scram Actions were completed successhlly with the exception of those related to the Feedwater system.
The SRO has directed RPV pressure control using the Safety Relief Valves (SRVs).
Use OT-1 10, Figure 1 to determine how reactor pressure is required to be controlled with the SRVs.
RPV pressure will be controlled by:
A. prolonged operation of a single SRV repeatedly.
B. prolonged operation rotating among the ADS SRVs.
C. prolonged operation rotating among ALL of the SRVs.
D. allowing the SRVs to cycle automatically on pressure.
NRC Question Data Sheet Source:
Answer Key New Exam Item 0 Modified Bank Item 17 Previous NRC Exam 0 Other Exam Bank I Question ID# 013 Both R(
Choice Reference@):
0 bjective:
Knowledge/Ability WA Learning Correct:
OT-110, Reactor High Level including Figure 1 P BI G PLOT-1 540.04 295025 EA?.03 Importance:
RO / SRO High Reactor Pressure 4.4 / 4.4 Distractors:
A B
C D
SRO Basis or Justification Conditions are on the unsafe side of the OT-110, Figure 1 curve indicating that the steam lines may be flooded. Under these conditions, OT-110 directs a single SRV to be used to control pressure to limit the potential damage from the flooded main steam line.
This would be the normal method of pressure control if the candidate incorrectly plots the OT-1 1 0, Figure 1 curve and does not identify that the main steam lines may be flooded.
This method would seem to spread the stress among many SRVs limiting an excessive buildup of stress on any one SRV.
This action is required when only the accumulators are available to provide instrument nitrogen to the SRVs. Under these circumstances, this could have negative consequences since the higher pressures for automatic opening would place even more stress on the system piping.
I 1
I Psychometrics I
I I
Level of Knowledge I Difficulty I Time Allowance (minutes) I SRO I
I REQUIRED MATERIALS:
OT-110, Figure 1 Notes and Comments Prepared By:
Philip E. Nielsen (717) 456-4122 PBAPS Regulatory Exam Author Page 26 of 150
Peach Bottom February 2005 Initial License Operator NRC Examination
- 14.
Unit 3 was operating at fill power when the following transient occurred:
e e
A small break Loss of Coolant Accident (LOCA) occurred due to a failure in a steam line weld.
A Hydraulic ATWS occurred when a GP-4, Manual Reactor Scram was performed.
Reactor Power is 18%
Drywell Pressure is 6.8 psig Torus Pressure is 5.6 psig Torus Temperature is 190°F and steady Torus Level is 15.7 feet.
The A RHR Loop is in Torus cooling at 22000 gpm using both pumps The B RHR Loop is in Torus cooling at 10500 gpm using the D RHR Pump.
The B RHR Pump is unavailable due to tripping on overcurrent.
Evaluate these conditions using the T-102 sheet 3 curves to determine any changes that are required to the Torus Cooling flows while still maintaining MAXIMUM available Torus Cooling.
A. A Loop flow is acceptable.
B Loop flow is acceptable.
B. A Loop flow is acceptable.
B Loop flow must be reduced to 10,000 gpm.
C. A Loop flow must be reduced to 21,000 gpm.
B Loop flow is acceptable.
D. A Loop flow must be reduced to 20,000 gpm.
B Loop flow must be reduced to 10,000 gpm.
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NRC Question Data Sheet Choice Correct:
C
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Answer Key Question ID# 014 Both RO/SRO Basis or Justification This answer is correct based on using the 14.5 to 19.99 feet curve and selecting the 3 to 5.99 psig Torus Pressure Line as the limit.
D The candidate would identify this answer if they incorrectly used Drywell Pressure instead of Torus Pressure when plotting the curve.
Distractors:
The candidate would identify this answer if they selected the wrong (12.3 -
14.49 Torus Level) curve.
The candidate would identify this answer if they confused the temperature I B l and Drywell Pressure Lines.
Level of Knowledge HIGH Difficulty Time Allowance (minutes)
SRO 3.0 5
NIA Psychometrics I
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Reference@):
0 bjective:
Learning 0
Modified Bank Item 0 ILT Exam Bank 0
Other Exam Bank T-102, Primary Containment Control Sheets 1 & 3 PLOT401 0.5g Source Documentation Source:
I New Exam Item 0
Previous NRC Exam Knowledge/Ability WA 295026 EK2.01 Importance: RO / SRO Suppression Pool High Water Temperature 3.9 14.0 REQUIRED MATERIALS:
Provide T-102, Sheet 3 Notes and Comments Prepared By:
Philip E. Nielsen (717) 4564122 PBAPS Regulatory Exam Author Page 28 of 150
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Peach Bottom February 2005 Initial License Operator NRC Examination
- 15.
Unit 3 was operating at full power when the following transient occurred:
0 0
0 A small break Loss of Coolant Accident (LOCA) occurred.
The unit was shutdown when it was determined that Drywell Pressure could not be maintained less than 1.2 psig.
The following are the current plant conditions:
+ Drywell Temperature 260°F and rising slowly
+ Drywell Pressure 10 psig and rising slowly
+ Torus Level 15.2 feet and rising slowly
+ Reactor Level -60 inches and dropping
+ Reactor Pressure 825 psig with a normal depressurization in progress Use the attached Drywell Spray Initiation Limit (DWSIL) Curve (DW/T-2) to determine if it is acceptable to spray the Drywell under these conditions and any required T-102, Primary Containment Control, limitations.
A. Do NOT spray the Drywell until Temperature exceeds 28 1 OF.
B. Do NOT spray the Drywell once the Torus Level exceeds 16 feet.
C. Spray the Drywell; terminate sprays before Drywell Pressure drops below 2 psig.
D. Spray the Drywell; terminate sprays before conditions exceed the DWSIL Curve.
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NRC Question Data Sheet C
Answer Key Drywell sprays are required to be terminated if DW pressure drops below 2 psig to protect the containment from damage.
Question ID# 015 Both RO/SRO I
A Choice These conditions permit DW spray when plotted on the DWSIL Curve and, in fact, T-102 directs sprays before reaching 281 O F.
I Basis or Justification B
D Correct:
Torus Level at 16 feet is not an appropriate limitation for Drywell sprays.
T-102 specifies that sprays should be stopped before Torus level exceeds 18 feet.
The DWSIL Curve is only used to determine if it is appropriate to initiate DW sprays. This could be easily confused by the candidate because the provided curve indicates "DO NOT SPRAY" on the "UNSAFE" side.
Distractors:
Level of Knowledge Difficulty Time Allowance (minutes)
HIGH 3.25 4
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Knowledge/Ability 295028 EA1.01 WA High Drywell Temperature Importance:
Reference@):
Learning Objective:
Source Documentation New Exam Item 0
Modified Bank Item 0 ILT Exam Bank 0
Previous NRC Exam 0
Other Exam Bank T-204, Initiation of Drywell Sprays using RHR PLOT-1560.03 REQUIRED MATERIALS:
Provide T-102 Curve - Drywell Spray Initiation Limit (DWSIL) DWTT-2 Notes and Comments Prepared By:
Philip E. Nielsen (71 7) 4 5 6 4 22 PBAPS Regulatory Exam Author Page 30 of 150
Peach Bottom February 2005 Initial License Operator NRC Examination
- 16.
Unit 3 was operating at full power when it experienced a Main Turbine trip due to a loss of EHC and an ATWS. The following conditions are present:
0 Reactor Power is 15%.
0 0
Torus Temperature is 180°F.
0 Reactor Level is being lowered in accordance with T-117 ATWS.
Reactor Pressure is 875 psig and being maintained 850 -1050 psig using SRVs.
Torus Level is 14 feet.
Using T-102 Curve T/T-1 Heat Capacity Temperature Limit, determine the operational implications of these conditions on Reactor Pressure control.
A. An Emergency Blowdown is required.
B. The current Reactor Pressure band is acceptable.
C. Reactor Pressure must be maintained at or above 900 psig.
D. Reactor Pressure must be maintained at or below 899.9 psig.
NRC Question Data Sheet Choice Correct:
D Distractors:
A B
C Answer Key Question ID# 016 Both ROISRO Basis or Justification Current conditions require that pressure be maintained below the curve for the existing DW Level and Temperature conditions.
An emergency blowdown is NOT required since pressure can be maintained at or below 899.9 psig.
The current pressure band is NOT acceptable because it would allow RPV pressure to exceed 899.9 psig.
The candidate would select this answer if he didnt recognize that the safe area of this curve is below the curve. This is a common error because this va ties between curves.
Level of Knowledge Difficulty Time Allowance (minutes)
HIGH 3.0 4
295030 EKI.03 Low Suppression Pool Water Level Reference(s):
Learning Objective:
Importance:
RO I SRO 3.8 14.1 Knowledge/Ability KIA Source Documentation 0 New Exam Item 0 Modified Bank Item ILT Exam Bank 0 Previous NRC Exam 0 Other Exam Bank T-102, Primary Containment Control PLOT-1560.03 (Description of K&A, from catalog)
Knowledge of the operational implications of the following concepts as they apply to Low Suppression Pool Water Level:
Heat Capacity REQUIRED MATERIALS:
Provide T-102 Curve TTT-1 Heat Capacity Temperature Limit Notes and Comments Prepared By:
Philip E. Nielsen (717) 456-4122 PBAPS Regulatory Exam Author Page 32 of 150
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Peach Bottom February 2005 Initial License Operator NRC Examination
- 17.
Use the attached P&IDs to determine when the Startup Bypass Valve (CV-2558) is used and how reactor water level would respond to a loss of instrument air if the Startup Bypass Valve were controlling level?
The Startup Bypass Valve (CV-2558) is used under conditions and reactor level will (2) if instrument air is lost.
(1) reactor pressure A. (1) low (less than 450 psig)
(2) lower B. (1) low (less than 450 psig)
(2) rise C. (1) high (greater than 450 psig)
(2) lower D. (1) high (greater than 450 psig)
(2) rise
NRC Question Data Sheet Level of Knowledge HIGH Answer Key Difficulty Time Allowance (minutes)
SRO 3.75 4
N/A Question ID## 017 Both RO/SRO I
Source:
Reference(s):
Learning Objective:
I Choice New Exam Item Modified Bank Item ILT Exam Bank 0
Previous NRC Exam 0
Other Exam Bank SO 5.7.E-2, Long Path Recirc for Startup Level Control PLOT-5006.06a, 7 I
Basis or Justification I
Knowledge/Ability WA Correct:
295031 G2.1.28 Importance:
RO / SRO Reactor Low Water Level 3.2 / 3.3 Distractors:
A C
SO 5.7.E-2, Long Path Recirc for Startup Level Control, directs the use of the CV-2558 valve and indicates that at 450 psig a RFP must be started.
The valve fails closed on loss of instrument air resulting in a level drop.
The candidate will select this answer if he believes that the CV-2558 fails open (as does the RFP discharge valve bypass controller) on a loss of instrument air.
The candidate will select this answer if he does not recognize that this valve is only used at low pressures because it bypasses all of the RFPs.
D The candidate will select this answer if he doesnt know that this valve is only used under low pressure conditions and incorrectly believes that the CV-2558 fails open (as does the RFP discharge valve bypass controller) on a loss of instrument air.
I Psychometrics I
controls.
REQUIRED MATERIALS:
M-300, Sheets 1 - 2 M-308, Sheet 1 Notes and Comments Prepared By:
Philip E. Nielsen (717) 456-4122 PBAPS Regulatory Exam Author Page 34 of 150
Peach Bottom February 2005 Initial License Operator NRC Examination
- 18.
Unit 2 has experienced an ATWS with a Group 1 Isolation. The following conditions currently exist:
0 Reactorpower: 24%
0 Reactor Level: -100 inches 0
Torus Temperature: 115°F 0
0 SRVs A, B, C, and G are open to control pressure T-117, LeveVPower Control, has directed that RPV Injection be terminated and prevented using T-240.
Under these conditions, the basis for performing T-240, Termination and Prevention of Injection into the RPV, is to limit:
A. power generation that threatens primary containment.
B. uncontrolled injection of large amounts of cold water.
C. neutron flux oscillations that challenge fuel clad integrity.
D. power excursions while establishing Minimum Alternate Flooding Pressure.
NRC Question Data Sheet Correct:
Answer Key A
Question ID# 018 Both RC Level of Knowledge HIGH Choice Difficulty Time Allowance (minutes)
SRO 3.75 4
N/A KnowledgelAbility WA I C
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295037 G 2.4.6 Importance:
RO I SRO SCRAM Condition Present and Power Above the APRM Downscale or Unknown 3.1 /4.0 I D SRO
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Basis or Justification The candidate must determine based on conditions that level is being lowered for power control and then apply the T-117 basis to recognize that the concern is for the containment challenge. (T-117, Step LQ-11)
This the basis for performing T-240 prior to a Blowdown (T-117, Step LQ-21 1 This is the basis for performing T-240 to lower level to below -60" (T-117, Step LQ-13)
This is the basis for performing T-240 prior to establishing MAF pressure in T-116 (T-116, Step RF-25)
Psychometrics Source:
Reference(s):
Learning Objective:
Source Documentation 0
New Exam Item 0
Modified Bank Item ILT Exam Bank 0 Previous NRC Exam 0 Other Exam Bank T-117 LeveVPower Control Bases Step LQ-11 PLOT-1560.09 REQUIRED MATERIALS:
Notes and Comments Question received minor wording changes for clarity but was not significantly modified.
Prepared By:
Philip E. Nielsen (717) 456-4122 PBAPS Regulatory Exam Author Page 36 of 150
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Peach Bottom February 2005 Initial License Operator NRC Examination
- 19.
The following plant conditions exist on Unit 3:
0 0
0 A steam leak exists in the Turbine Building T-104, Radioactivity Release, has been entered due to Vent Stack Radiation above the Hi Hi Alarm setpoints.
An Equipment Operator then reports that Turbine Building Ventilation has tripped.
What is the required operational response to the Turbine Building Ventilation trip and its basis?
A. Restart ventilation to monitor the release.
B. Restart ventilation to lower the radioactive release.
C. Maintain ventilation tripped to lower the radioactive release.
D. Maintain ventilation tripped to prevent an unmonitored release.
NRC Question Data Sheet
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A I
Answer Kev I
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T-104 directs that ventilation be restarted to provide a monitored (and elevated) release (which will also help to maintain personnel accesability).
Question ID# 01 9 Both RO/SRO I
B C
Correct:
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Restarting ventilation will lower the radiation levels in the Turbine Building, but will NOT reduce the radioactive release.
Maintaining ventilation tripped would make the area inaccessible and might permit a ground level unmonitored release.
I Basis or Justification I
D
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Maintaining ventilation tripped would make the area inaccessible and might permit a ground level unmonitored release.
Level of Knowledge Difficulty Time Allowance (minutes)
LOW 2.5 3
1 Source:
Reference(s):
0 bject ive :
KnowledgelAbility WA Learning 0 New Exam Item 0
Modified Bank Item ILT Exam Bank 0
Previous NRC Exam 0
Other Exam Bank T-104, Radioactivity Release PLOT-1560.3,9 295038 EA1.06 Importance:
RO I SRO High Off-Site Release Rate 3.5 13.6 REQUIRED MATERIALS:
Notes and Comments This question was reworded for clarity, but was not Significantly Modified.
Prepared By:
Philip E. Nielsen (717) 456-4122 PBAPS Regulatory Exam Author Page 38 of 150
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Peach Bottom February 2005 Initial License Operator NRC Examination
- 20.
ON-1 14, Actual Fire Reported in the Power Block, Diesel Generator Building, Emergency Pump, Inner Screen, or Emergency Cooling Tower Structure, contains the following note:
NOTE If power is lost to the Motor Driven Fire Pump (OOPO64) controller (E-224 (1 1 13)) for more than 8 seconds, THEN the Motor Driven Fire Pump automatic start feature is defeated. This interlock does NOT affect the ability to manually start the pump. Guidance for resetting the auto start logic can be found in SO 37B. 1.A, Common Plant Fire Water System Lineup for Automatic Operation.
The basis for this feature is to prevent:
A. the pump fiom automatically starting with reduced bus voltage.
B. overloading the diesel generators during a loss of off-site power.
C. a simultaneous start with the Diesel Driven Fire Pump and a water hammer.
D. a spurious start due to loss of power to the fire header pressure instrumentation.
NRC Question Data Sheet Correct:
Distractors:
Answer Kev B
A C
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Question ID# 020 Both RO/SRO Level of Knowledge Difficulty Time Allowance (minutes)
LOW 3.0 3
SRO N/A D
Source:
Reference( s):
Learning Objective:
Basis or Justification 0
New Exam Item 0
Modified Bank Item ILT Exam Bank 0
Previous NRC Exam 0
Other Exam Bank ON-114 Actual Fire Reported in the Power Block, DG Building, Emergency Pump, Inner Screen, or Emergency Cooling Tower Structure Bases Note before Step 6.
PLOTPBIG-1550.04 ON-I 14, Bases for the note says that the defeat of the auto start feature occurs after an 8 second loss of power to prevent an auto start during a LOOP event which could cause an EDG to exceed its 200 hour0.00231 days <br />0.0556 hours <br />3.306878e-4 weeks <br />7.61e-5 months <br /> rating.
The candidate could believe that this interlock is to protect the Fire Pump from damage due to low bus voltage.
Knowledge/Ability KIA
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The candidates are trained to have concern for situations that may cause water hammers and resultant equipment damage and possible personnel safety issues, however, this is not the concern in this situation.
Loss of instrumentation is a plausible concern for causing a causing an undesired system operation.
600000 AK3.04 Importance:
RO I SRO Plant Fire On Site 2.8 / 3.4 REQUIRED MATERIALS:
Notes and Comments This question was reworded for clarity, but was not Significantly Modified.
Prepared By:
Philip E. Nielsen (717) 456-4122 PBAPS Regulatory Exam Author Page 40 of 150
Peach Bottom February 2005 Initial License Operator NRC Examination
- 21.
Given the following:
0 Unit 2 is operating at full power 0
The Unit Reactor Operator (URO) is conducting an In Depth Control Board Walk-Down in accordance with PBAPS Operations Standing Order 04-01.
During his Panel 20C007A walkdown, the URO notes that Off-Gas System Flow on FI-4020 (lower indication) is reading 104 scfm.
Based on these plant conditions and the Off-Gas System Routine Inspection, this value of Off-Gas Flow is:
A. BELOW the expected range; initiate a full Off-Gas Routine Inspection B. WITHIN the expected range; continue with the In Depth Walk-Down C. WITHIN the expected range; but investigate based on the power level D. ABOVE the expected range; monitor for a loss of condenser vacuum
NRC Question Data Sheet Choice Correct:
D Distractors:
A B
C Basis or Justification Based on the Routine Inspection, the expected range for Off-Gas System Flow is 20-45 scfm. This indication is above the expected range and a loss of condenser vacuum is imminent.
104 scfm is above, not below, the expected range. A value too low is also worth investigating for a system irregularity or indication failure.
104 scfm is above, not within, the expected range. If the value was normal, continuing the walk-down would normally be appropriate.
104 scfm is above, not within, the expected range. If the value was within the expected range but at one of the extremes or not what was expected due to power level, then an investigation would be appropriate.
Level of Knowledge Difficulty Time Allowance (minutes)
LOW 3.0 3
SRO N/A REQUIRED MATERIALS:
Source:
Reference(s):
Learning Objective:
Notes and Comments New Exam Item 0
Modified Bank Item 0
ILT Exam Bank 0
Previous NRC Exam 0
Other Exam Bank SO 8.8.A-2 Off-Gas System Routine Inspection PBAPS Operations Standing Order 04-01, Expectations for Control Board Walk-Downs PBIGPLOT-1540.1 Prepared By:
Philip E. Nielsen (717) 456-4122 PBAPS Regulatory Exam Author Knowledge/Ability K/A Page 42 of 150 295002 AA2.04 Importance:
RO / SRO Loss of Main Condenser Vacuum 2.8 / 2.9
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Peach Bottom February 2005 Initial License Operator NRC Examination
- 22.
Unit 3 was manually scrammed following a loss of feedwater. Current plant conditions are as follows:
0 HPCI and RCIC started automatically and injected until they automatically stopped injecting due to high level.
0 Reactor Water Level is 46 inches and dropping slowly.
0 Reactor Pressure is 1020 psig and being controlled automatically by EHC.
Complete the following sentence related to the RCIC systems response to these plant conditions.
The RCIC (1) and the RCIC system will not automatically reinject until Reactor Water Level has dropped to (2)
A. (1) Turbine is tripped (2) + 29 inches.
B. (1) Turbine is tripped (2) - 48 inches.
C. (1) Turbine Supply Valve (MO-131) is closed (2) + 29 inches.
D. (1) Turbine Supply Valve (MO-131) is closed (2) - 48 inches.
NRC Question Data Sheet Level of Knowledge Difficulty Time Allowance (minutes)
LOW 2.5 4
Answer Key SRO NIA Question ID# 022 Both ROISRO
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Knowledge/Ability KIA Choice 295008 AK2.06 Importance:
RO I SRO High Reactor Water Level 3.4 13.6 B
I C
Basis or Justification RClC does not trip on high level (because it can not auto reset a trip like HPCI), instead the Turbine Supply valve (MO-131) goes closed. The valve will not reopen until a -48" initiation signal is received (unlike HPCI which will restart at +29").
RClC does not trip on high level and will not reinject until -48" RClC does not trip on high level.
The RClC Turbine Valve (MO-131) is shut, but the system will not reinject until Reactor Water Level drops to -48 Source:
0 bjective:
Source Documentation New Exam Item 0
Modified Bank Item 0 ILT Exam Bank 0 Previous NRC Exam 0
Other Exam Bank OT-110, Reactor High Level, Step 3.6 PLOT-5013.4b REQUIRED MATERIALS:
Notes and Comments Prepared By:
Philip E. Nielsen (717) 456-4122 PBAPS Regulatory Exam Author Page 44 of 150
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Peach Bottom February 2005 Initial License Operator NRC Examination
- 23.
Unit 3 conditions are as follows:
0 Reactor Power is 100%
0 The C Safety Relief Valve is leaking by slightly.
0 Torus Temperature is 92°F and rising 0
Torus Cooling is in service on the B Loop of RHR 0
River Water Temperature is 85°F Evaluate these conditions and recognize the LOWEST temperature at which actions will need to be taken in accordance with Technical Specifications for exceeding the Suppression Pool Average Temperature Limiting Condition for Operation (LCO).
A. 96°F B. 106°F C. 111°F D. 121°F
NRC Question Data Sheet Correct:
Distractors:
Answer Key A
B
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Question ID# 023 Both RO/SRO Level of Knowledge Difficulty Time Allowance (minutes)
HIGH 3.5 4
Reference@):
0 bjective:
Learning I D New Exam Item 0 Modified Bank Item ILT Exam Bank 0 Previous NRC Exam 0 Other Exam Bank TS 3.6.2.1 Suppression Pool Average Temperature PLOT-5007.01 k Basis or Justification Knowledge/Ability KIA Entry into TS is required at >95OF, unless HPCl testing is in progress (when entry into the TS is not required until Torus Temp is >105OF).
29501 3 G2.1.33 Importance:
RO I SRO High Suppression Pool Temperature 3.4 14.0 With the reactor at power, if HPCl testing is in progress, entry into the TS is not required until Torus Temp is >105"F.
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>llO°F Torus Temperature is a TS value that requires the Reactor to be scrammed.
>120°F Torus Temperature is a TS value that requires the Reactor to be depressurized.
REQUIRED MATERIALS:
Notes and Comments Prepared By:
Philip E. Nielsen (71 7) 456-4122 PBAPS Regulatory Exam Author Page 46 of 150
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Peach Bottom February 2005 Initial License Operator NRC Examination
- 24.
T-220-2, Driving Control Rods during a Failure to Scram, directs the operator to bypass the Rod Worth Minimizer (RWM).
The reason for this step is to permit the:
A. control rods to be manually driven in any sequence.
B. continuous use of the Emergency In/Notch Override switch.
C. operator to exit the current Reactivity Maneuvering Approval (ReMA).
D. simultaneous use of Control Rod Insertion by Manual Scram (T-216-2).
NRC Question Data Sheet A
Answer Key Question ID# 024 Both RO/SRO Basis or Justification The note, just prior to Step 4.3 of T-220-2, states that the RWM is bypassed to permit inserting control rods in any sequence.
Choice Correct:
B C
D Distractors:
The same note gives permission to continuously use the Emergency IrdNotch Override switch when going rod to rod, however, this is unrelated to bypassing the RWM.
Bypassing the RWM does not permit the operator to exit the ReMA, however, during an ATWS condition rod sequence is not limited to the approved sequence.
The note, just prior to step 4.1, states that using T-220-2 may prevent recharging the HCU headers as is required in T-216-2.
Level of Knowledge Difficulty Time Allowance (minutes)
LOW 2.0 3
SRO N/A 29501 5 AK3.01 Incomplete SCRAM Source:
Importance: RO / SRO 3.4 13.7 I Reference(s):
Source Documentation New Exam item 0 Modified Bank Item 0 ILT Exam Bank 0 Previous NRC Exam 0 Other Exam Bank T-220-2 Driving Control Rods during Failure to Scram, Notes before Steps 4.1 &
PLOT-1560.11 Lea m i ng 0 bjective:
Knowledge/Ability K/A (Description of K&A, from catalog)
Knowledge of the reasons for the following responses as they apply to an Incomplete SCRAM:
Bypassing rod insertion blocks REQUIRED MATERIALS:
Notes and Comments Prepared By:
Philip E. Nielsen (717) 456-4122 PBAPS Regulatory Exam Author Page 48 of 150
Peach Bottom February 2005 Initial License Operator NRC Examination
- 25.
In accordance with T-102, Primary Containment Control, Step TL-20, Torus Level is maintained below the TL-1 curve (attached) is to prevent:
A. exceeding the Torus Level TS Limiting Condition for Operation (LCO).
B. flooding the Safety Relief Valve Solenoids rendering the SRVs inoperable.
C. direct pressurization of the Primary Containment without pressure suppression.
D. covering the highest vent capable of passing all the decay heat fiom the reactor.
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NRC Question Data Sheet Level of Knowledge Difficulty Time Allowance (minutes)
LOW 3.5 4
Answer Key SRO N/A lhuestion ID# 025 Both RO/SRO Source:
Reference(s):
Learning Objective:
I Choice New Exam Item 0
Modified Bank Item 0
ILT Exam Bank T-102 Bases Step TIL-23 TRIPEAMP CURVES, TABLES, & LIMITS - BASES, Step 24 0
Previous NRC Exam 0
Other Exam Bank PLOT-1 560.03,09 I
A I
Distractors:
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Knowledge/Ability 295029 G2.4.6 WA High Suppression Pool Water Level Basis or Justification Importance: RO / SRO 3.1 / 4.0 The SRV Tail Pipe Limit Curve is specifically designed to prevent the back pressure during SRV operation from damaging the SRV components potentially causing direct pressurization of containment without pressure suppression.
The Suppression Pool Water Level LCO is exceeded when level is above 14.9 feet, significantly below the values in curve T/L-l.
The concern for rendering the SRVs inoperable due to solenoid flooding is real but does not occur until 21 feet Torus Level as opposed to the just over 17 feet limit on the SRV Tail Pipe Limit Curve.
The concern for exceeding the level of the highest vent that can pass all of the decay heat after shutdown is real but it is not a concern until level has risen to the point that Maximum Containment dP is exceeded which will not occur until level is significantly greater than 21 feet.
REQUIRED MATERIALS:
T-102 Curve T/L-1, SRV Tail Pipe Limit Notes and Comments Prepared By:
Philip E. Nielsen (717) 456-4122 PBAPS Regulatory Exam Author Page 50 of 150
Peach Bottom February 2005
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Initial License Operator NRC Examination
- 26.
The following conditions exist on Unit 3:
A transient occurred resulting in significant he1 damage.
A high radiation condition exists in the Reactor Building due to shine.
The Reactor Building has become a High Radiation Area (General Area dose rates of 12omrem/hr) and has no current valid Radiation Work Permit (RWP).
Operations personnel must enter the Reactor Building for one hour to help mitigate the transient and save plant equipment.
In accordance with RP-AA-460, Controls for High and Very High Radiation Areas, the MINIMUM requirement for an operator to enter the area is that they MUST receive:
A. coverage by a qualified Advanced Rad Worker (ARW).
B. coverage by an ANSI Radiation Protection Technician (RPT).
C. permission fiom the Radiation Protection Manager (RPM).
D. permission from the Emergency Director (ED) after Emergency Plan activation.
NRC Question Data Sheet Level of Knowledge D i fTi cu I ty LOW 3.0 4
Time Allowance (minutes)
Answer Key SRO NIA Question ID# 026 Both RO/SRO 295033 EK1.02 High Secondary Containment Area Radiation Levels Choice Correct:
Importance:
RO I SRO 3.9 14.2 Distractors :
B A
C D
Basis or Justification
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RP-AA-403 requires that this coverage be provided to meet the objectives of the RWP program.
An ARW qualified individual is NOT sufficient to provide the required coverage.
The procedure requires the RPT to notify RP Management as soon as possible, but their permission is not required for entry.
The EDS permission is not required unless a dose extension is required for entry into the High Radiation Area.
Source :
Reference@):
Learning Objective:
KnowledgelAbility WA Source Documentation New Exam Item 0 Modified Bank Item 0 ILT Exam Bank 0
Previous NRC Exam 0
Other Exam Bank RP-AA-403 Radiation Work Permit Program RP-AA-460 Controls for High and Very High Radiation Areas PLOT-1760.04 (Description of K&A, from catalog)
Knowledge of the operational implications of the following concepts as they apply to High Secondary Containment Area Radiation Levels:
Personnel protection REQUIRED MATERIALS:
Notes and Comments Prepared By:
Philip E. Nielsen (717) 456-4122 PBAPS Regulatory Exam Author Page 52 of 150
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Peach Bottom February 2005 Initial License Operator NRC Examination
- 27.
A Designated Alternate (DA) is moving an old jet pump in the Unit 2 Fuel Pool when it falls off the auxiliary hoist. It is reported to the Control Room that a jet pump fell on an irradiated fuel bundle and damaged some fuel pins.
The Control Room also receives the following alarms and indications:
Refueling Floor Vent Exhaust Hi Radiation (21 8 A-1)
Reac. Bldg. Zone Vent Exhaust Hi Radiation (218 B-1) 0 Reac. Bldg. or Refueling Floor Vent Exh. Hi Rad Trip (218 D-4) 0 Refueling Floor Radiation Trip Units A and D High Lights are lit Evaluate these conditions and determine the ventilation response.
A. Reactor Building Ventilation trips.
Refuel Floor Ventilation trips.
SBGT initiates and aligns to the entire Reactor BuildingRefuel Floor.
B. Reactor Building Ventilation continues to run.
Refuel Floor Ventilation trips.
SBGT initiates and aligns to the Refuel Floor.
C. Reactor Building Ventilation continues to run.
Refuel Floor Ventilation continues to run.
SBGT initiates and aligns to the Refuel Floor.
D. Reactor Building Ventilation continues to run.
Refuel Floor Ventilation continues to run.
SBGT remains in standby.
NRC Question Data Sheet Level of Knowledge Difficulty Time Allowance (minutes)
HIGH 3.5 4
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Answer Key SRO N/A Question ID# 027 Both RC Choice Source:
Reference(s):
Learning Objective:
Correct:
0 New Exam Item 0 Modified Bank Item ILT Exam Bank 0
Previous NRC Exam 0 Other Exam Bank ARC 218 A-I Refueling Floor Vent Exhaust Hi Rad ARC 218 B-I Reac. Bldg. Zone Vent Exhaust Hi Rad ARC 218 D-4 Reac. Bldg. or Refueling Floor Vent Exhaust Hi Rad Trip PLOT-5007G. I c
Distractors:
Knowledge/Ability K/A A
295034 EA1.03 Importance:
RO I SRO Secondary Containment Ventilation High Radiation.
4.0 13.9 B
C D
'S RO I
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Basis or Justification The trip of "A" and 'D" Refuel Floor rad will result in a Group Ill isolation.
The Group Ill isolation will trip both Reactor Bldg. and Refuel Floor Ventilation and align SBGT to the entire Reactor Building and Refuel Floor.
Reactor Building will also trip even though the high radiation was on the Refuel Floor. SBGT is aligned to both the Refuel Floor and Reactor Building.
Both Reactor Building and Refuel Floor Ventilation will trip and SBGT will be aligned to both areas.
Both Reactor Building and Refuel Floor Ventilation will trip and SBGT will be aligned to both areas.
Secondary Containment Ventilation EQUIRED MATERIALS:
Notes and Comments Prepared By:
Philip E. Nielsen (717) 4564122 PBAPS Regulatory Exam Author Page 54 of 150
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Peach Bottom February 2005 Initial License Operator NRC Examination
- 28.
Unit 2 has experienced a Loss of Coolant Accident. The following conditions are present:
0 To control Reactor Level:
Reactor Pressure is 200 psig Drywell Pressure is 8 psig.
Reactor Level is -120 inches and rising due to Low Pressure ECCS injection.
+ All Core Spray Pumps started automatically and have been secured.
+ A, B, C RHR Pumps started automatically and have been secured.
+ D RHR Pump started automatically and continues to inject at full flow.
+ All Condensate Pumps are tripped.
Seven (7) minutes have elapsed since the LOCA signal sealed in 0
The Reactor Operator (RO) has been directed to control level by throttling the discharge path of the D RHR Pump. Evaluate the given conditions to determine the method for throttling the discharge path and any RHR System Operating procedure pump flow limitations.
The RO must throttle on the (1) and the MINIMUM steady state flow rate is throttled to maintain at or above (2)
A. (1) RHR Loop B Inboard Discharge Valve (MO 10-025B)
(2) 4000gpm B. (1) RHR Loop B Inboard Discharge Valve (MO-2-10-025B)
(2) 6500gpm C. (1) RHR Outboard Discharge Valve (MO-2-10-154B)
(2) 4000gpm D. (1) RHR Outboard Discharge Valve (MO-2-10-154B)
(2) 6500gpm
~ _ _ _
~~
NRC Question Data Sheet Psychometrics Level of Knowledge Difficulty Time Allowance (minutes)
SRO HIGH 2.5 4
N/A I
Answer Key Source:
Reference(s):
Learning Objective:
I Question ID# 028 Both RO/SRO New Exam Item 0 Modified Bank Item 0 ILT Exam Bank 0
Previous NRC Exam 0
Other Exam Bank SO 10.7.8-2 RHR System Automatic Response during LOCA conditions PLOT-5010.4j,5d I
Knowledge/Ability WA Correct:
Distractors:
203000 A4.02 Importance:
RO / SRO RHR/LPCI Injection 4.1 /4.1 C
A B
D Basis or Justification The MO-1548 valve is available to be throttled 5 minutes after the LOCA signal seals in. The minimum flow specified by the procedure is 4000 gpm.
The MO-25B Valve is sealed open as long as the LOCA signal is present.
Evaluating current conditions (<450 psig Reactor Pressure with >2psig Drywell Pressure) shows that the LOCA signal cannot be cleared. The indicated minimum flow of 4000 gpm is correct.
The MO-25B Valve is sealed open as long as the LOCA signal is present.
Evaluating current conditions ( ~ 4 5 0 psig Reactor Pressure with >2psig Drywell Pressure) shows that the LOCA signal cannot be cleared. The indicated minimum flow of 6500 gpm is also wrong and corresponds to the max permissible RHR flow while in Shutdown Cooling with core instrumentation not supported by blade guides or fuel bundles on all sides.
The MO-154B valve is available to be throttled 5 minutes after the LOCA signal seals in, however, the minimum flow specified by the procedure is 4000 gpm rather than 6500 gpm. 6500 gpm is the max permissible RHR flow while in Shutdown Cooling with core instrumentation not supported by blade guides or fuel bundles on all sides.
I System Valves REQUIRED MATERIALS:
Notes and Comments Prepared By:
Philip E. Nielsen (717) 4564122 PBAPS Regulatory Exam Author Page 56 of 150
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Peach Bottom February 2005 Initial License Operator NRC Examination
- 29.
SO 10.7.A-2, Residual Heat Removal System LPCI Mode Manual Start, cautions the operator to recognize that under specific conditions existing 480V and 120V loads supplied by the Emergency Bus are anticipated to drop out (lose power) when starting a RHR Pump.
Existing 480V and 120V loads supplied by the emergency bus are expected to lose power upon starting a RHR pump when the Emergency Bus is powered by its (1) due to (2)
A. (1) Normal Off-Site Feed (2) normal Emergency Bus load shedding.
B. (1) Normal Off-Site Feed (2) a momentary low voltage condition.
C. (1) Emergency Diesel Generator (2) normal Emergency Bus load shedding.
D. (1) Emergency Diesel Generator (2) a momentary low voltage condition.
NRC Question Data Sheet Level of Knowledge Difficulty Time Allowance (minutes)
LOW 3.25 3
Answer Kev SRO N/A
~
Question ID# 029 Both RO/SRO Source:
Reference(s):
Learning Knowledge/Ability WA Objective:
Choice Correct:
New Exam Item 0
Modified Bank Item 0 ILT Exam Bank 0
Previous NRC Exam 0
Other Exam Bank SO 10.7.A-2 RHR System LPCl Mode Manual Start PLOT-5054.1 a 203000 G2.1.32 Importance:
RO I SRO RHWLPCI Injection 3.4 13.8 PLOT-501 0.1 g D is t ractors:
D A
B C
Basis or Justification
~~
When starting a RHR pump with the E-bus supplied by the EDG, voltage may droop momentarily resulting in individual 480V and 120V loads tripping on undervoltage. The E-bus will not experience normal loadshedding.
The Normal Off-Site Feed is not expected to droop voltage during an RHR pump start enough to cause the 480V and 12OV loads to drop out. The cautioned about effect is from individual loads tripping on undervoltage.
Normal E-bus Load Shedding occurs when the e-bus is deenergized and reenergized with the EDG.
The Normal Off-Site Feed is not expected to droop voltage during an RHR pump start causing individual 480V and 120V loads to drop out.
It is correct that this effect occurs when the E-bus is powered by the EDG, however, the power losses are not due to a Normal Load Shedding but rather due to individual loads tripping on undervoltage.
RHRILPCI Injection: Ability to explain and apply system limits and precautions.
REQUIRED MATERIALS:
Notes and Comments Prepared By:
Philip E. Nielsen (717) 456-4122 PBAPS Regulatory Exam Author Page 58 of 150
Peach Bottom February 2005 Initial License Operator NRC Examination
- 30.
Select from the following the answer that describes the correct Shutdown Cooling suction valve electrical power supplies.
The MO 10- 17, S/D Cooling Outboard is powered by a The MO-2-10-18, S/D Cooling Inboard is powered by a (1)
(2)
A. (1) 250 VDC Safety Related Bus (2) 250 VDC Safety Related Bus B. (1) 480V Emergency Bus MCC (2) 250 VDC Safety Related Bus C. (1) 250 VDC Safety Related Bus (2) 480V Emergency Bus MCC D. (1) 480V Emergency Bus MCC (2) 480V Emergency Bus MCC
NRC Question Data Sheet Correct:
Distractors:
Answer Key C
A Question ID# 030 Both RO/SRO r
Psychometrics Level of Knowledge Difficulty Time Allowance (minutes)
NIA 205000 K2.02 Shutdown Cooling Mode I B Importance:
Basis or Justification I
MO-2-10-17 is powered by a 250 VDC Safety Related Bus.
MO-2-10-18 is powered by a 480V E Bus supplied MCC.
~~
MO-2-10-17 is powered by a 250 VDC Safety Related Bus.
MO-2-10-18 is powered by a 480V E Bus supplied MCC.
~
MO-2-10-17 is powered by a 250 VDC Safety Related Bus.
MO-2-10-18 is powered by a 480V E Bus supplied MCC.
~
MO-2-10-17 is powered by a 250 VDC Safety Related Bus.
MO-2-10-18 is powered by a 480V E Bus supplied MCC.
Source:
Reference(s):
Learning Objective:
KnowledgeIAbility WA Source Documentation New Exam Item Modified Bank Item 0
ILT Exam Bank Previous NRC Exam 0 Other Exam Bank SO 10.1.8-2 RHR System Shutdown Cooling Mode Manual Start PLOT-501 0.2b (Description of K U, from catalog)
Knowledge of the electrical power supplies to the following:
Shutdown Cooling Mode: Motor operated valves REQUIRED MATERIALS:
Notes and Comments Prepared By:
Philip E. Nielsen (717) 456-4122 PBAPS Regulatory Exam Author Page 60 of 150
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Peach Bottom February 2005 Initial License Operator NRC Examination 3 1.
Unit 2 was manually scrammed due to a steam leak in the primary containment.
Current plant conditions are as follows:
0 0
0 Drywell Pressure is 5 psig and rising.
HPCI started automatically and injected until it tripped on high level.
Reactor Water Level is +46 inches and dropping slowly.
Reactor Pressure is 1000 psig and being controlled automatically by EHC.
The HPCI System:
A. will reset and reinject automatically when Reactor Level drops to +29 inches.
B. will reset and reinject automatically when Reactor Level drops to -48 inches.
C. cannot be manually reset for reinjection until Reactor Level drops to +29 inches.
D. cannot be manually reset for reinjection until Reactor Level drops to -48 inches.
NRC Question Data Sheet Level of Knowledge Difficulty HIGH 2.0 Answer Key Time Allowance (minutes)
SRO 4
NIA Question ID# 031 Both R(
Source:
Reference@):
Objective:
Learning I
Choice
[x1 New Exam Item 0
Modified Bank Item 0
ILT Exam Bank 0
Previous NRC Exam 0 Other Exam Bank SO 23.7.A-2 HPCl System Automatic Initiation Response PLOT-5023.4~
I correct:
Knowledge/Ability KIA Distractors:
206000 K4.03 Importance: RO I SRO HPCl 4.2 / 4.1 I
A B
C D
SRO Basis or Justification HPCl resets turbine trips automatically when the high level trip signal clears. The high Reactor Level signal clears at +29". The system will reinject automatically because an initiation signal (DW Press) still exists.
HPCl resets turbine trips automatically when the high level trip signal clears, however, the high Reactor Level signal clears at +29", not -48". The system will reinject automatically because an initiation signal (DW Press) still exists.
HPCl resets turbine trips automatically when the high level trip signal clears and does not require manual action for this to occur. The high Reactor Level trip signal does clear at +29". The system will then reinject automatically because an initiation signal (DW Press) still exists.
HPCl resets turbine trips automatically when the high level trip signal dears and does not require manual action for this to occur. The high Reactor Level trip signal does clears at +29", not -48". The system will then reinject automatically because an initiation signal (DW Press) still exists.
Notes and Comments Prepared By:
Philip E. Nielsen (717) 4564122 PBAPS Regulatory Exam Author Page 62 of 150
Peach Bottom February 2005 Initial License Operator NRC Examination
- 32.
Following a Unit 3 transient with a Group I isolation, HPCI is operating in CST-CST mode with the following conditions present:
0 0
0 0
0 0
HPCI Flow Rate is set to 5000 gpm HPCI Discharge Pressure is 800 psig HPCI Flow Controller is in Automatic set to 5000 gpm.
MO-2-23-19, To Feed Line Valve, is closed.
Reactor Level is +20 inches and being maintained with RCIC injection.
Reactor Pressure is 1000 psig and rising slowly The Reactor Operator (RO) has been directed to raise HPCI system discharge pressure to maximize HPCI System steam usage to assist in Reactor Pressure control.
To RAISE the HPCI Discharge Pressure and steam usage in accordance with the Rapid Response Card (RRC-23. 1-2), the RO must throttle:
A. CLOSED on the MO-2-23-21, Full Flow Test Valve.
B. OPEN on the MO-2-23-21, Full Flow Test Valve.
C. CLOSED on the MO-2-23-24, Cond. Tank Return.
D. OPEN on the MO-2-23-24, Cond. Tank Return.
NRC Question Data Sheet A
B C
D Answer Key Question ID# 032 Both RO/SRO Basis or Justification The procedure directs throttling closed on the MO-21, Full Flow Test Valve, to cause increased resistance to flow requiring more steam to be used to maintain the set flowrate.
The candidate could believe that throttling open on the MO-21, Full Flow Test Valve, will use more steam because it will be pumping at a higher rate.
This is not accurate, because the flow controller will maintain flow automatically.
Although flow to the Condensate Storage Tank must be restricted to raise pressure, the procedure does not permit throttling of the Condensate Tank Return Valve.
The Condensate Storage Tank Return Valve will be maintained full open by procedure.
Choice Level of Knowledge Difficulty HIGH 2.25 Correct:
Time Allowance (minutes)
SRO 4
NIA Distractors:
Source:
Reference@):
Objective:
Knowledge/Ability KIA Learning New Exam Item 0
Modified Bank Item 0
ILT Exam Bank 0
Previous NRC Exam 0
Other Exam Bank SO 23.1.B-3 HPCl System Manual Operation PLOT-5023.1 c 206000 Al.07 Importance: RO I SRO HPCl 3.7 13.6 REQUIRED MATERIALS:
Notes and Comments Prepared By:
Philip E. Nielsen (717) 456-4122 PBAPS Regulatory Exam Author Page 64 of 150
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Peach Bottom February 2005 Initial License Operator NRC Examination
- 33.
Unit 3 has experienced a Loss of Coolant Accident. The following plant conditions exist:
0 Reactor Level initially dropped to -180 inches 0
Reactor Level has now recovered to -150 inches and rising slowly 0
All RHR and Core Spray Pumps started automatically and are injecting.
0 The operators have monitored RHR and Core Spray operation but have not manually manipulated any associated controls.
0 Reactor Pressure is 200 psig 0
Drywell Pressure is 16 psig With these conditions present, all off-site power is lost. All of the Emergency Diesel Generators (EDGs) start and reenergize their expected busses.
Based on these conditions, complete the following statement to describe how the Core Spray Pumps will be restarted to control Reactor Level when power is restored.
The Core Spray Pumps will Emergency Diesel Generators reaches 95% of rated voltage.
when the A. require a manual restart B. automatically restart immediately C. automatically restart after a six (6) second time delay D. automatically restart sequentially after 13 seconds (A, C) and 23 seconds (B, D)
NRC Question Data Sheet Correct:
Answer Key Question ID# 033 Both RO/SRO C
I Choice Level of Knowledge Difficulty Time Allowance (minutes)
HIGH 3.0 4
SRO N/A Basis or Justification With the EDGs at 95% of their rated voltage, all Core Spray Pumps will start after a 6 second time delay.
The candidate could have concerns about the CS Pump Breaker Locking out and requiring manual action to restart, but in this situation, the Core Spray Pumps will restart automatically.
The system imposes a 6 second time delay for Core Spray Pump starts to allow the bus to recover from the immediate RHR Pump Starts.
The sequential loading of the busses reflect those times delays that would occur if the E-Bus was powered by its alternate off-site source rather than the EDG.
209001 K1.08 LPCS Importance:
Reference@):
Learning Objective:
Knowledge/Ability WA Source Documentation New Exam Item 0
Modified Bank Item 0 ILT Exam Bank 0
Previous NRC Exam 0
Other Exam Bank SO 14.7.A-2 Core Spray System Automatic Response during LOCA PLOT-5014.1 I (Description of K&A, from catalog)
Knowledge of physical connections and/or cause-effect relationships between LPCS and the following:
A.C. Electrical Power REQUIRED MATERIALS:
Notes and Comments Prepared By:
Philip E. Nielsen (717) 456-4122 PBAPS Regulatory Exam Author Page 66 of 150
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Peach Bottom February 2005 Initial License Operator NRC Examination
- 34.
Unit 2 is operating at full power when the Standby Liquid Control injection sparger becomes completely clogged with debris. This has resulted in the pressure input from this line being 10 psig lower than actual.
Evaluate this condition to determine the impact, if any, on indicated Core Plate Flow as read on the Control Room Flow Recorder (FR-095).
Indicated Core Plate Flow on FR-095 will be:
A. unaffected.
B. failed downscale.
C. lower than actual.
D. higher than actual.
NRC Question Data Sheet C
A B
D Answer Key Question ID# 034 Both RO/SRO The SBLC Injection line is the below core tap for Core Plate dP and flow.
With the below core plate tap having a lower pressure signal, the indicated dP will be lower, resulting in a lower than actual Core Plate Flow Indication.
If the candidate believes that the Core Plate Flow indication uses taps unrelated to the SBLC injection line as do Control Rod Drive dP and Core Spray Line Break Detection, he would believe that the indication would be unaffected.
The candidate may believe that the Core Plate Flow indication will fail upscale if it receives faulty inputs (as some other new recorders do).
If the candidate believes that SBLC injects through the above core plate line, they would determine that indicated Core Plate Flow would be high.
Choice Level of Knowledge D ifficu I ty Time Allowance (minutes)
HIGH 3.5 5
I Basis or Justification SRO NIA Correct:
Source:
Reference(s):
Objective:
Knowledge/Ability WA Learning Distractors:
New Exam Item 0
Modified Bank Item 0
ILT Exam Bank Previous NRC Exam 0 Other Exam Bank M-351 & M-352 Piping and Instrument Diagrams (P&IDs)
PLOT-5011.3~
21 1000 K3.03 Importance: RO I SRO SLC 2.6 12.7 REQUIRED MATERIALS:
Notes and Comments Prepared By:
Philip E. Nielsen (717) 456-4122 PBAPS Regulatory Exam Author Page 68 of 150
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Peach Bottom February 2005 Initial License Operator NRC Examination
- 35.
Unit 2 has experienced an Anticipated Transient Without Scram (ATWS) condition resulting in the need to inject Standby Liquid Control (SBLC). The SBLC Keylock Switch has been placed in the Start System A position.
Evaluate these conditions to determine the condition of the amber colored Squib Continuity Lights located on the C05 panel indicating a successfbl SBLC initiation.
Squib Continuity Light A will be (1) and Squib Continuity Light B will be (2).
A. (1) lit (2) lit B. (1) lit (2) off
- c. (1) off (2) lit D. (1) off (2) off
NRC Question Data Sheet Choice Correct:
A Basis or Justification The Squib Continuity Lights are in the initiation circuit in a manner that makes them stay on, even after the valves have been fired, until the SBLC Distractors:
Since the A System was started, the candidate could believe that its associated continuity light would go out when the valve fired.
l c I Pump is secured.
Since the A System was started, the candidate could believe that only its continuity light would remain lit after injection.
B If the candidate recognizes that both valves fire when the A system is started, he could expect that their associated continuity lights would go out.
I D I Level of Knowledge Difficulty Time Allowance (minutes)
LOW 2.25 3
SRO N/A Source:
Reference( s):
Objective:
Knowledge/Ability KIA Lea m i ng REQUIRED MATERIALS:
New Exam Item 0 Modified Bank Item a
ILT Exam Bank 0
Previous NRC Exam Other Exam Bank SO 11.1.B-2 Standby Liquid Control System Initiation PLOTdOll.4d 21 1000 A4.03 Importance: RO / SRO SLC 4.1 14.1 Notes and Comments Prepared By:
Philip E. Nielsen (717) 456-4122 PBAPS Regulatory Exam Author Page 70 of 150
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Peach Bottom February 2005 Initial License Operator NRC Examination
- 36.
Unit 2 conditions:
0 Reactor Power is 100%.
0 Both W S Busses were aligned to their normal RPS MG Set power supplies.
0 A loss of one off-site startup feed occurred causing a 4KV Emergency Bus Fast Transfer.
0 The transfer occurred as designed and restored power to the impacted Emergency Busses fiom the other startup feed.
Evaluate this condition and complete the following sentence to identify the designed response of the A and B W S MG Sets to this transient.
Following the fast transfer, the EARLIEST power will be restored is in approximately seconds to maintain continuous power to the 120 VAC RPS Bus.
A. 0.25 B. 3.25 C. 8.0 D. 13.0
NRC Question Data Sheet Choice Answer Key Basis or Justification B
A Correct:
The 4KV Bus will fast transfer in 0.25 seconds, then 3 seconds later the Emergency Bus MCC will reclose providing 48OVAC power back to the RPS MG Set.
Although the 4KV Bus will fast transfer in 0.25 seconds, there is another 3 seconds until the Emergency Bus MCC is reenergized.
Distractors:
Level of Knowledge Dificu I ty Time Allowance (minutes)
HIGH 2.0 3
SRO N/A 8 seconds corresponds to the time delay before the RPS MG Set Supply Breakers trip on a loss of power.
212000 K2.01 RPS 13 seconds corresponds to 10 seconds for the Emergency Diesel to start and 3 additional seconds for the Emergency Bus MCC to reenergize. The 8 second time delay trip is designed to cause a loss of RPS (and scram) rather than allowing the system to continue operation after a complete loss of off-site power.
Importance:
I Reference(s):
Learning KnowledgelAbility Source Documentation New Exam Item 0
Modified Bank Item 0
ILT Exam Bank 0
Previous NRC Exam 0
Other Exam Bank SO 54.7.A 4KV Transfer Load Shedding and Sequential Loading on Bus Undervoltage PLOT-5060F.2b (Description of K&A, from catalog)
Knowledge of the electrical power supplies to the following:
RPS motor-generator sets REQUIRED MATERIALS:
Notes and Comments Prepared By:
Philip E. Nielsen (717) 4564122 PBAPS Regulatory Exam Author Page 72 of 150
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Peach Bottom February 2005 Initial License Operator NRC Examination
- 37.
Unit 3 experienced a full scram on low condenser vacuum. The following conditions currently exist:
All routine scram actions were completed satisfactorily.
0 Draining of the Scram Discharge Volume (SDV) is on hold while Radiation Protection Technicians conduct surveys.
All other scram conditions were clear or automatically bypassed.
0 RPS was successfully reset using GP-1 lE, Reactor Protection System - Scram and ARI Reset.
The Reactor Engineers have requested that the Mode Selector Switch (MSS) be placed in REFUEL so that a group of rods could be stroked.
0 Just as the mode switch was placed in REFUEL, the B RPS MG Set trips causing the B RPS Bus to be deenergized.
Evaluate these conditions to determine the response of the RPS system and the reason for the response.
Based on these conditions, the plant will receive a:
A. HALF Scram ONLY due to the loss of power to By RPS.
B. FULL Scram due ONLY to the loss of power to B RPS.
C. FULL Scram due ONLY to taking the MSS out of SHUTDOWN D. WLL Scram due to BOTH the loss of power to B RPS and moving the MSS.
NRC Question Data Sheet Level of Knowledge HIGH I
Answer Key I
Difficulty Time Allowance (minutes)
SRO 3.75 3
NIA I Question ID# 037 Both R(
Source:
Reference( s):
Learning Objective:
I Choice New Exam Item 0
Modified Bank Item 0
ILT Exam Bank GP 11.E RPS Scram and ARI Reset PLOT-5060F.4k, 6a 0 Previous NRC Exam 0 Other Exam Bank Correct:
Knowledge/Ability WA Distractors:
21 2000 G2.1.28 Importance:
A C
D SRO Basis or Justification A full scram would result because a loss of either RPS MG Set will cause a loss of the power to the bypass circuitry for the SDV Hi Level Bypass. If SDV level is still greater than 50 gallons, a full scram will result.
A full scram will occur based on these conditions, therefore indicating a Half Scram Only is incorrect.
A full scram will occur but it is NOT related to putting the MSS in REFUEL.
Placing the MSS in Startup or Run will also defeat the bypass and cause a full scram.
A full scram will occur but it is NOT related to putting the MSS in REFUEL.
Placing the MSS in Startup or Run will also defeat the bypass and cause a full scram.
REQUIRED MATERIALS:
Notes and Comments Prepared By:
Philip E. Nielsen (71 7) 456-41 22 PBAPS Regulatory Exam Author Page 74 of 150
Peach Bottom February 2005 Initial License Operator NRC Examination
- 38.
Unit 2 is in MODE 2 performing a reactor startup. Control Rod withdrawal has begun, but the reactor is still subcritical.
Which of the following describes the response if power to the 2B 24/48 VDC Distribution Panel (20D45) is lost?
A. WRNM channels A, C, E, and G ODAs deenergize on panel 20C05.
B. WRNM channels B, D, F, and H ODAs deenergize on panel 20C05.
C. A Channel RPS Half Scram and a Rod Block D. B Channel RPS Half Scram and a Rod Block
NRC Question Data Sheet Level of Knowledge LOW Question ID# 038 Both R(
Choice Difficulty Time Allowance (minutes)
SRO 3.25 3
N/A Correct:
Source:
Reference(s):
Learning 0 bjective:
Distractors:
0 New Exam Item 0
Modified Bank Item ILT Exam Bank ARC 21 0 H-3 B WRNM TripAnop PLOT 5060C2c,4a,4b Previous NRC Exam 2001 0
Other Exam Bank D
Knowledge/Ability KIA A
21 5003 K6.02 Importance: RO / SRO IRM (WRNM at PBAPS) 3.6 13.8 B
C SRO Basis or Justification This loss of power will result in a half scram and a rod block The A,C,E and G ODAs will not deenergize.
The B, D, F, and H ODAs will not deenergize.
This power supply goes to the B not A logic train.
REQUIRED MATERIALS:
Notes and Comments Prepared By:
Philip E. Nielsen (717) 4564122 PBAPS Regulatory Exam Author Page 76 of 150
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Peach Bottom February 2005 Initial License Operator NRC Examination
- 39.
Unit 2 is operating at full power with APRM Channel 1 bypassed due to an INOP fad t trip.
An unrelated problem now results in the following condition:
APRM Channel 3 if found to be failed upscale APRM/OPRM HI-HVINOP (21 1 A-3) Alarms Evaluate these conditions to determine the plant response and required procedural response to the event.
A. A FULL S c m will occur, enter and execute the Scram procedure, T-100.
B. A HALF Scram will occur, bypass the APRM and reset the half per GP-1 1E.
C. NO Scram Channels trip, bypass the APRM in accordance with ARC 21 1 A-3.
D. NO Scram Channels trip, evaluate keeping the APRM tripped with Tech Specs.
NRC Question Data Sheet Correct:
Distractors:
I Answer Key I
D A
B I Question ID# 039 Both RO/SRO I
Level of Knowledge Difficulty HIGH 2.5 I
Choice Time Allowance (minutes)
SRO 3
NIA Source:
Reference(s):
Learning 0 bjective:
C
[XI New Exam Item Modified Bank Item ILT Exam Bank 0
Previous NRC Exam 0
Other Exam Bank ARC 21 1 A-3 PLOT-5060.3a Basis or Justification I
Knowledge/Ability KIA The RPS system requires two operable APRMs to trip to result in a scram signal. The candidate must know that it would be incorrect to bypass the APRM with one already bypassed.
Two APRMs upscale or inop would result in a full reactor scram requiring entry into T-1 00, Scram. The candidate must recognize that this does not apply to bypassed APRMs.
With the old GE APRMs, this condition would have caused a half scram.
The candidate must know that since we recently installed the new GE NUMAC system two operable APRMs must trip to give a scram condition.
The RPS system requires two operable APRMs to trip to result in a scram signal. The candidate must know that it would be incorrect to bypass the APRM with one already bypassed.
21 5005 A2.02 Importance: RO I SRO APRM/LPRM 3.6 13.7 REQUIRED MATERIALS:
Notes and Comments Prepared By:
Philip E. Nielsen (71 7) 456-41 22 PBAPS Regulatory Exam Author Page 78 of 150
Peach Bottom February 2005 Initial License Operator NRC Examination
- 40.
Unit 2 was operating at full power when the following conditions occurred.
2 DB HPCI 250V DC BUS LO VOLTAGE (220 G-3) Alarms 0
0 0
The PRO reports that most position indicating lights for the HPCI valves are off.
The crew enters SE-13, Loss of a 125 or 250 VDC Safety Related Bus.
The investigation determines that 2DB-R-B (20D1 l), a 250 VDC Division II Bus is deenergized Evaluate these conditions to determine the effect on the operation of the Reactor Core Isolation Cooling (RCIC) system.
A. Operation in all modes will be unaffected.
B. Operation in the CST-CST Mode is unavailable.
C. RCIC Will NOT auto inject on a Low Reactor Water Level.
D. RCIC will NOT auto swap fi-om CST to Torus suction on Low CST Level.
NRC Question Data Sheet Level of Knowledge Difficulty Time Allowance (minutes)
HIGH 3.25 3
Answer Key 1
SRO N/A Question ID# 040 Both R(
Source:
Reference(s):
Learning Objective:
I Choice New Exam Item 0
Modified Bank Item ILT Exam Bank 0
Previous NRC Exam 0
Other Exam Bank ARC 220 G3,2DB HPCl250 VDC BUS LO VOLTAGE SE-13, Loss of a 125 or 250 VDC Safety Related Bus PLOT-5013.2c,6a Correct:
Knowledge/Ability K/A Distractors:
21 7000 K6.01 Importance:
A C
D SRO I
~
Basis or Justification Operation in CST-CST requires the MO-23-24, Common Test Return to CST, to be opened. It is a HPCl valve and loses power with the loss of 20D11 with the other HPCl valves.
Although this power supply loss does not directly remove power from any RClC components, operation in the CST-CST mode is unavailable as explained in B above.
The annunciator for REACTOR WATER LEVEL LOW LOW (221 E-5) is located on with the HPCl annunciators and the candidate could believe that this function would fail with the loss of HPCl power.
The CONDENSATE STOR TANK LEVEL LOW-LOW (221 C-3) alarm is located with the HPCl alarms and the candidate could believe that this function would be deenergized when HPCl loses power. There are actually separate RClC relays.
Page 80 of 150
Peach Bottom February 2005 Initial License Operator NRC Examination
- 41.
Unit 2 was operating at full power:
I&C has been conducting testing on the Discharge Pressure switches for the Core Spray (CS) and Residual Heat Removal (RHR) pumps and reports the following:
0 0
0 The Discharge Pressure switches from C and D CS Pumps are NOT responding.
The Discharge Pressure switches from A and B CS Pumps are operable.
All of the FWR Pump Discharge Pressure switches are operable.
Unit 2 now experiences a Small Break Loss of Coolant Accident (LOCA) due to a steam leak in the Drywell. The following conditions exist:
0 The Reactor is scrammed.
0 Drywell Pressure is 12 psig.
0 Reactor Pressure is 1000 psig.
0 No High Pressure Feed is available.
0 Reactor Level is -170 inches.
0 All RHR and Core Spray Pumps are running.
0 The Automatic Depressurization System (ADS) has NOT yet been inhibited.
Given these conditions, the Automatic Depressurization System (ADS) logic can:
A. NOT successfully initiate an automatic depressurization.
B. NOT initiate an automatic depressurization until the 9 Minute Timer times out.
C. initiate on inputs from either a RHR pump OR the A and B CS Pumps.
D. initiate on inputs from a RHR pump, but NOT the A and B CS Pumps.
~-
NRC Question Data Sheet Choice Correct:
D Distractors:
A Basis or Justification ADS Logic can be initiated, but it requires a RHR pump running input because A & B CS pumps can not make up the proper logic for CS alone.
For CS, the ADS logic requires A OR 6 AND C OR D.
ADS Logic can be initiated, but it requires a RHR pump running input because A & B CS pumps can not make up the proper logic for CS alone.
B C
The 9-Minute Timer does bypass one requirement for a blowdown to be initiated, however, it is the requirement to have a 2# Drywell pressure signal. It has no effect on the pump running sensing logic.
ADS Logic can NOT be initiated based on Core Pump discharge pressure because it does not meet the logic requirements of A OR B AND C OR ID.
~
Source:
Level of Knowledge Difficulty Time Allowance (minutes)
HIGH 3.75 3
Reference(s):
Learning 0 bjective:
Knowledge/Abi I ity K/A SRO N/A Source Documentation New Exam Item 0
Modified Bank Item 0
ILT Exam Bank 0 Previous NRC Exam 0 Other Exam Bank M-1 -S 52, Automatic Depressurization System PLOT-5001 G.6b 218000 K6.02 ADS Importance:
RO / SRO 4.1 !4.1 (Description of K&A, from catalog)
Knowledge of the effect that a loss or malfunction of the following will have on the ADS:
Low pressure core spray system pressure REQUIRED MATERIALS:
Notes and Comments Prepared By:
Philip E. Nielsen (717) 456-4122 PBAPS Regulatory Exam Author Page 82 of 150
Peach Bottom February 2005 Initial License operator NRC Examination
- 42.
Unit 2 was operating at 75% power with Torus Cooling in service to support HPCI testing when the following transient occurred:
0 BLOWDOWN TIMERS INITIATED (227 D-4) alarms.
0 The SRO and the PRO have verified that a malfunction of the ADS Automatic Mode exists as confirmed by two independent indications and Adequate Core Cooling is assured.
0 The SRO then directs the PRO to inhibit ADS.
As a MINIMUM, the PRO must rotate (1)
ADS Inhibit Keylock Switch(es) to the INHIBIT position before the automatic depressurization initiates in approximately (2) seconds after the timers initiated alarm is received.
A. (1) EITHER (2) 100 B. (1) BOTH (2) 100 C. (1) EITHER (2) 640 D. (1)BOTH (2) 640
NRC Question Data Sheet
~
Psychometrics Level of Knowledge Difficulty Time Allowance (minutes)
N/A Answer Key Source:
Reference(s):
Learning Objective:
Question ID# 042 Both RO/SRO New Exam Item 0
Modified Bank Item 0
ILT Exam Bank 0
Previous NRC Exam 0 Other Exam Bank RRC 1G1-3 PLOT-5001 G.4a Choice Correct:
KnowledgelAbility
. WA Distractors:
21 8000 A4.02 Importance: RO / SRO ADS 4.2 / 4.2 B
A C
D Basis or Justification Both keylock inhibit switches are needed to prevent the auto depress. The auto depress would start approximately 100 seconds after the timers initiated alarm came in.
Rotating only one keylock inhibit switches would NOT prevent the auto depress. The auto depress would start approximately 100 seconds after the timers initiated alarm came in.
Rotating only one keylock inhibit switches would NOT prevent the auto depress. 640 seconds is a good distractor because the plant conditions demonstrate that a 2# Drywell pressure signal does NOT exist, but the key is that the timers initiated alarm does not come in until this requirement has been met or bypassed by the 9 minute timer.
Both keylock inhibit switches are needed to prevent the auto depress. 640 seconds is a good distractor because the plant conditions demonstrate that a 2# Drywell pressure signal does NOT exist, but the key is that the timers initiated alarm does not come in until this requirement has been met or bypassed by the 9 minute timer.
I ADS logic initiation REQUIRED MATERIALS:
Notes and Comments Prepared By:
Philip E. Nielsen (717) 456-4122 PBAPS Regulatory Exam Author Page 84 of 150
~
Peach Bottom February 2005 Initial License Operator NRC Examination
- 43.
Unit 2 was operating at full power when an Equipment Operator (EO) reported that he could smell burning insulation from Vital AC Electrical Panel 20Y33 and it was extremely hot to the touch. The SRO directed the Equipment Operator to deenergize 20Y33.
Use your evaluation of this situation to complete the following sentences identifying the response of the PCIS system and required procedural actions.
The Reactor Water Clean Up (RWCU) system will have received a isolation. Veri@ the isolations using (1)
(2)
A. (1) PARTIAL (2) GP-8.C, Group I, 11, and 111 INBOARD Half Isolation B. (1) PARTIAL (2) GP-8.D, Group I, 11, and I11 OUTBOARD Half Isolation
- c. (1) FULL (2) GP-8.C, Group I, 11, and I11 INBOARD Half Isolation D. (1) FULL (2) GP-8.D, Group I, 11, and I11 OUTBOARD Half Isolation
NRC Question Data Sheet Correct:
Answer Key C
Choice Distractors:
A Level of Knowledge LOW B
Difficulty Time Allowance (minutes)
SRO 3.75 3
NIA D
Knowledge/Ability KIA SRO Basis or Justification 223002 A2.01 Importance:
RO I SRO PClS 3.2 13.5 Although the loss of 20Y33 causes an INBOARD group II isolation, RWCU will receive a full isolation due to a logic power supply intertie.
Although it is an INBOARD isolation, loss of 20Y33 causes a full RWCU isolation. This is due to a logic power supply intertie.
Although it is an INBOARD isolation, loss of 20Y33 causes a full RWCU isolation. This is due to a logic power supply intertie.
Although RWCU will receive a full isolation, the loss of 20Y33 actually causes an INBOARD half isolation.
Source Documentation Source:
Reference( s) :
Learning Objective:
New Exam Item 0
Modified Bank Item 0
ILT Exam Bank Previous NRC Exam 0
Other Exam Bank GP-8C, Groups I, II, and Ill Inboard Half Isolation and associated Check-Off List PLOT-5007G.5a REQUIRED MATERIALS:
Notes and Comments Prepared By:
Philip E. Nielsen (717) 456-4122 PBAPS Regulatory Exam Author Page 86 of 150
Peach Bottom February 2005 Initial License Operator NRC Examination
- 44.
During a transient condition, the SRO has directed operation of Safety Relief Valves (SRVs) from the Alternative Shutdown Panel in the Recirc MG Set Room.
The (1) indication available comes from the SRV SRVs can be operated from this location and the position (2)
A. (1) A, B, and K (2) acoustic monitoring B. (1) A, B, and K (2) solenoid valve status C. (1) H, E, and L (2) acoustic monitoring D. (1) H, E, and L (2) solenoid valve status
NRC Question Data Sheet Level of Knowledge Difficulty Time Allowance (minutes)
LOW 3.25 3
Answer Key Question ID# 044 Both ROISRO SRO NIA Choice Distractors:
Source:
Reference (s) :
0 bjective:
KnowledgeIAbility KIA Learning I
[XI New Exam Item 0
Modified Bank Item 0
ILT Exam Bank 0 Previous NRC Exam 0 Other Exam Bank SE-10 Plant Shutdown from the Alternative Shutdown Panels PLOT-5001A 239002 K4.05 Importance:
C Basis or Justification The A, B, and K SRVs can be operated from the Alternative Control Station. Position indication is only by solenoid valve status.
This is incorrect because position indication is not from acoustic monitoring (as it is on the Remote Shutdown Panel), but from solenoid valve status.
~~
~~
~
~
~
This is incorrect because the H,E, and L SRVs are operated from the Remote Shutdown Panel, not the Alternative Shutdown Panel.
This is incorrect because the H, E, and L SRVs are operated from the Remote Shutdown Panel, not the Alternative Shutdown Panel.
REQUIRED MATERIALS:
Notes and Comments Prepared By:
Philip E. Nielsen (71 7) 4 5 6 4 22 PBAPS Regulatory Exam Author Page 88 of 150
Peach Bottom February 2005 Initial License Operator NRC Examination
- 45.
Unit 2 was operating at full power when the following conditions occurred:
0 0
0 0
The A Steam Line Flow Transmitter (DPT 2-6-5 1A) failed irrationally.
Digital Feedwater Control automatically selected single level control and RPV water level is stable.
After investigating the error, I&C reported that the steam flow transmitter will be unavailable until the next day when the required repair parts will anive.
A clearance remains in effect for the A Steam Flow Transmitter.
Under these conditions, the operator selects the LSF pushbutton and then the X pushbutton to place the Digital Feedwater Control System back into three element control.
The Digital Feedwater System will (1) and Reactor Feedwater Pump Turbine SPEED will (2)
A. (1) stay in single element control (2) remain steady B. (1) shift to three-element control (2) remain steady C. (1) shift to three-element control (2) rise until level stabilizes at a new higher level D. (1) shift to three-element control (2) lower until level stabilizes at a new lower level
NRC Question Data Sheet A
B C
D Answer Key Question ID# 045 Both RO/SRO Basis or Justification The Digital Feedwater Fault-Tolerance Override Logic will not permit the system to be placed in three-element control with an irrational steam flow detector. The system would permit three-element with a failed level detector but not a flow detector.
The candidate may expect the system to permit the operator to force the swap to three-element, because it does allow some things to be forced.
The system conditions would not change if the candidate believes the failed instrument is bypassed by the system when it fails irrationally.
The candidate may expect the system to permit the operator to force the swap to three-element, because it does allow some things to be forced.
The candidate may believe that with a failed steam flow instrument the system would be biased to allow level to rise to a new higher level causing RFP Turbine speed to rise.
The candidate may expect the system to permit the operator to force the swap to three-element, because it does allow some things to be forced.
The candidate may believe that with a failed steam flow instrument the system would be biased to allow level to lower to a new lower level causing RFP Turbine speed to lower.
Correct:
Level of Knowledge Difficulty Time Allowance (minutes)
LOW 3.0 3
1 Distractors:
SRO N/A Source:
Reference@):
Learning Objective:
Knowledge/Ability WA New Exam Item 0
Modified Bank Item 0
ILT Exam Bank Previous NRC Exam 0
Other Exam Bank SO 6C.l.D-2 Reactor Feedwater Automatic Level Control, Section 4.7 PLOT-5006.4n, 4r 259002 Al.07 Importance:
RO / SRO Reactor Water Level Control 2.6 / 2.6 Prepared By:
Philip E. Nielsen (717) 456-4122 PBAPS Regulatory Exam Author Page 90 of 150
Peach Bottom February 2005 Initial License Operator NRC Examination
- 46.
Unit 3 is operating at full power when it experiences a small steam leak that cannot be isolated in the Reactor Building. The following conditions are present:
0 0
0 A Group III isolation occurred as expected SBGT Filter Train Differential Pressure (a) is 2 W.G.
SBGT Filter Train Flow is 5500 c h Based on these conditions, use the SO 9A. 1.C, Figure 1 curve to determine the Standby Gas Treatment (SBGT) status and required actions.
The Standby Gas Treatment Filter Train is operating (1) the Expected (2)
Performance Region and the system must be 1
Diagnose A. (1) INSIDE (2) maintained in service B. (1) OUTSIDE (2) maintained in service C. (1) INSIDE (2) shutdown if dP drops below 1.8 W.G.
D. (1) OUTSIDE (2) shutdown immediately and declared inoperable
NRC Question Data Sheet B
A C
D Answer Key Question ID# 046 Both RO/SRO I
Plotting the point on figure 1 shows the system to be operating OUTSIDE of the Expected Performance Region. The system must be maintained in service due to plant conditions and the fact that a NOTE in SO 9A.1.C states that Figure 1 is for information only and is not and indication of operability status.
A majority of the curves used in our procedures require operation above the curve. This could lead a candidate to believing that the system is operating INSIDE the Expected Performance Region while it is above the curve.
If the candidate interprets the system performance to be INSIDE the expected region, dropping dP to e l.8" W.G. would put the system operation outside of the Expected Performance Region.
Plotting the point on figure 1 shows the system to be operating OUTSIDE of the Expected Performance Region, however, the system is considered to be inoperable based on a NOTE in SO 9A.l.C states that Figure 1 is for information only and is not and indication of operability status.
Choice
~
Level of Knowledge HIGH I
Basis or Justification Difficulty Time Allowance (minutes)
SRO 3.0 3
N Correct:
Source:
Reference(s):
Learning Objective:
KnowledgelAbility WA Distractors:
New Exam Item 0
Modified Bank Item 0
ILT Exam Bank 0
Previous NRC Exam 0 Other Exam Bank SO 9A.l.C procedure and Figure 1 Curve.
PLOT-5009A.3a 261 000 A3.01 Importance:
Psychometrics I
REQUIRED MATERIALS:
SO 9A.l.C Figure 1 Curve.
Notes and Comments Prepared By:
Philip E. Nielsen (717) 456-4122 PBAPS Regulatory Exam Author Page 92 of 150
~
Peach Bottom February 2005 Initial License Operator NRC Examination
- 47.
Given the following conditions:
0 0
Both units are operating at full power Both units electrical loads are aligned in the preferred line-up With all busses energized.
Under these conditions, the Plant Reactor Operator (PRO) notes the following:
0 0
0 E-212 breaker GREEN and RED indicating lights are NOT lit.
E-12 Bus Voltage is normal E-3 12 breaker GREEN indicating light is LIT.
El Emergency Diesel Generator is in a normal standby alignment The implication of these 4KV indications is that the E-212 breaker:
A. is racked out.
B. trip coil has lost continuity.
C. has received an automatic trip signal.
D. Alternative Shutdown Station control is in Emergency.
NRC Question Data Sheet Level of Knowledge HIGH Answer Key Question ID# 047 Both RO/SRO Difficulty Time Allowance (minutes)
SRO 2.5 3
N/A Correct:
Distractors:
Knowledge/Ability KIA I
262001 K5.02 Importance:
RO / SRO AC Electrical Distribution 2.6 12.9 B
A C
D Basis or Justification A loss of trip coil continuity will result in losing the Red light indication on a closed breaker. This relates to PBAPS Upset Report 2-88-01 E-12 Breaker Trip Precluded where a similar event occurred.
The breaker being racked out would cause this red and green light indication, but the candidate must recognize that with the bus at normal voltage and the E-12 & E-312 open, the E-212 must be closed.
The candidate must recognize that an automatic trip signal would cause the red light to go out, but would also the green light to light and the bus to fast transfer or deenergize.
Transferring an ASD related breaker to emergency would cause these indications, however, the candidate must recognize that the E-212 breaker is not ASD equipped.
I Psvchometrics I
Source:
Reference(s):
Learning Objective:
~
~
Source Documentation New Exam Item 0 Modified Bank Item 0 ILT Exam Bank 0
Previous NRC Exam 0 Other Exam Bank so 54.7.c PLOT-5054.56 REQUIRED MATERIALS:
Notes and Comments Prepared By:
Philip E. Nielsen (717) 456-4122 PBAPS Regulatory Exam Author Page 94 of 150
~-
Peach Bottom February 2005 Initial License Operator NRC Examination
- 48.
Unit 3 was operating at full power with the following conditions present:
0 0
E-134-T-B is deenergized 0
0 0
0 A RPS bus is powered by its Alternate Power Supply due to a failure of the 3A RPS MG Set.
INVERTER TROUBLE (320 F-5) alarms An EO investigating the condition reports that the Load on Inverter light is lit on the Static Inverter and all system voltages appear to be normal.
The EO is then directed to transfer the Static Inverter to its alternate power The EO immediately placed the Manual Bypass Switch (MBS) on the Static Inverter in the Load to Bypass position.
supply-Under these conditions, power to the Power Range Neutron Monitoring system will:
A. be unaffected because the MBS will not transfer load to a dead bus.
B. lose power to 1 and 3 APRMs, causing a full scram signal.
C. lose power to 1 and 3 2/4 logic modules, causing a half scram signal.
D. lose power to 1 and 3 2/4 logic modules, causing a full scram signal.
NRC Question Data Sheet Level of Knowledge Difficulty HIGH 3.5 I
Answer Kev Time Allowance (minutes)
SRO 4
NIA Choice Correct: I-262002 K1.19 UPS Distractors:
Importance: RO I SRO 2.9 13.1 C
A Basis or Justification Loss of power to the '1" and '3" 2/4 logic modules will result in a half scram signal due to trips in the A1 and A2 Logic Channels. This answer is correct, but may not be comfortable to the candidates because a rule of thumb is that we no longer get half scrams from the APRMs (as we used to with the old GE system).
The "Load to Bypass" push button will not transfer to a dead bus, however the MBS will transfer to a dead bus resulting in a loss of power from to Y-50 and the RPS Alternate power supply.
The APRMs are powered from both the "A" and "B" RPS busses, so the APRMs do not lose power under these conditions. This is a change with the PRNM system compared to the old GE system.
Loss of power to the "1" and "3" 2/4 logic modules will NOT result in a full scram because they both input to the "A" RPS (Channels A1 and A2)
Source:
Reference@):
Learning 0 bjective:
KnowledgelAbility KIA Source Documentation New Exam Item 0 Modified Bank Item ILT Exam Bank 0
Previous NRC Exam 0 Other Exam Bank SO 58B.7.A-3 PLOT-5058.1 r (Description of K&A, from catalog)
Knowledge of the physical connections andlor cause-effect relationships between UPS and the following:
Power range neutron monitoring system REQUIRED MATERIALS:
Notes and Comments Prepared By:
Philip E. Nielsen (717) 456-4122 PBAPS Regulatory Exam Author Page 96 of 150
Peach Bottom February 2005 Initial License Opemitor NRC Examination
- 49.
Unit 2 was operating at full power when the following occurs:
0 2AD004/2BD004 BATTERY GROUND (220 H-5) alarms An EO sent to investigate reports that the Ground Lamp Indications are as follows:
+ Ground Lamp A is BRIGHTLY lit
+ Ground Lamps B are OUT
+ The Ground Detection Ammeter is reading mid-scale.
0 Annunciator 220 H-5 does not clear when the EO depresses the local reset.
0 The crew begins to search for the ground by isolating loads in accordance with A 0 57A.1-2, 125/250 VDC Balance of Plant Station Battery Ground Investigation.
When the grounded load is isolated, all of the Ground Lamp Indications will be (1) lit and the ground detection ammeter will approach (2)
A. (1) brightly (2) zero B. (1) brightly (2) full scale C. (1)dimly (2) zero D. (1) dimly (2) full scale
NRC Question Data Sheet Distractors:
Answer Key Question ID# 049 Both RO/SRO A
B Choice Correct:
Level of Knowledge D ifficu I ty Time Allowance (minutes)
LOW 2.75 3
C SRO N/A Knowledge/Abi lity WA D
263000 A3.0 1 Importance: RO / SRO DC Electrical Distribution 3.2 13.3 Basis or Justification A 0 57A.1-2 notes state that the Ground Lamp indications are dimly lit and the Ammeter will read near zero when the ground is isolated.
Although the ammeter should approach zero, Ground Lamp indications that are brightly lit indicate a positive ground.
Ground Lamp indications that are brightly lit indicate a positive ground and a full scale ammeter indicates a significant ground is present.
Although the Ground Light Indications should be dimly lit, a full scale ammeter reading indicates a significant ground is present.
I Source:
Reference(s):
Learning 0 bj ect ive :
Source Documentation 0 New Exam Item 0 Modified Bank Item ILT Exam Bank Previous NRC Exam 0801#221 0 Other Exam Bank A 0 57A. 1-2, I 29250 VDC Balance of Plant Station Battery Ground Investigation ARC 220 H-5,2AD004/2BD004 BAlTERY GROUND PLOT-5057.1 d REQUIRED MATERIALS:
Notes and Comments Prepared By:
Philip E. Nielsen (717) 456-4122 PBAPS Regulatory Exam Author Page 98 of 150
Peach Bottom February 2005 Initial License Operator NRC Examination
- 50.
Peach Bottom experienced a complete loss of off-site power. The following conditions are present:
0 0
0 0
The E-1 Emergency Diesel Generator (EDG) failed to start The other EDGs started and loaded their busses normally No backfeed operations have been completed for the E-1 supplied busses Unit 2 has experienced a complete loss of high pressure feed and RPV water level is dropping continually.
Based on these conditions, which Core Spray Pump(s) will be available to line up and inject to the Unit 2 reactor vessel when the reactor is depressurized.
The Core Spray Pumps AVAILABLE TO INJECT will be:
A. A and C ONLY B. B and D ONLY.
C. B, C, and D ONLY.
NRC Question Data Sheet Level of Knowledge Difficulty HIGH 2.75 Answer Key Question ID# 050 Both RO/SRO Time Allowance (minutes)
SRO 4
NIA I
Choice 264000 K3.01 EDGs Correct:
Importance:
RO / SRO 4.2 14.4 Distractors:
B A
C D
Basis or Justification Only the B Loop of Core Spray will be available because although only the A Pump has lost power, the Inboard Injection Valve will also be closed and deenergized due to the E-1 loss.
The A Loop of Core Spray is notavailable because the Inboard Injection Valve is closed and deenergized due to the E-1 loss.
This distractor is likely to be selected because candidates may not realize that the A Loop Inboard Injection Valve will be closed and deenergized due to the E-I loss.
The candidates may believe that the Core Spray System would have additional redundant backups making it available much like the RHR Injection Valves are powered by a swing bus so they dont become unavailable.
Source: r Reference(s):
Learning Objective:
Knowledge/Ability WA Source Documentation New Exam Item 0
Modified Bank Item ILT Exam Bank 0
Previous NRC Exam 0
Other Exam Bank SO 14.1.A-2 COL, Core Spray Check Off List Item 162 PLOT-5052.3a (Description of K&A, from catalog)
Knowledge of the effect that a loss or malfunction of the EDG system will have on the following:
Emergency Core Cooling Systems REQUIRED MATERIALS:
Notes and Comments Prepared By:
Philip E. Nielsen (717) 4564122 PBAPS Regulatory Exam Author Page 100 of 150
Peach Bottom February 2005 Initial License Operator NRC Examination 5 1.
Unit 2 was operating at full power when the following occurred:
0 0
0 0
The E-22 4KV Bus normal off-site feed breaker tripped The E-22 4KV Bus alternate off-site feed breaker failed to close in The E-2 Emergency Diesel Generator (EDG) started automatically and supplied the E-22 Bus with power.
The failures on both off-site feeder breakers have now been corrected.
E-2 Emergency Diesel Generator is currently operating in the synchronization (2)
(1)
Mode and when the DG Auto Start Bypass Pushbutton is depressed, the A. (1) Droop (Parallel)
(2) must be completed in 3 minutes or the EDG will return to the original mode B. (1) Droop (Parallel)
(2) does NOT need to be completed in 3 minutes because the EDG will NOT return to its original mode C. (1) Isochronous (Unit)
(2) must be completed in 3 minutes or the EDG will return to the original mode D. (1) Isochronous (Unit)
(2) does NOT need to be completed in 3 minutes because the EDG will NOT return to its original mode
NRC Question Data Sheet Level of Knowledge Difficulty Time Allowance (minutes)
HIGH 3.75 4
Answer Key SRO NIA Question ID# 051 Both RO/SRO Source:
Reference@):
Choice 0
New Exam Item Modified Bank Item ILT Exam Bank 0
Previous NRC Exam 0
Other Exam Bank SO 52B.2.A, Diesel Generator Shutdown; Diesel Carrying One 4KV Emergency Correct:
KnowledgeIAbility WA Distractors:
264000 K5.05 Importance:
A B
C Basis or Justification On a dead bus start the EDG operates in the Isochronous Mode. After depressing the DG Auto Start Bypass pushbutton, it will not return to the original mode after 3 minutes because no MCA signal is present.
With no MCA signal present, the EDG will not retum to its original mode 3 minutes after the DG Auto Start Bypass Pushbutton is depressed. On a dead bus start the EDG does operate in the Isochronous Mode.
Although, with no MCA signal present, the EDG will not return to its original mode 3 minutes after the DG Auto Start Bypass Pushbutton is depressed, on a dead bus start the EDG operates in the Isochronous Mode.
On a dead bus start the EDG does operate in the Isochronous Mode, but, after depressing the DG Auto Start Bypass pushbutton, it will not retum to the original mode after 3 minutes because no MCA signal is present.
Learning 0 bjective:
PLOT-5052.5e Page 102 of 150
Peach Bottom February 2005 Initial License Operator NRC Examination
- 52.
Unit 2 is experiencing a Loss of Instrument Air transient.
As Instrument Air Pressure LOWERS fiom its normal value, the MAXIMUM pressure when the Backup Air Compressor will automatically start is (1) and the MAXIMUM pressure when the Backup Air Control Valve (AO-80250D) will automatically open is (2)
A. (1) 90 psig (2) 90 psig B. (1) 90 psig (2) 100 psig C. (1) 100psig (2) 90 psig D. (1) 1OOpsig (2) 1OOpsig
NRC Question Data Sheet Level of Knowledge LOW Answer Key Question ID## 052 Both R(
Diffi cu Ity Time Allowance (minutes)
SRO 2.0 3
N/A r-Choice 300000 K4.01 Instrument Air I
C Correct:
Importance: RO / SRO 2.0 / 2.9 Distractors :
A I
B I
Basis or Justification The Backup Air Compressor starts at 100 psig and the Backup Air Control Valve opens at 90 psig.
The Backup Air Compressor actually starts at 100 psig.
The Backup Air Compressor actually starts at 100 psig and the Backup Air Control Valve actually opens at 90 psig.
The Backup Air Control Valve doesnt actually open until 90 psig.
I Psvchometrics I
Source:
Reference(s):
Learning Objective:
Knowledge/Ability KIA Source Documentation New Exam Item 0 Modified Bank Item 0 ILT Exam Bank 0 Previous NRC Exam 0 Other Exam Bank ON-119, Loss of Instrument Air PLOT-5036.4a (Description of KW, from catalog)
Knowledge of the Instrument Air system design feature(s) and/or interlocks which provide for the following:
0 ManuaVautornatic transfer of control REQUIRED MATERIALS:
Notes and Comments Prepared By:
Philip E. Nielsen (717) 456-4122 PBAPS Regulatory Exam Author Page 104 of 150
~
Peach Bottom February 2005 Initial License Operator NRC Examination
- 53.
Given the following conditions:
0 Unit 3 is operating in MODE 1 at 25% power.
0 The 3A Turbine Building Closed Cooling Water (TBCCW) Pump control switch is in OFF due to the discharge valve being stuck closed.
0 The 3B TBCCW Pump tripped on overcurrent and could not be restarted.
0 All other systems responded as designed.
In accordance with ON-1 18, Loss of TBCCW, which of the following components will need to be secured due to rising temperatures to prevent damage during continued plant operation.
A. Condensate Pumps B. Isophase Bus Coolers C. Control Rod Drive Pumps D. Instrument Air Compressors
~~
NRC Question Data Sheet Level of Knowledge Difficulty Time Allowance (minutes)
LOW 2.25 3
Answer Key Question ID# 053 Both RO/SRO SRO NIA I
Choice Source :
Reference(s):
Objective:
Lea mi ng I
Correct:
New Exam Item 0
Modified Bank Item 0
ILT Exam Bank 0
Previous NRC Exam Other Exam Bank ON-118, Loss of TBCCW Bases PLOT-5034.3 Distractors:
Knowledge/Ability WA A
400000 K3.01 Importance:
RO I SRO Component Cooling Water 2.9 13.3 C
D Basis or Justification The Condensate Pumps do not have a backup cooling supply because they are not considered vital to plant safety.
The lsophase Bus Coolers do not have a backup cooling supply, but the ON-118 bases states that if generator load is 48,000 amps, the environment can absorb the heat generated by the bus bars.
Cooling to CRD Pumps is backed up by the RBCCW system.
Cooling to Instrument Air Compressors is backed up by the RBCCW system.
REQUIRED MATERIALS:
Notes and Comments Prepared By:
Philip E. Nielsen (717) 456-4122 PBAPS Regulatory Exam Author Page 106 of 150
-~
Peach Bottom February 2005 Initial License Operator NRC Examination
- 54.
A Unit 2 reactor startup is in progress with control rod withdrawals occurring.
0 0
0 0
0 Rod Worth Minimizer (RWM) Group 1 contains 12 control rods that are to be drawn from notch 00 to notch 48.
10 control rods from this group have been withdrawn to notch 48 A pen-and-ink change is made to the Approved Startup Sequence to skip the last 2 rods in RWM Group 1.
The RWM computer sequence is not changed.
The Reactor Operator attempts to select the first rod in R W M Group 2.
Which of the following describes the response of the RWM to these conditions?
The RWM will:
A. insert a select block to prevent the selection of the Group 2 rod.
B. latch into Group 2, resulting in 2 insert errors and an insert rod block.
C. display a select error, but NO insert errors, since it is still latched into Group 1.
D. NOT latch into Group 2 until the first Group 2 rod is moved. Then it will display 2 insert errors and an insert rod block.
NRC Question Data Sheet Correct:
Answer Kev C
Question ID# 054 Both RO/SRO Level of Knowledge HIGH Choice Difficulty Time Allowance (minutes)
SRO 3.5 4
N/A Source :
Reference&):
Objective:
Knowledge/Ability WA Learning Distractors: 2 0
New Exam Item 0
Modified Bank Item 0
ILT Exam Bank 0
Previous NRC Exam Other Exam Bank - LORT SO 62.7.A-2, Receipt of Rod Blocks PLOT-5062A.5 201 006 A I.03 Importance: RO / SRO RWM 2.9 13.0 B
D Basis or Justification The RWM will stay latched to Group 1 because latching to Group 2 would result in 2 insert errors, which the system will not latch into automatically.
The system will have a select error for the selection of a rod outside of the latched group, but will not give a select block.
The RWM NOT latch into Group 2 because this would result in 2 insert errors, which the system will not latch into automatically.
If rod motion were attempted, it would result in a withdrawal error and rod block.
(Description of K&A, from catalog)
Ability to predict and/or monitor changes in parameters associated with operating the RWM system controls including:
Latched group indications REQUIRED MATERIALS:
Notes and Comments Prepared By:
Philip E. Nielsen (717) 456-4122 PBAPS Regulatory Exam Author Page 108 of 150
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Peach Bottom February 2005 Initial License Operator NRC Examination
- 55.
Unit 2 was operating at 60% power when the following transient occurred:
0 The SRO directed a GP-4, Manual Reactor Scram.
0 Reactor Level dropped to +2 inches before turning and beginning to rise.
0 The Reactor Operator tripped the Reactor Feed Pumps when he saw level rising.
Evaluate these plant conditions to determine the most limiting recirculation system response and the reason for that response.
The Recirculation Pumps will receive a:
A. 30% runback to ensure adequate Reactor Feedwater Flow is available.
B. 30% runback to ensure adequate Recirc Pump Net Positive Suction Head.
C. 45% runback to ensure adequate Reactor Feedwater Flow is available.
D. 45% runback to ensure adequate Recirc Pump Net Positive Suction Head.
NRC Question Data Sheet Distractors:
Answer Key Question ID# 055 Both RO/SRO A
Although a 30% runback will occur its reason is to provide adequate Recirc Pump NPSH.
I Choice I
Basis or Justification I
C Correct:
~~
Although a 45% runback will also be received, the 30% runback is more limiting and is actually based on providing adequate Recirc Pump NPSH.
~
With a reactor scram and total feedwater flow < 20%, a 30% runback will I B l occur to ensure adequate Recirc Pump Net Positive Head.
LOW 2.5 4
NIA Source:
Reference@):
0 bject ive :
Knowledge/Ability WA Learning Although a 45% runback will also be received, the 30% runback is more I D I limiting.
New Exam Item 0
Modified Bank Item 0 ILT Exam Bank 0
Previous NRC Exam Other Exam Bank OT-100, Reactor Low Water Level PLOT-5002.04b 202002 K4.06 Importance:
RO I SRO Recirculation Flow Control 3.1 13.1 Psychometrics Level of Knowledae 1
D i ffi cu I tv I Time Allowance (minutes) I SRO REQUIRED MATERIALS:
Notes and Comments Prepared By:
Philip E. Nielsen (717) 456-4122 PBAPS Regulatory Exam Author Page 110 of 150
Peach Bottom February 2005 Initial License Operator NRC Examination
- 56.
Given the following Unit 2 Reactor Water Cleanup (RWCU) System conditions:
0 RWCU Pump Motor Winding Temperature is 150°F.
0 Pump Flow at 65 gpm for 20 seconds.
0 NRHX Outlet Temperature is 180°F.
0 Reactor Water Level is 16 inches.
Select the following statement that describes the Reactor Water Cleanup response and reason for the response to these conditions.
The Reactor Water Cleanup:
A. Pump will trip on low pump flow.
B. Pump will trip due to high motor winding temperature.
C. System will isolate due to low reactor water level.
D. System will isolate due to high NRHX Outlet temperature.
NRC Question Data Sheet Level of Knowledge D ifficu I ty LOW 3.0 Answer Key Question ID# 056 Both ROISRO Time Allowance (minutes)
SRO 3
NIA I
Choice Knowledge/Ability WA Correct:
Distractors:
204000 A4.01 Importance:
C D
Basis or Justification I
The RWCU pump will trip when motor winding temperature exceeds 149°F.
When RWCU pump flow drops to <70 gpm, an alarm is received, but the pump trip no longer occurs (it did in the past).
Although 16 inches reactor water level will result in a low level alarm, the RWCU system will not isolate until level drops to 1 inch.
180°F is an elevated NRHX Outlet Temperature, but the isolation does not occur until temperature reaches 200°F.
Source Documentation Source:
Reference( s):
Learning Objective:
Ix] New Exam Item 0
Modified Bank Item 0
ILT Exam Bank 0
Previous NRC Exam 0
Other Exam Bank ARC 215 A-2, A Clean-up Recirc Pump Mtr Wdg Temp High-High PLOT-5012 REQUIRED MATERIALS:
Notes and Comments Prepared By:
Philip E. Nielsen (71 7) 456-41 22 PBAPS Regulatory Exam Author Page 112 of 150
Peach Bottom February 2005 Initial License Operator NRC Examination
- 57.
A plant start up is in progress on Unit 2 with the following conditions present:
0 Reactor Power is 14%.
0 0
0 Rod Withdrawal is in progress.
A control rod is being withdrawn using single notches to a target position of 12.
When the rod reaches position 8, rod indication on both the full core and the four rod display go blank due to a position indicating reed switch failure.
This same reed switch was failed during a startup earlier this year.
Evaluate these conditions to determine:
(1) the plant response, and (2) the procedural actions that are required to continue the startup.
A. (1) A Rod Withdraw Block (21 1 D-3) ONLY (2) Bypass the RWM using A 0 62A. 1-2, Rod Worth Minimizer System Manual Bypass.
B. (1) A Rod Drift Alarm (21 1 D-4) ONLY (2) Withdraw the rod to the next position using SO 62.1-2, Withdrawinghserting a Control Rod One Notch.
C. (1) A Rod Drift Alarm (21 1 D-4) and a Rod Withdraw Block (21 1 D-3)
(2) Insert the rod to Full-In using the EMER IN control switch using ON-121, Drifting Control Rod.
D. (1) A Rod Drift Alarm (21 1 D-4) and a Rod Withdraw Block (21 1 D-3)
(2) Insert a substitute rod position using A 0 59A.2-2, Adding and Removing Substitute Control Rod Positions.
NRC Question Data Sheet Level of Knowledge Difficulty Time Allowance (minutes)
HIGH 3.25 4
Answer Key Question ID# 057 Both RO/SRO SRO NIA Choice Correct:
Distractors:
214000 A2.01 RPlS D
A Importance:
C Basis or Justification The Rod Drift Alarm will be the result of the loss of the even reed switch.
The Rod Withdraw Block will be from the RWM when it loses rod indication.
A0 62C.1-2, Rod Position Determination due to RPlS Failure, directs inserting a substitute control rod position using A 0 59A.2-2.
A Rod Drift alarm will also be received.
Bypassing the RWM is NOT directed in this situation where one reed switch has failed. The procedures clearly lead you to entering a substitute position to clear the condition.
A Rod Withdraw Block will also be received from the RWM.
Control Rod Withdrawal with the SO is not possible at this time.
No indication of rod motion is indicated. Even if the crew entered ON-121 for the rod drift, it would send them to A 0 62C.1-2 rather than direct insertion of the control rod.
Source:
Reference( s):
Learning Objective:
KnowledgelAbility WA Source Documentation New Exam Item 0
Modified Bank Item 0
ILT Exam Bank ARC 211 D-4, Rod Drift ARC 21 1 D-3, Rod Withdraw Block A0 62C.1-2, Rod Position Determination due to RPlS Failure.
A0 59A.2-2 Adding and Removing Substitute Control Rod Positions 0 Previous NRC Exam 0
Other Exam Bank PLOT-5003.10 (Description of K&A, from catalog)
Ability to (a) predict the impact of the following on the RPIS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:
Failed reed switches
,epared By:
Philip E. Nielsen (717) 456-4122 PBAPS Regulatory Exam Author Page 114 of 150
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Peach Bottom February 2005 Initial License Operator NRC Examination
- 58.
Given the following conditions:
0 0
0 0
0 Both Units have been shutdown due to steam leaks in the Drywell LOCA signals sealed in to both units 15 minutes ago.
With normal level, low pressure ECCS pumps have been secured.
The Electrical Plant is in its normal alignment.
- 3 Emergency Aux. Bus is more heavily loaded than #2 Emergency Am. Bus.
The CRS directs the PRO to place Torus Cooling in service on Unit 3 using an RHR pump powered from #2 Emergency Aw. Bus.
Select the RHR Pump to be used for Torus Cooling on Unit 3.
A. 3ARHR B. 3BRHR C. 3CRHR D. 3DRHR
NRC Question Data Sheet Choice Answer Key Basis or Justification Correct:
D
'3D" RHR Pump is available during the duel unit LOCA and it is powered from #2 Emergency Aux. Bus.
Distractors:
A
'3A" RHR Pump is NOT available during a duel unit LOCA B
'38" RHR Pump is NOT available during a duel unit LOCA C
'3C" RHR Pump is available during the duel unit LOCA, but it is powered from #3 Emergency Aux. Bus.
Pumps Level of Knowledge Difficulty Time Allowance (minutes)
HIGH 3.25 4
REQUIRED MATERIALS:
SRO NIA Notes and Comments Source:
Reference(s):
Learning 0 bjective:
Prepared By:
Philip E. Nielsen (717) 456-4122 PBAPS Regulatory Exam Author New Exam Item 0 Modified Bank Item 0 ILT Exam Bank SO 10.1.D-3 Torus Cooling SO 10.7.8-3 RHR System Automatic Response during LOCA PLOT-5010.1 g,2a 0
Previous NRC Exam 0
Other Exam Bank Page 1 16 of 150 Knowledge/Ability K/A 21 9000 K2.02 Importance: RO / SRO RHWLPCI: Torus/Pool Cooling Mode 3.1 / 3.3
Peach Bottom February 2005 Initial License Operator NRC Examination
- 59.
Given the following:
0 Unit 2 is operating at full power 0
Torus pressure is currently 0.1 psid lower than Reactor Building pressure.
Loss of all pneumatic supplies to the Reactor Building to Torus air-operated butterfly valve (AO-2502A) will cause the valve to fail:
A. OPEN and WILL result in de-inerting the Primary Containment Atmosphere.
B. OPEN, but will NOT result in de-inerting the Primary Containment Atmosphere.
C. CLOSED and WILL prevent pressure equalization fkom the Reactor Building to Torus.
D. CLOSED, but will NOT prevent pressure equalization from the Reactor Building to Torus.
NRC Question Data Sheet Correct:
I Answer Key B
I Question ID## 059 Both RO/SRO Level of Knowledge Difficulty Time Allowance (minutes)
LOW 3.5 3
I Choice SRO N/A Knowledge/Ability WA l
A Distractors :
223001 K6.13 Importance:
RO I SRO Primary Containment and Aux 3.2 13.4 I
C I
D Basis or Justification The A 0 valve fails open, but will not de-inert the Primary Containment because there is a self-operating check valve in series with the A 0 that does not open until 1 psid.
The A 0 valve fails open, but will not de-inert the Primary Containment because there is a self-operating check valve in series with the A0 that does not open until 1 psid.
The A0 valve fails open and does not prevent equalizing pressure.
The A 0 valve fails open and does not prevent equalizing pressure.
Source:
Reference( s):
Learning Objective:
Source Documentation 0
New Exam Item 0
Modified Bank Item ILT Exam Bank Previous NRC Exam 099W119 0 Other Exam Bank ON-I 19, Loss of Instrument Air PLOT-5007.69 1 Applicable plant air system/ nitrogen make-up system I
REQUIRED MATERIALS:
Notes and Comments Prepared By:
Philip E. Nielsen (717) 456-4122 PBAPS Regulatory Exam Author Page 118 of 150
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Peach Bottom February 2005 Initial License Operator NRC Examination
- 60.
Unit 2 experienced a small break LOCA Condition, the following conditions exist:
0 The Reactor has been manually shutdown.
0 Reactor Level has been maintained continuously +5 to +35 inches with feedwater.
0 Reactor Pressure is 400 psig with a depressurization in progress.
0 Torus Cooling is in service on the B RHR Loop using the D RHR Pump.
0 Drywell Pressure is 4 psig.
0 Drywell Temperature is 2 10°F.
The SRO has directed that the PRO place Torus Sprays in service using the B RHR Loop.
Based on these conditions, what action(s) is (are) required for the spray logic to permit the B Torus Spray Valve (MO-2-10-34B) to be throttled opened?
A. ONLY the Ctmt. Spray Valve Cont. (S 17B) must be placed momentarily in manual.
B. ONLY the Ctmt. Spray Vlv. Reset Pushbutton (S33B) must be momentarily depressed.
C. The Ctmt. Spray Override 2/3 Core Coverage Keylock Switch (S18B) must be placed in override AND the Ctmt. Spray Valve Cont. (S17B) must be placed momentarily in manual.
D. The Ctmt. Spray Override 2/3 Core Coverage Keylock Switch (S18B) must be placed in override AND the Containment Spray Valves Reset Pushbutton (S33B) must be momentarily depressed
NRC Question Data Sheet Correct:
Answer Key A
Question ID# 060 Both RO/SRO Level of Knowledge Difficulty Time Allowance (minutes)
HIGH 2.75 3
SRO N/A Distractors: 7 Source:
C New Exam Item 0
Modified Bank Item 0
ILT Exam Bank 0
Previous NRC Exam 0
Other Exam Bank D
Reference(s):
Learning Basis or Justification T-204-2, Initiation of Drywell Sprays with RHR PLOT301 0.1 u Only the use of the S17 switch is required since the conditions are met that the S18 overrides (LOCA signal sealed in and >223 inches reactor level)
The S33 pushbutton is used to restore the system after the LOCA signal is clear.
The S18 switch is not required because it is used to override a "Lack of LOCA Signal" or not having 2/3 core coverage. Since these conditions are met, the switch is not required (or procedurally directed).
The S18 switch is not required because it is used to override a "Lack of LOCA Signal" or not having 2/3 core coverage. Since these conditions are met, the switch is not required (or procedurally directed).
0 bjective:
I Knowledge/Ability 1 226001 A4.03 I Importance: RO I SRO I
(Description of K&A, from catalog)
Ability to manually operate and/or monitor in the control room:
Spray Valves REQUIRED MATERIALS:
Notes and Comments Prepared By:
Philip E. Nielsen (717) 456-4122 PBAPS Regulatory Exam Author Page 120 of 150
-~
Peach Bottom February 2005 Initial License Operator NRC Examination
- 61.
Unit 3 was operating at full power when the following transient occurred:
0 A loss of W2 13KV Auxiliary Bus resulted in rapidly lowering reactor level.
0 An electric ATWS occurred when a manual scram was attempted.
T-22 1, Main Steam Isolation Valve Bypass has been completed 0
Level has been lowered to a band of -172 inches to -140 inches for power control 0
Fuel damage from the ATWS transient has caused Main Steam Line Radiation to rise.
The Group 1 Isolation Setpoint for Main Steam Line Radiation is (1)
Normal Full Power Background (NFPB) and based on these conditions, the MSIVs (2) isolate when the radiation exceeds the setpoint.
A. (1) 140%
(2) WILL B. (1) 140%
(2) will NOT C. (1) 10Times (2) WILL D. (1) 10Times (2) will NOT
~~
~
NRC Question Data Sheet Level of Knowledge Difficulty Time Allowance (minutes)
HIGH 2.75 3
Answer Key Question ID# 061 Both RO/SRO SRO N/A Choice Distractors:
239001 A3.01 Main and Reheat Steam.
Importance:
D Basis or Justification The isolation setpoint is 1OX NFPB. The MSlVs will isolate because T-221 does not bypass the High MSL Radiation Isolation.
140% is the High Main Steam Flow setting, NOT the radiation setpoint.
These setpoints could be easily confused by a candidate.
140% is the High Main Steam Flow setting, NOT the radiation setpoint.
The MSlVs will isolate because T-221 does not bypass the High MSL Radiation Isolation. The candidate that does not understand the bypass will believe that it is always bypassed.
The isolation setpoint is 1OX NFPB, but the MSlVs WILL isolate because T-221 does not bypass the High MSL Radiation Isolation. The candidate that does not understand the bypass will believe that it is always bypassed.
Source:
Reference@):
Learning Objective:
Knowledge/Ability KIA Source Documentation New Exam Item Modified Bank Item ILT Exam Bank 0
Previous NRC Exam 0
Other Exam Bank T-221-2 Main Steam Isolation Valve Bypass PLOT-50079.4 h (Description of K&A, from catalog)
Ability to monitor automatic operations of the Main and Reheat Steam System including:
Isolation of the Main Steam System REQUIRED MATERIALS:
Notes and Comments Prepared By:
Philip E. Nielsen (717) 456-4122 PBAPS Regulatory Exam Author Page 122 of 150
Peach Bottom February 2005
~
Initial License Operator NRC Examination
- 62.
Unit 2 was operating at 90% power when the CRS directed that the EHC Pressure Regulator be swapped fiom A in Control to B in Control.
This task is accomplished by adjusting the Pressure Setpoint Bias Potentiometer located (1) and must be completed by (2)
A. (1) in the Cable Spreading Room (2) an I&C Technician B. (1) in the Cable Spreading Room (2) a Licensed Operator C. (1) at the Main Turbine Front Standard (2) an I&C Technician D. (1) at the Main Turbine Front Standard (2) a Licensed Operator
NRC Question Data Sheet LOW Answer Key 2.25 3
NIA I
Question ID# 062 Both RC Source:
Choice New Exam Item 0
Modified Bank Item 0 Previous NRC Exam 0
Other Exam Bank Correct:
Knowledge/Ability KIA Distractors:
241000 G2.1.30 Importance:
RO I SRO ReactorITurbine Pressure Regulator 3.9 13.4 B
A C
D SRO Basis or Justification The potentiometer is located on a panel in the Cable Spreading Room.
The task MUST be accomplished by a licensed operator because it can change pressure, which changes power and is therefore a reactivity manipulation.
The potentiometer is located in the Cable Spreading Room, but although I&C is used extensively during system troubleshooting, a licensed operator is required to shift potentiometers. This is an infrequent task and this is not obvious to the candidate.
Although some turbine controls are located at the front standard, the potentiometer is located in the Cable Spreading Room.
~~
~
~
Although some turbine controls are located at the front standard, the potentiometer is located in the Cable Spreading Room. Also, a licensed operator is required to perform the task due to its being a reactivity manipulation.
Psychometrics Level of Knowledge I Difficulty I Time Allowance (minutes) 1 SRO Reference@):
Learning Objective:
0 ILT Exam Bank A 0 1 D. 1-2 Swapping EHC Pressure Regulators PLOT-5001 DL.4a REQUIRED MATERIALS:
Notes and Comments Prepared By:
Philip E. Nielsen (717) 4564122 PBAPS Regulatory Exam Author Page 124 of 150
Peach Bottom February 2005 Initial License Operator NRC Examination
- 63.
When placing a Reactor Feedwater Pump in service in accordance with SO 6C.l.C-2, Startup of Second or Third Reactor Feedwater Pump, the operator is cautioned to ensure that the Reactor Feedwater Pump discharge pressure is greater than reactor pressure.
The intent of this precaution is to prevent (1)
The procedure requires (2) if this situation occurs.
A. (1) nuclear flux spikes and possible fuel damage (2) tripping the RFP B. (1) nuclear flux spikes and possible fuel damage (2) shutting the RFP Discharge Valve C. (1) check valve slamming and possible water hammer (2) tripping the RFP D. (1) check valve slamming and possible water hammer (2) shutting the RFP Discharge Valve
NRC Question Data Sheet Choice Correct:
C Distractors:
A B
D Basis or Justification The procedural concern for maintaining this pressure relationship is for RFP discharge check valve slamming. The caution requires tripping the RFP if check valve slamming occurs.
In the same set of cautions, there is one related to flux spikes. However, this one is addressing the need to monitor power while slowly opening the discharge valve.
In the same set of cautions, there is one related to flux spikes. However, this one is addressing the need to monitor power while slowly opening the discharge valve. This distractor might be tempting because closing the discharge valve will remove the new cooler water that is being injected.
Although this is the issue the caution is addressing, the operator is required to trip the RFP rather than just shutting the discharge valve if check valve slamming occurs.
Level of Knowledge Difficulty Time Allowance (minutes)
LOW 2.5 3
Reference(s):
Learning 0 bjective:
REQUIRED MATERIALS:
New Exam Item 0 Modified Bank Item ILT Exam Bank 0 Previous NRC Exam 0 Other Exam Bank SO 6C.1.C-2, Startup of Second or Third Reactor Feedwater Pump PLOT-5006.5a Notes and Comments KnowledgeIAbility KIA Prepared By:
Philip E. Nielsen (717) 456-4122 PBAPS Regulatory Exam Author 259001 K5.02 Importance: RO I SRO Reactor Feedwater 2.5 12.5 Page 126 of 150
~
~
Peach Bottom February 2005 Initial License Operator NRC Examination
- 64.
Both units are operating at full power with the following conditions present:
0 0
Main Control Room Radiation Monitor RI-0760D is failed with a GP-25 Appendix 14 trip inserted.
CONTROL ROOM RAD MONITOR DIV. I1 INITIATED (003 A-3) is lit due to the GP-25 trip.
Later, an annunciator is received and the PRO reports:
0 CONTROL ROOM VENT SUPPLY FAN HI-LO (003 A-1) is alarming.
Flow Recorder FR-0765 indicates 200scfm and dropping.
0 Main Control Room Radiation Monitor RI-0760B is failed upscale high.
Evaluate these conditions to determine the status of Control Room Emergency Ventilation and the reason for that status.
The Control Room Emergency Ventilation System has:
A. STARTED due to the low flow condition.
B. STARTED due to the radiation monitor failure.
C. NOT started as indicated by the low flow condition.
D. NOT started since the full initiation logic is not satisfied.
~
NRC Question Data Sheet Correct:
Distractors:
1 Answer Key B
A I Question ID# 064 Both ROISRO Level of Knowledge Difficulty Time Allowance (minutes)
HIGH 3.25 4
SRO N/A Basis or Justification 272000 K3.10 Radiation Monitoring The CREV system is in service due to the combination of RI-0760B (failed high) and RI-0760D (GP-25 trip).
Importance:
~~
This is a good distractor because CREV will initiate on low flow, but in this case the low flow is being caused by the isolation of normal Control Room Ventilation.
The low flow signal is actually from normal Control Room Ventilation and is normal during a CREV initiation.
This is a solid distractor because this logic system is different, in that "B" and "D" make up the logic for initiation even though only a Div II alarm is received. For RPS or PCIS, "B" and "D" would only give a half initiation.
Source:
Reference(s):
Learning 0 bjective:
Knowledge/Ability WA Source Documentation New Exam Item 0 Modified Bank Item 0 ILT Exam Bank 0
Previous NRC Exam 0
Other Exam Bank GP-25 APPENDIX 14, MCR Ventilation Isolation, Division II SO 40D.1.A Control Room Ventilation Startup and CREV High Radiation PLOT-5040D. 1 a (Description of K&A, from catalog)
Knowledge of the effect that a loss or malfunction of Radiation Monitoring will have on the following:
Control Room Ventilation REQUIRED MATERIALS:
Notes and Comments Prepared By:
Philip E. Nielsen (717) 456-4122 PBAPS Regulatory Exam Author Page 128 of 150
Peach Bottom February 2005 Initial License Operator NRC Examination
- 65.
Unit 2 is operating at full power when the jet pump mixer for Jet Pump 1 1 becomes displaced. Evaluate the following to determine the effect this will have on Recirculation System indications.
In accordance with ON-100, Failure of a Jet Pump, this failure will cause a sudden RISE in:
A. Core Plate Flow.
B. dp on Jet Pump 12.
C. A Recirc Loop Drive Flow.
D. By Recirculation Pump Speed.
NRC Question Data Sheet
~
Psychometrics Level of Knowledge Difficulty Time Allowance (minutes)
N/A
~
Answer Key Source:
Reference(s):
Objective:
Learning Question ID# 065 Both R(
Choice New Exam Item 0 Modified Bank Item ILT Exam Bank 0 Previous NRC Exam 0 Other Exam Bank ON-100, Failure of a Jet Pump PLOT-5004.1 b Correct:
7 Knowledge/Ability KIA c
290002 K1.02 Importance: RO / SRO Reactor Vessel lnternals 3.2 J 3.2 Distractors: 7 I B 0
~
~~
~~
~
Basis or Justification ON-100 specifies that this condition will cause a sudden rise in redrc drive flow to the loop containing the defective jet pump. Since the A Redrc loop is connected to Jet Pumps 11-20, A Drive Flow will RISE.
ON-100 states that Core Plate Flow will LOWER due to the drop in core flow.
Jet Pump 12 shares the riser with 1 1. ON-100 states that this dP should LOWER.
Good distractor because common sense would associate Jet Pump 1 1 with the B Recirc Loop, however it is actually connected to the A Loop.
Recirculation Pump Speed in the affected loop will actually lower.
REQUIRED MATERIALS:
Notes and Comments Prepared By:
Philip E. Nielsen (717) 456-4122 PBAPS Regulatory Exam Author Page 130 of 150
Peach Bottom February 2005
~
Initial License Operator NRC Examination
- 66.
Given the following:
0 Both Units are operating at full power.
0 The entire shift team was present at a 0730 Shift Turnover Meeting conducted by the Control Room Supervisor (CRS) and the Shift Manager (SM).
0 At 1 100, the Fourth Reactor Operator (RO), who has been coordinating clearances, enters the Unit 2 Controls area to relieve the Unit 2 RO for lunch.
0 The Unit 2 RO will be eating in the Control Room Lunchroom.
What is the MINIMUM TURNOVER ACTIVITY and the MAXIMUM DURATION of this mid-shift turnover in accordance with OP-AA-112-101, Shift Turnover and Relief.
A. Tour the Main Control Boards with the off-going RO ONLY and the relief duration shall be a MAXIMUM of 30 minutes.
B. Tour the Main Control Boards with the off-going RO ONLY and the relief duration shall be a MAXIMUM of 60 minutes.
C. Review the Shift Turnover Checklist including any deviations ONLY and the relief duration shall be a MAXIMUM of 30 minutes.
D. Review the Shift Turnover Checklist including any deviations ONLY and the relief duration shall be a MAXIMUM of 60 minutes.
NRC Question Data Sheet Correct:
Answer Key Question ID# 066 Both RO/SRO D
Choice Level of Knowledge Difficulty Time Allowance (minutes)
LOW 2.5 3
SRO N/A l A Distractors:
Source:
Reference(s):
0 bject ive:
Lea rn i ng B
New Exam Item 0
Modified Bank Item 0
ILT Exam Bank 0
Previous NRC Exam 0 Other Exam Bank OP-AA-112-101 I Shift Turnover and Relief.
PLOT-1570.17 C
Knowledge/Ability K/A Basis or Justification Section 4.1.7 states that mid-shift relief for a maximum of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> only requires a review of the Shift Turnover Checklist and any deviations from the indicated status.
A Control Board tour is NOT required and the turnover is valid for 60 minutes.
A Control Board tour is NOT required.
PWG 2.1.03 Importance: RO / SRO Conduct of Operations 3.01 3.4 This mid-shift turnover is valid for 60 minutes.
REQUIRED MATERIALS:
Notes and Comments Prepared By:
Philip E. Nielsen (717) 456-4122 PBAPS Regulatory Exam Author Page 132 of 150
Peach Bottom February 2005 Initial License Operator NRC Examination
- 67.
Unit 2 is operating with the following conditions present:
Main Generator Load is 1080MWe Generator Hydrogen Pressure is 60 psig Power Factor (PF) is 0.95 lagging The Power System Director (PSD) directs you to REDUCE reactive loading fi-om 400 to 300 MVARs.
What is the MAXIMUM real load permitted by the attached Generator Capacity Curve based on this new reduced level of reactive loading?
A. 101OMWe B. ll00MWe C. 1130MWe D. 1240MWe
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~-
NRC Question Data Sheet Level of Knowledge HIGH I
Answer Kev Difficulty Time Allowance (minutes)
SRO 3.0 3
N/A I Question ID# 067 Both RO/SRO Source:
Reference(s):
Objective:
Learning I
Choice New Exam Item 0
Modified Bank Item 0
ILT Exam Bank Modified SO 50.1.A, Figure 1, Main Generator Capability Curve 0
Previous NRC Exam 0
Other Exam Bank PLOT-5050.10 Correct:
Knowledge/Ability KIA Distractors:
PWG 2.1 2 5 importance: RO / SRO Conduct Of Operations 2.8 / 3.1 L
C A
B D
Basis or Justification Correct based on the plotting of the junction of 300 MVARs with the 60#
Hydrogen Pressure line.
Based on plotting 300 MVARs on the incorrect 45# Hydrogen Pressure line.
Based on plotting the junction of.95 PF with the 60# Hydrogen Pressure line.
Based on plotting 300 MVARs on the incorrect (but normal) 75 # Hydrogen Pressure line.
REQUIRED MATERIALS:
SO 50.1.A-2, Figure 1 Generator Capacity Curve Notes and Comments Prepared By:
Philip E. Nielsen (717) 456-4122 PBAPS Regulatory Exam Author Page 134 of 150
Peach Bottom February 2005
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Initial License Operator NRC Examination
- 68.
Select the situation describing the proper use of an Alternate Verification Technique to minimize radiation exposure during the performance of a valve lineup in accordance with HU-AA-101, Human Performance Tools and Verification Practices.
A. A qualified Maintenance Technician is used to verify the position of the valve.
B. A qualified operator uses the system discharge flow indication to verify the position the valve.
C. A second qualified operator, who has not witnessed the activity, verifies the stem position of the valve.
D. Two qualified operators verify the correct component identification and then check the valve hand wheel in the CLOSED direction.
NRC Question Data Sheet Level of Knowledge LOW Answer Key Difficulty Time Allowance (minutes)
SRO 2.75 3
N/A Question ID# 068 Both RO/SRO PWG 2.1.29 Conduct of Operations Choice
~
~~~
Importance:
B Distractors:
A C
D Basis or Justification HU-AA-101 permits the use of Alternate Verification Techniques such as flow indication in various situations such as those with ALARA concerns.
This describes a Worker Verification (WV).
This describes Independent Verification (IV).
This describes Concurrent Verification (CV)
Psychometrics Source:
Reference(sk Learning Objective:
KnowledgelAbility WA Source Documentation New Exam Item 0
Modified Bank Item 0
ILT Exam Bank Previous NRC Exam 0
Other Exam Bank HU-AA-101 Human Performance Tools and Verification Techniques (Description of K&A, from catalog)
Knowledge of how to conduct and verify valve lineups.
REQUIRED MATERIALS:
Notes and Comments Prepared By:
Philip E. Nielsen (71 7) 456-41 22 PBAPS Regulatory Exam Author Page 136 of 150
Peach Bottom February 2005 Initial License Operator NRC Examination
- 69.
A Reactor startup is in progress. Plant conditions are as follows:
0 0
0 Reactor pressure is 750 psig.
Mode Switch is in SHUTDOWN.
All rods are filly inserted.
When directed by the CRS to place the Mode Switch to STARTUP/HOT STANDBY, the RO inadvertently places the Mode Switch to RUN.
Based on the above conditions after the Mode Switch is placed in RUN the MSIVs will -(
- 1) andRPS will (2)
(1)
(2)
A. CLOSE NOT TRIP B. CLOSE TRIP C. REMAINOPEN TRIP D. REMAINOPEN NOT TRIP
NRC Question Data Sheet Choice Answer Key Basis or Justification Correct:
B MSlVs will close due to <850# in RUN and RPS will trip on MSlV closure.
Distractors:
A RPS will NOT be reset.
C D
MSlVs will be CLOSED.
MSlV will be CLOSED and RPS will be TRIPPED.
KnowledgeIAbility I PWG 2.2.2 Level of Knowledge Difficulty Time Allowance (minutes)
HIGH 3.0 3
1 Importance:
Reference (s ) :
0 bjective:
Learning REQUIRED MATERIALS:
0 New Exam Item 0
Modified Bank Item 0
ILT Exam Bank GP-2 Normal Plant Startup.
Ix) Previous NRC Exam 2002 0 Other Exam Bank PLOT-506OF 4k; PLOT-5007G 4d Notes and Comments KIA I Generic - Equipment Control Prepared By:
Philip E. Nielsen (717) 456-4122 PBAPS Regulatory Exam Author 4.0 I 3.5 Page 138 of 150
Peach Bottom February 2005 Initial License Operator NRC Examination
- 70.
Surveillance testing of the 2B Standby Liquid Control (SLC) Pump is in progress.
The attached Surveillance Test (ST) is complete through step 6.3.20.
Complete the remaining ST steps to determine the status of steps 6.3.22 and 6.3.23 and the appropriate action(s).
A. Step 6.3.22 and 6.3.23 are SAT. Continue with the ST.
B. Step 6.3.22 is SAT. Step 6.3.23 is UNSAT. Stop the ST and notify Shift Management.
C. Step 6.3.22 is UNSAT. Step 6.3.23 is SAT. Note the UNSAT step in the remarks section and continue with the ST.
D. Step 6.3.22 and 6.3.23 are UNSAT. Stop the ST and notify Shift Management.
NRC Question Data Sheet Level of Knowledge HIGH Answer Key Difficulty Time Allowance (minutes)
SRO 2.5 4
NIA Question ID# 070 Both RC Choice Source:
Reference(s):
Objective:
Learning Correct:
0 New Exam Item 0
Modified Bank Item ILT Exam Bank 0 Previous NRC Exam 0 Other Exam Bank ST-0-01 1-301-2, SLC Pump Functional Test DBIG-PLOT-1570.16 Distractors:
KnowledgeIAbility KIA B
PWG 2.2.12 Importance:
RO I SRO Equipment Control 3.0 13.4 A
C D
SRO Basis or Justification Using the indicated formula, the calculated flow rate should be 59.7 gpm.
This makes step 6.3.22 SAT and step 6.3.23 UNSAT.
Step 6.3.23 is UNSAT, ST must be stopped and management informed.
Step 6.3.22 is SAT, Step 6.3.23 is UNSAT, and the ST must be stopped and management informed.
Based on the calculation, Step 6.3.22 is SAT.
REQUIRED MATERIALS:
ST page for question.
Notes and Comments Question received minor rewording but is not significantly modified.
Prepared By:
Philip E. Nielsen (717) 456-4122 PBAPS Regulatory Exam Author Page 140 of 150
Peach Bottom February 2005 Initial License Operator NRC Examination
- 71.
Unit 3 is in MODE 5 with refbeling activities in progress:
Evaluate the following conditions to determine the one which would require the Reactor operator to notifL the Fuel Handling Director to suspend core alterations in accordance with FHdC, Core Component Movement - Core Transfers.
A. The A Fuel Pool Cooling Service Water Booster Pump Overcurrent annunciator is received.
B. Shutdown Cooling (SDC) has been removed from service to complete a swap of SDC Loops.
C. Wide Range Neutron Count Rate doubles when a fifth fuel bundle is seated around the A WRNM detector.
D. The white rod permissive light on the C05 panel is NOT lit when the refuel platform is over the core with fbel loaded on the main hoist.
~~
NRC Question Data Sheet Choice Answer Key Basis or Justification Correct:
Distractors:
~~~~
C FH-6C requires that the FHD be notified to secure fuel handling when the WRNM count rate doubles after the fourth fuel bundle is placed around the detector. (Step 10.2.9)
Loss of a Fuel Pool Cooling Service Water Pump does NOT require securing fuel handling.
Swapping of SDC loops does not require securing fuel handling.
A B
D Under the conditions described, the white light should be extinguished.
I I
Level of Knowledge Difficulty Time Allowance (minutes)
LOW 3.0 3
SRO N/A Source:
Reference( s):
Objective:
Learning 0
New Exam Item 0
Modified Bank Item ILT Exam Bank 0 Previous NRC Exam 0 Other Exam Bank FH-6C Core Component Movement - Core Transfer.
NLSRO-0763.6 REQUIRED MATERIALS:
KnowledgelAbility WA Notes and Comments This question had minor wording changes and the distractors were reordered, but the question was not significantly modified.
PWG 2.2.30 Importance:
RO / SRO Equipment Control 3.5 / 3.3 Prepared By:
Philip E. Nielsen (717) 456-4122 PBAPS Regulatory Exam Author Page 142 of 150
Peach Bottom February 2005 Initial License Operator NRC Examination
- 72.
Given the following worker characteristics:
0 40 year old male.
0 0
0 Active as a radiation worker for the last ten years.
Lifetime exposure is currently 7.324 Rem.
His entire lifetime dose has been received at Peach Bottom.
No Administrative Dose Control Level (ADCL) extensions have been authorized.
In accordance with RP-AA-203, Exposure Control and Authorization, what is this workers Exelon ADCL on the morning of January 1 before entering the Radiological Controls Area (RCA).
The workers Exelon Administrative Dose Control Level (ADCL) is:
A. 750Mrem.
B. 1000 Mrem.
C. 2000Mrem.
D. 4000Mrem.
NRC Question Data Sheet Level of Knowledge D imcu Ity Time Allowance (minutes)
LOW 2.25 3
Answer Key SRO NIA Question ID# 072 Both RC PWG 2.3.01 Radiation Controls Choice Importance:
Distractors:
C A
B D
SRO Basis or Justification 2000 Mrem is the correct ADCL for this worker.
750 Mrem would be the ADCL for a worker where his last quarters exposure was unknown (for example: received at another location).
1000 Mrem would be the ADCL for a worker with a high lifetime exposure.
4000 Mrem was the previous PECO Administrative Limit for dose.
Source:
Reference@):
Learning Objective:
KnowledgeIAbility KIA Source Documentation New Exam Item 0 Modified Bank Item 0 ILT Exam Bank 0 Previous NRC Exam 0 Other Exam Bank EP-AA-203, Exposure Control and Authorization PLOT1 770.04 (Description of K&A, from catalog)
Knowledge of 10 CFR 20 and related facility radiation control requirments.
REQUIRED MATERIALS:
Notes and Comments Prepared By:
Philip E. Nielsen (717) 456-4122 PBAPS Regulatory Exam Author Page 144 of 150
Peach Bottom February 2005 Initial License Operator NRC Examination
- 73.
Given the following conditions:
0 A female, not declared to be pregnant, fully qualified radiation worker at Peach Bottom supported the April 2004 Limerick Generating Station outage where her Total Effective Dose Equivalent (TEDE) received was 250 &em.
Her current TEDE from PBAPS for 2004 is 200 &em.
She also received an estimated 10 mRem to her knee in June 2004 fiom a diagnostic test.
0 0
Determine the MAXIMUM annual non-emergency Total Effective Dose Equivalent (TEDE) that she can receive for the remainder of 2004 WITHOUT exceeding the Federal Exposure Limits?
A. 4540mRem B. 4550mRem C. 4790mRem D. 4800mRem
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~
NRC Question Data Sheet Correct:
Answer Key B
Question ID# 073 Both R(
Choice Level of Knowledge HIGH Difficulty Time Allowance (minutes)
SRO 2.5 3
NIA I
Distractors:
Source:
Reference@):
Learning 0 bject ive:
KnowledgeIAbility KIA I
C 0
New Exam Item 0
ILT Exam Bank 0 Previous NRC Exam 0 Other Exam Bank Modified NRC Bank Item RP-AA-203 PLOT-1730.3 PWG 2.3.04 Importance: RO I SRO Radiological Controls 2.5 13.1 D
SRO Basis or Justification 5000 mRem (Fed Limit) - 250 mRem (LGS) - 200 mRem (PBAPS) = 4550 mRem (Remaining Exposure) 5000 mRem (Fed Limit) - 250 mRem (LGS) - 200 mRem (PBAPS) - 10 mRem (Incorrectly subtracted non-occupational knee exposure) = 4540 mRem (Remaining Exposure) 5000 mRem (Fed Limit) - 10 mRem (Incorrectly subtracted non-occupational knee exposure) - 200 mRem (PBAPS) = 4790 mRem (Remaining Exposure). Calculation also missing 250 mRem (LGS).
5000 mRem (Fed Limit) - 200 mRem (PBAPS) = 4800 mRem (Remaining Exposure). Calculation missing 250 mRem (LGS).
REQUIRED MATERIALS:
Notes and Comments This question was significantly modified by changing the rad worker to a female, changing all of the exposure values, changing the correct answer and all of the distractors, and adding some non-occupational exposure to be considered in the problem.
Prepared By:
Philip E. Nielsen (717) 456-4122 PBAPS Regulatory Exam Author Page 146 of 150
Peach Bottom February 2005 Initial License Operator NRC Examination
- 74.
In accordance with OP-AA-103-102, Watchstanding Practices, with no prior approval of the Control Room Supervisor, the Reactor Operator may prioritizdimit the announcement of Main Control Room Annunciators to only those that the operator believes to be important:
A. during a routine Shift Turnover Meeting.
B. following a planned change in plant conditions.
C. during a planned HLA Briefing prior to RPS testing.
D. following a drywell leak and automatic reactor scram.
NRC Question Data Sheet Choice Answer Key Basis or Justification Correct:
Distractors:
D Step 4.3.4 states that during transient conditions with higher level procedures in effect, announcement of alarms may be suspended and that the announcing of alarms must not take precedence over stabilizing the plant.
Shift Turnover is not a condition permitting the suspension of Alarm Announcement.
A B
C Even with a planned change in plant conditions, the alarm must be announced unless the SRO has given advance permission for that specific annunciator.
Annunciators must be announced even if they interfere with a planned HLA briefing.
Level of Knowledge Difficulty Time Allowance (minutes)
LOW 2.25 3
Source:
Reference(s):
Objective:
Learning Notes and Comments New Exam Item 0
Modified Bank Item 0 ILT Exam Bank 0
Previous NRC Exam 0
Other Exam Bank OP-AA-102-103, Watchstanding Practices DBIG-PLOT-1570.17 Prepared By:
Philip E. Nielsen (717) 456-4122 PBAPS Regulatory Exam Author Knowledge/Ability WA Page 148 of 150 PWG 2.4.12 Importance:
RO I SRO Emergency Procedures and Plan 3.4 13.9
Peach Bottom February 2005 Initial License Operator NRC Examination
- 75.
Unit 2 was operating at full power when the following transient occurred:
0 0
0 B Reactor Feedwater Pump Tripped.
Reactor Level is 15 inches and dropping slowly.
Reactor Power is 100% and steady.
Based on these plant conditions, the Reactor Operator must immediately:
A. run both Recirculation Pumps manually back to 30% speed.
B. run both Recirculation Pumps manually back to 45% speed.
C. perform a plant shutdown in accordance with GP-4, Manual Reactor Scram.
D. lower power in accordance with GP-5, Power Operations, until water level is restored.
NRC Question Data Sheet Level of Knowledge Difficulty HIGH 2.5 Answer Kev Time Allowance (minutes)
SRO 4
NIA Question ID# 075 Both RO/SRO Source:
Reference(s):
0 bjective:
Learning Choice Correct:
New Exam Item 0 Modified Bank Item 0 ILT Exam Bank OT-100, Reactor Low Level 0
Previous NRC Exam 0
Other Exam Bank PBIGPLOT-1540.03 Distractors:
KnowledgeIAbility KIA A
PWG 2.4.49 Importance: RO I SRO Emergency Procedures and Plan 4.0 14.0 C
D Basis or Justification Reactor water level below 17" with a RFP with less than 20% flow should have caused a 45% Recirculation Runback. Operators are required to make failed automatic actions occur.
The candidate would select this answer if he believed that the Recirculation Runback failure should have caused a 30% runback.
A full reactor scram is not required to recover level under these circumstances. A reactor scram would indicate that the operator was slow in performing his required immediate verifications and actions.
Although lowering power is a normally required action for a lack of make-up capacity, it is completed using GP-9, not GP-5. It is also not correct for these circumstances because the Recirculation Runback had failed and must be made to happen.
REQUIRED MATERIALS:
Notes and Comments Prepared By:
Philip E. Nielsen (717) 456-4122 PBAPS Regulatory Exam Author Page 150 of 150
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Peach Bottom February 2005 Initial Senior Reactor Operator NRC Examination
- 1.
Unit 2 was operating at 2800 Mwt when the following transient occurred:
The A Recirc Pump tripped.
0 The B Recirc Pump Speed remains at 1200 RPM.
Power dropped to 1757 MWt.
0 APRMs are oscillating between 48 and 53% power in approximately 4-5 second random intervals.
0 Indicated Core Flow on FR-2-2-3-095 (Black Pen) is 52 Mlbm/hr.
0 Inactive Loop Flow on FI-2-2-3-092A is 5 Mlbm/hr.
Assess plant conditions using the attached GP-5-1, PBAPS Power Flow Operation Map, then identify the correct procedural response to this issue.
The plant is operating in:
A. Region 1, SCRAM the reactor and enter T-100 due to being in Region 1.
B. Region 2, SCRAM the reactor and enter T-100 due to indications of Thermal Hydraulic Instability.
C. Region 2, immediately exit Region 2 by raising B Recirc Pump speed using SO 2A. 1.D-2, without exceeding 56 Mlbm/hr.
D. the normal operating region, continue with the follow-up actions of OT-112, Unexpected/Unexplained Change in Core Flow.
NRC Question Data Sheet Choice Answer Kev Basis or Justification A
I Correct:
I C
I The calculation of core flow 52-2(5) = 42 Mlbm/hr / 102.5 Mlbdhr = 41%.
Pwr 1757-3514 = 50% This is region 2, action required is to raise reurc flow to exit the region immediately.
The correctly plotted position on Power Flow Map is in Region 2, a scram NOT required.
Distractors: I------
B D
The indications provided do provide an indicated power oscillation, but it does not meet the criteria for THI, scram is NOT required.
If the candidate does not multiply two times the inactive flow (common error), he will believe that the operating point is just inside the normal Level of Knowledge Difficulty Time Allowance (minutes)
HIGH 2.5 4
SRO 1 OCFR55.43.b.5 Source :
295001 AA2.01 Partial or Complete Loss of Forced Core Flow Circulation Reference(s):
Learning Objective:
Importance:
SRO 3.8 Knowledge/Ability KIA Source Documentation New Exam Item 0
Modified Bank Item 0
ILT Exam Bank 0
Previous NRC Exam 0
Other Exam Bank OT-112, Unexpected/Unexplained Change in Core Flow PBIGPLOT-1540.3,4 (Description of K&A, from catalog)
Knowledge of the operational implications of the following concepts as they apply to Partial or Complete Loss of Forced Core Flow Circulation:
Power/Flow Map REQUIRED MATERIALS:
GP-5-1, PBAPS Power Flow Operation Map Notes and Comments Prepared By:
Philip E. Nielsen (71 7) 456-41 22 PBAPS Regulatory Exam Author
Peach Bottom February 2005 Initial Senior Reactor Operator NRC Examination
- 2.
Unit 2 was operating in MODE 1 when the following reports were received fiom maintenance and engineering:
Station Battery 2CDO1 terminal voltage is 98 VDC.
The repairs cannot be made in a timely manner A Unit 2 shutdown to MODE 3 is:
A. required in a maximum of 14 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br />.
B. required in a maximum of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
C. required in a maximum of 7 days and 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
D. NOT required unless 4KV bus de-energization is required.
NRC Question Data Sheet Level of Knowledge Difficulty Time Allowance (minutes)
HIGH 2.0 4
Answer Key SRO 1 OCFR55.43. b. 1 1 Question # 2 SRO Source:
Reference@):
Objective:
Knowledge/Ability KIA Learning I
Choice New Exam Item 0
Modified Bank Item 0
ILT Exam Bank 0
Previous NRC Exam 0
Other Exam Bank T.S. 3.8.4 PLOT-5057.8 295004 G2.2.22 Importance:
SRO Partial or Complete Loss of DC Pwr.
4.1 Correct
Distractors: L A
B C
D Basis or Justification TS 3.8.4.C requires restoration within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
TS 3.8.4.D requires MODE 3 in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
2 + 12 = 14 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br />.
If candidate mistakenly applies the TS 3.8.4.B (for Unit 2), it is 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> to restore.
12 + 12 = 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
If candidate mistakenly applies the TS 3.8.4.A, it is 7 days to restore.
7 days + 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
If the candidate misapplies the note under condition A, it could be interpreted no actions are required unless a 4KV bus is deenergized.
REQUIRED MATERIALS:
TS 3.8.4 Notes and Comments Prepared By:
Philip E. Nielsen (717) 456-4122 PBAPS Regulatory Exam Author
Peach Bottom February 2005 Initial Senior Reactor Operator NRC Examination
- 3.
Technical Specifications require that RPS initiate a scram by a given Drywell Pressure.
What is the bases for the value of the High Drywell Pressure Limiting Condition for Operat ion.
A. To prevent exceeding the Drywell Spray Initiation Limit.
B. To minimize instrument failures due to the drywell environment.
C. To prevent exceeding static head limitations on the drywell support structure.
D. To minimize the energy required to be absorbed by the Primary Containment.
NRC Question Data Sheet B
C Answer Key Question # 3 SRO Environmental condition impact on drywell instrumentation is of concern, but is not the basis for the High Drywell Pressure scram.
Containment integrity is of concern, but is not based on exceeding static head limitations.
Choice Correct:
Level of Knowledge LOW Distractors:
D ifficu I ty Time Allowance (minutes)
SRO 3.5 3
1 OCFR55.43. b. 1 D
Source:
Reference( s):
Objective:
Learning A
New Exam Item 0
Modified Bank Item 0
ILT Exam Bank 0 Previous NRC Exam 0 Other Exam Bank TS Bases 3.3.1.1 PLOT-5002B.9 Basis or Justification Knowledge/Ability WA
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TS Bases 3.3.1.1 states that the High Drywell Pressure setpoint is based upon minimizing the energy that must be absorbed by the primary containment.
Although the bases is for minimizing energy that must be absorbed by containment, it is not based on preventing exceeding the DWSIL curve.
295024 G 2.2.25 Importance: SRO High Drywell Pressure 3.7 Psychometrics REQUIRED MATERIALS:
Notes and Comments Prepared By:
Philip E. Nielsen (717) 456-4122 PBAPS Regulatory Exam Author
Peach Bottom February 2005 Initial Senior Reactor Operator NRC Examination
- 4.
Unit 2 has experienced a drywell steam leak with an Anticipated Transient Without Scram (ATWS). Current conditions are as follows:
0 Reactor Pressure is being maintained 800- 1000 psig.
0 Level has been lowered to control reactor power.
The current RPV level band is - 100 to - 150 inches.
0 HPCI and RCIC are running to control reactor level.
The Reactor Operator reports the following:
0 RPV Level Indications
+ NarrowRange
+5 inches.
+ WideRange
- 1 15 inches.
+ FuelZone
-125 inches.
+ Refuel Range (LI-86)
-21 inches 0
0 TI-2501 point 126 is not available TI-2501 point 127 indicates 510°F Interpret these drywell temperatures using Table DW/T-1 from T-102, Primary Containment Control to determine which RPV level indication ranges can be used and what actions are required for level control.
A. Narrow Range and Wide Range. Declare level unknown due to extreme level divergence and enter T-116, RPV Flooding.
B. Fuel Zone Range and Refuel Range. Declare level unknown due to extreme level divergence and enter T-116, RPV Flooding.
C. Narrow Range and Refuel Range. Direct a HPCI shutdown to restore the level band using T-240, Termination and Prevention of Injection.
D. Wide Range and Fuel Zone Range. Direct using these indicators to maintain the level band using T-240, Termination and Prevention of Injection.
NRC Question Data Sheet Level of Knowledge D ifficu I ty Time Allowance (minutes)
HIGH 3.5 5
Answer Kev SRO 10CFR55.43.b.5 Question # 4 SRO Source:
Reference(s):
Objective:
Knowledge/Ability KIA Learning Choice New Exam Item
[7 Modified Bank Item 0
ILT Exam Bank T-102 and curve DW/T-1 0
Previous NRC Exam c]
Other Exam Bank PLOT-1 560.4,5,7 295028 EA2.01 Importance:
SRO High Drywell Temperature
4.1 Correct
Distractors:
D A
B C
Basis or Justification Interpreting drywell temperature on DWn-1 indicates that the Wide Range and Fuel Zone indications are still functional. Since level is in the correct band, direct maintaining level.
Narrow Range indication is inaccurate based on the Min Indicated Level vs Max Run Temp Chart. If the candidate believed these indications were valid, the wide divergence in indications might result in a T-116 entry on level unknown.
Refuel Range indication is inaccurate based on its section of the DWn-1 curve. If the candidate believed these indications were valid, the wide divergence in indications might result in a T-116 entry on level unknown.
Although these indications are relatively close in indication, they are both inaccurate based on DWn-1. If the candidate believed they were accurate and level was actually between +5" and -21", it would be appropriate to secure HPCl to restore level to the specified band.
REQUIRED MATERIALS:
T-102, Primary Containment, DW/T-1 Curve Notes and Comments Prepared By:
Philip E. Nielsen (71 7) 456-41 22 PBAPS Regulatory Exam Author
Peach Bottom February 2005 Initial Senior Reactor Operator NRC Examination
- 5.
The following Unit 2 conditions exist during an Anticipated Transient Without Scram (ATWS):
Reactor Power is 35%
Reactor Pressure is 1 100 psig with SRVs lifting on overpressure.
+ The Main Turbine is tripped due to low condenser vacuum, which is currently 6 inches Hg.
+ The only source of Instrument Nitrogen to the SRVs is the accumulators.
Torus Water Level is 13 feet and steady after the isolation of a leak on the D FUR Pump discharge while in torus cooling.
Torus Temperature is 160°F and rising 0
0 Interpret the Heat Capacity Temperature Limit (HCTL) Curve (TR-1) to determine the actions required as Torus temperature continues to RISE.
A. Use T-101, RPV Control, to lower pressure with Bypass Jack operation to maintain pressure BELOW the HCTL Curves.
B. Use T-101, RPV Control, to lower pressure with manual SRV operation to maintain pressure BELOW the HCTL Curves.
C. Perform a T-112, Emergency Blowdown, when Torus Temp exceeds 175°F AND the ATWS has been terminated.
D. Perform a T-112, Emergency Blowdown, when Torus Temp exceeds 175°F REGARDLESS of Reactor Power.
NRC Question Data Sheet Choice Answer Key Basis or Justification Level of Knowledge Difficulty Time Allowance (minutes)
HIGH 3.5 4
Correct:
SRO 1 OCFR55.43.b.5 At 175°F Torus Temp, an Emergency Blowdown is required regardless of reactor power.
Source:
Reference (s) :
Objective:
Learning Distractors:
(XI New Exam Item 0
Modified Bank Item 0
ILT Exam Bank 0 Previous NRC Exam 0 Other Exam Bank T-102 Curve T/T-I Heat Capacity Temperature Limit Curve PLOT-1560.3 A
Knowledge/Ability KIA B
295030 EA2.02 Importance: SRO Low Suppression Pool Water Level 3.9 C
Although lowering pressure to stay below the HCTL curve would be desirable, the Bypass Jack is NOT available due to low condenser vacuum sealing out the BPVs.
Although lowering pressure to stay below the HCTL curve would be desirable, the SRVs are NOT available since only the accumulators are available for SRV manual actuation.
At 175°F Torus Temp, an Emergency Blowdown is required regardless of reactor power.
REQUIRED MATERIALS:
T-102 Curve T/T-1 Heat Capacity Temperature Limit Notes and Comments Prepared By:
Philip E. Nielsen (71 7) 456-41 22 PBAPS Regulatory Exam Author
Peach Bottom February 2005
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Initial Senior Reactor Operator NRC Examination
- 6.
An Anticipated Transient Without Scram (ATWS) is in progress on Unit 3. The following conditions exist.
Reactor Power of 38%
The Main Turbine is tripped on low vacuum Condenser vacuum is steady at 15 inches of Hg.
ALL Bypass Valves are full open.
Reactor Pressure being controlled 950-1050 psig with SRVs.
Drywell Pressure 4 psig and steady due to a small steam leak.
Torus Temperature 1 10°F and rising slowly T-221, MSIV Bypass, is complete.
T-240, Termination and Prevention of Injection, has been initiated. The following occur as RPV Level lowers:
-1 10 inches, all SRVs are shut with pressure steady and controlled by EHC.
- 130 inches, B APRM Downscale is received.
-145 inches, ALL APRM Downscales have been received and are steady.
-172 inches, the RO announces he is at -172 inches and requests to reinject.
Interpret these conditions and responses to the lowering level to determine the appropriate procedurally required Level Control Band given specifically in T-1 17, Level Power Control and how level should be maintained.
Maintain RPV Water Level in the band of:
A. -1 10 to -172 using HPCI.
B. -130 to -172 using feedwater.
C. -145 to -195 using HPCI.
D. -172 to -195 using feedwater.
NRC Question Data Sheet C
Answer Key Question # 6 SRO Basis or Justification The criteria for terminating the lowering of level was met at -145 inches. T-117 directs a band of -195 to the level to which it was intentionally lowered.
Choice Correct:
A B
Distractors:
~~~~
Although SRVs went closed at -1 10 inches, the criteria to stop lowering was not met because Drywell Pressure is >2 psig. Also the bottom of the band should be -195 inches. Although the MSlVs remain open, feedwater is not available due to low vacuum.
Although the first APRM downscale came in at -1 30, core wide power was not yet below 4%. Also the bottom of the band should be -195 inches.
D Criteria for terminating the level lowering was met when all the APRM downscales came in, before actual level reached -172 inches. Although the MSlVs remain open, feedwater is not available due to low vacuum.
Level of Knowledge Difficulty Time Allowance (minutes)
HIGH 3.5 4
SRO 1 OCFR55.43.b.5 Source:
Reference@):
0 bject ive:
Learning New Exam Item 0
Modified Bank Item 0
ILT Exam Bank T-I 17 LeveVPower Control 0
Previous NRC Exam 0
Other Exam Bank PLOT-1560.3 REQUIRED MATERIALS:
Knowledge/Ability 295037 EA2.02 KIA SCRAM Condition Present and Power Above APRM Downscale or Unknown Notes and Comments Importance: SRO 4.2 Prepared By:
Philip E. Nielsen (717) 456-4122 PBAPS Regulatory Exam Author
Peach Bottom February 2005 Initial Senior Reactor Operator NRC Examination
- 7.
Consider the following timeline:
101 0 - Unit 2 Turbine Building 1 16 Elevation fire alarm received.
1012 - Incident Commander (I-C) responds to the scene with Member #l.
1015 - I-C reports that an actual fire exists in Turbine Building 116.
1020 - I-C reports that the onsite Fire Brigade is fighting the fire.
1030 - I-C reports that the fire is not yet under control and fire fighting is still in progress.
In accordance with FF-01, Fire Brigade, and ON-1 14 for an actual reported fire, the Control Room is required to notify the Incident Commander that they will be requesting off-site assistance at (1) and the CRS is required to direct A. (1) 1020 (2) an immediate rapid shutdown using GP-4, Manual Reactor Scram.
B. (1) 1025 (2) the Floor Supervisor to relieve the I-C using FF-01, Fire Brigade.
C. (1) 1030 (2) a Controlled Shutdown using GP-3, Normal Plant Shutdown.
D. (1) 1035 (2) isolation of the RPV Condensing Chamber Backfill using ON-1 14, Att. 1.
NRC Question Data Sheet C
A Answer Key Question # 7 SRO Basis or Justification 1030 is the correct time, based on 20 minutes after the fire alarm. A controlled shutdown is required based on ON-114 direction after requesting off-site assistance for fire fighting.
Time is incorrectly based on 10 min. after the fire alarm. A GP-4 is only directed if the fire jeopardizes normal plant shutdown OR ECCS.
Choice B
Correct:
Time is incorrectly based on 10 min. after the actual fire report or 15 minutes after the fire alarm (this time corresponds to when an emergency classification is made per ON-114). The Floor Supervisor role by FF-01 is not to relieve the I-C, but to coordinate between the I-C and the Control Room.
Distractors:
D Time is incorrectly based on 20 minutes after the report of an actual fire.
RPV Condensing Chamber Backfill is only isolated with a Reactor Building Fire.
Level of Knowledge LOW Difficulty Time Allowance (minutes)
SRO 2.5 4
1 OCFR55.43.b.5 600000 G2.4.30 Plant Fire On Site Source:
Importance: SRO
3.6 Reference(s)
Learning 0 bjective:
Knowledge/Ability WA Source Documentation New Exam Item 0
Modified Bank Item ILT Exam Bank FF-01, Fire Brigade.
ON-114, Actual Fire Reported in the Power Block, Diesel Generator Building, Emergency Pump, Inner Screen, or Emergency Cooling Tower Structures 0
Previous NRC Exam 0 Other Exam Bank PBIGPLOT-1550.19 (Description of K&A, from catalog)
Plant Fire On Site: Knowledge of which events related to system operations/status should be reported to outside agencies.
REQUIRED MATERIALS:
Notes and Comments Prepared By:
Philip E. Nielsen (717) 456-4122 PBAPS Regulatory Exam Author
Peach Bottom February 2005 Initial Senior Reactor Operator NRC Examination
- 8.
Unit 3 was operating at full power when a failure in the Digital Feedwater Control System resulted in the following:
Feedwater Control automatically shifted to single element control.
0 The Feedwater Control system is making NO additional adjustments to the feedwater pump speeds.
The RO reports that actual Steam Flow is stable and higher than Feed Flow.
With no automatic response by the Feedwater Level Control System, identifl the required procedural response.
A. Direct a power reduction using GP-3, Normal Plant Shutdown.
B. Direct restoring 3-element control using SO 6C.l.D-3, FW Auto Control.
C. Direct taking manual control of the Master Level Controller in accordance with SO 6C. 1.D-3, FW Auto Control.
D. Direct a manual scram before reactor level exceeds +35 inches in accordance with OT-1 10, Reactor High Level.
NRC Question Data Sheet Level of Knowledge Difficulty Time Allowance (minutes)
HIGH 2.5 4
Answer Key SRO 1 OCFR55.43.b.5 Question # 8 SRO 295009 AA2.02 Low Reactor Water Level Choice Importance: SRO
3.7 Correct
Distractors:
C A
B D
Basis or Justification Taking manual control of the Master Level Control is appropriate since it is not responding to conditions in automatic.
Action to reduce power to match available feed flow would be appropriate, but OT-100, Reactor Low Level, directs the use of GP-9, Fast Power Reduction, NOT GP-3, Normal Plant Shutdown.
The system does not need to be forced to three element as it should have responded automatically even in single element.
The operator is directed in OT-I 10 to control level below +35 inches, but not to shutdown before exceeding it.
Source:
Reference s :
Learning t--
0 bjective:
Knowledge/Ability 1 KIA Source Documentation New Exam Item 0
Modified Bank Item 0
ILT Exam Bank OT-I 00, Reactor Low Level 0 Previous NRC Exam 0 Other Exam Bank PBIGPLOT-1540.3 (Description of K&A, from catalog)
Ability to determine and/or interpret the following as it relates to Low Reactor Water Level:
Steam flow/Feed flow mismatch REQUIRED MATERIALS:
Notes and Comments Prepared By:
Philip E. Nielsen (717) 456-4122 PBAPS Regulatory Exam Author
~~
~~
Peach Bottom February 2005 Initial Senior Reactor Operator NRC Examination
- 9.
Given the following:
Peach Bottom has experienced a radiological accident condition.
0 2 VENT EXH STACK RAD MONITOR HIITROUBLE A (218 B-5) a l m s 2 VENT EXH STACK RAD MONITOR HIITROUBLE B (218 C-5) alarms The PRO reports that the Unit 2 Vent Stack Radiation is reading above the HI alarm level.
The source of the radiation release is the and the CRS must (2)
(1)
A. (1) Standby Gas Treatment Exhaust (2) enter and execute T-104, Radioactivity Release.
B. (1) Radwaste Building Ventilation Exhaust (2) direct the termination of radwaste processing using the appropriate radwaste procedures.
C. (1) Recombiner Building Ventilation Exhaust (2) direct the evacuation of all unnecessary recombiner personnel using GP-15, Local Evacuation.
D. (1) PEARL Building Ventilation Exhaust (2) direct restarting ventilation using SO 40AA. 1.A, Setup and Operation of PEARL Heating and Ventilation.
NRC Question Data Sheet HIGH Answer Key 2.0 4
I 1 OCFR55.43.b.4,5 1 I Question # 9 SRO I
Source:
Reference@):
Learning 0 bject ive:
Knowledge/Ability K/A Correct:
New Exam Item 0
Modified Bank Item 0
ILT Exam Bank ON-104, Vent Stack High Radiation P B I GPLOT-1 540.9a 295017 AA2.04 Importance: SRO 0 Previous NRC Exam 0 Other Exam Bank High Off-Site Release Rate 4.3 I I I
Basis or Justification I
~
B
~
The Radwaste Building Vent Exhausts to the Unit 2 Vent Stack. Radwaste operations must be terminated under these conditions.
A C
D The SBGT Exhaust is to the Main Stack. T-104 would not be entered until radiation reached the HI HI alarm point.
The Recombiner Building Exhaust is to the Unit 3 Vent Stack. Evacuating all unnecessary personnel from the recombiner would be correct if it was the source of the leak.
The PEARL Building Exhaust is to the Unit 3 Vent Stack. Restarting ventilation would be appropriate if the PEARL was the source.
I 1
I Psychometrics I
Level of Knowledge 1 D ifficu I ty I Time Allowance (minutes) I SRO I
I I
REQUIRED MATERIALS:
Notes and Comments Prepared By:
Philip E. Nielsen (717) 456-4122 PBAPS Regulatory Exam Author
~
~~
Peach Bottom February 2005 Initial Senior Reactor Operator NRC Examination
- 10.
When a Control Rod Drive (CRD) pump trips during plant operation, if Charging Water Header pressure drops below 940 psig, and at least two accumulator trouble alarms are lit, the timing of a required scram is based on reactor pressure.
What is the Tech Spec requirement and its basis?
With reactor pressure:
A. less than 900 psig, scram immediately, because reactor pressure alone is adequate to ensure the complete insertion of the control rods.
B. less than 900 psig, scram in 20 minutes, to allow adequate time for the recovery of the CRD pump and charging pressure during low power, low risk conditions.
C. greater than 900 psig, scram immediately, to ensure as many control rods as possible are inserted prior to a potential loss of scram capability.
D. greater than 900 psig, scram in 20 minutes, to provide adequate time to recover a CRD pump while reactor pressure alone can hlly insert the control rods.
NRC Question Data Sheet Correct:
D i s t ractors:
Answer Key D
A Level of Knowledge LOW I
B Difficulty Time Allowance (minutes)
SRO 3.0 3
1 OCFR55.43.b.2 I
C Source:
Reference@):
0 bjective:
KnowledgelAbility KIA Learning
~
Basis or Justification New Exam Item 0 Modified Bank Item 0 ILT Exam Bank TS 3.1.5 and Bases.
0 Previous NRC Exam Other Exam Bank PLOT-5003A.9 295022 G2.2.25 Importance: SRO Loss of CRD Pumps 3.7 Tech Spec Bases 3.1.5 states that with reactor pressure 2 900 psig, it is permissible to allow 20 minutes to recover CRD because reactor pressure alone has the ability to fully insert the Control Rods.
With reactor pressure e900 psig, an immediate scram is required, but the rods may NOT have adequate force to be fully inserted on a scram.
With reactor pressure e900 psig, you are not permitted to wait 20 minutes because scram capability is rapidly degrading.
With reactor pressure 2 900 psig, an immediate scram is NOT required because the reactor pressure will fully insert the control rods.
REQUIRED MATERIALS:
Notes and Comments Prepared By:
Philip E. Nielsen (717) 4564122 PBAPS Regulatory Exam Author
Peach Bottom February 2005
~ _ _ _
Initial Senior Reactor Operator NRC Examination 1 1.
Given the following conditions:
0 0
0 Unit 2 has experienced a Loss of Coolant Accident (LOCA).
HPCI is injecting into the RPV at maximum flow to control level.
HPCI suction is currently aligned to the Torus.
A break in the Torus occurs and causes Torus level to begin dropping at a rapid rate.
The FIRST time HPCI operation will be impacted is when Torus Level lowers to below:
A. 9.5 ft, when a HPCI Shutdown is required in accordance with SO 23.2.A-2.
B. 9.5 ft, when aligning the HPCI suction to the CST is required using SO 23.7.B-2.
C. 6 ft, when a HPCI Shutdown is required in accordance with SO 23.2.A-2.
D. 6 ft, when aligning the HPCI suction to the CST is required using SO 23.7.B-2.
NRC Question Data Sheet Level of Knowledge Difficulty LOW 2.5 Answer Key Time Allowance (minutes)
SRO 4
1 OCFR55.43.b.5 Question # 11 SRO Source:
Reference@):
Learning Objective:
Knowledge/Ability KIA Choice New Exam Item 0
Modified Bank Item 0
ILT Exam Bank 0
Previous NRC Exam 0
Other Exam Bank T-102, Primary Containment, and bases step T/L-13 PLOT-5023.6e 206000 A2.07 Importance: SRO HPCl
3.6 Correct
Distractors:
A B
C D
Basis or Justification T-102, Primary Containment Control, requires a HPCl shutdown at 9.5 feet due to uncovering the HPCl exhaust line which could cause pressuring of the containment.
A candidate could select this response believing that it is the HPCl suction from the Torus that is uncovered.
Although a HPCl shutdown is required, action is required significantly before Torus level Reaches 6 feet.
Action is required to shutdown HPCl significantly before 6 feet in the Torus.
Verifying HPCl suction alignment is not directed.
REQUIRED MATERIALS:
Notes and Comments Prepared By:
Philip E. Nielsen (717) 456-4122 PBAPS Regulatory Exam Author
Peach Bottom February 2005 Initial Senior Reactor Operator NRC Examination
- 12.
Unit 2 has experienced Loss of Off-site Power (LOOP). The following conditions are present:
0 The Reactor is Shutdown 0
Reactor Pressure is 1000 psig and rising 0 Reactor Level is -30 inches and lowering 0 Group I, 11, and I11 isolations have gone to completion.
0 All Diesels have automatically started and are supplying their associated busses.
As conditions deteriorate, with no operator action:
(1) describe the response of the ADS valves to a valid ADS initiation signal AND (2) provide the action necessary to support long term ADS valve operation.
The ADS valves initially would (1) must be directed to support long term ADS (2)
SRV operation.
A. (1) NOT open (2) SO 16.1.A-2, Inst. Nitrogen System Startup and Normal Operations, B. (1) NOT open (2) T-261, Placing the Backup Inst. Nitrogen supply from CAD Tank in service, C. (1) OPEN (2) SO 16.1.A-2, Inst. Nitrogen System Startup and Normal Operations, D. (1) OPEN (2) T-261, Placing the Backup Inst. Nitrogen supply from CAD Tank in service
NRC Question Data Sheet Choice Answer Kev Basis or Justification Level of Knowledge Difficulty Time Allowance (minutes)
HIGH 2.5 4
Correct:
SRO 1 OCFR55.43.b.5 Distractors:
21 8000 A203 ADS D
Importance:
SRO 4.6 A
B The valves would initially open due to the ADS SRV Instrument Nitrogen Accumulators. T-261 is necessary to establish long term Nitrogen to the ADS SRVs.
The Valves would open due to their accumulators. Instrument Nitrogen cannot be started up normally due the continued presence of the isolation signal (as determined by reactor level c 1 inch).
T-261 is required for long term operation of the ADS SRVs, but the Valves would open initially due to their accumulators.
l c The valves would initially open due to the ADS SRV Instrument Nitrogen Accumulators, but normal instrument nitrogen cannot be realigned due to the isolation signal present (as determined by reactor level c 1 inch).
Source :
Reference( s):
Learning Objective:
Knowledge/Ability WA Source Documentation New Exam Item 0
Modified Bank Item 0
ILT Exam Bank 0 Previous NRC Exam 0 Other Exam Bank T-101, RPV Control, Step RCIP-4 SO 16.1.A-2, Inst. Nitrogen Normal System Startup, Prerequisites T-261, Placing Instrument Nitrogen Backup with CAD tank in service PLOT-5001 G.4d, 6d (Description of K U, from catalog)
Ability to (a) predict the impacts of the following on ADS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:
Loss of air supply to ADS valves REQUIRED MATERIALS:
Notes and Comments Prepared By:
Philip E. Nielsen (71 7) 456-41 22 PBAPS Regulatory Exam Author
Peach Bottom February 2005 Initial Senior Reactor Operator NRC Examination
- 13.
The following conditions exist during an Anticipated Transient Without Scram (ATWS) on Unit 2:
0 Reactor Power is 30%
0 0
Reactor Pressure is 1000 psig Reactor Level is 23 inches.
The following occurred when the SBLC Pump Selector was placed in Start Sys A.
SBLC Pump A red light lit with 1080 psig indicated discharge pressure.
Both Squib Valve Continuity Lights are lit.
GROUP IYIII OUTBOARD ISOL RELAYS NOT RESET (214 E-1) alarmed MO-2-12-015, Cleanup Inlet Isolation (Outboard) Valve, Stroked Closed MO-2-12-018, Cleanup Inlet Isolation (Inboard) Valve, Stroked Closed All RWCU Pumps tripped.
(1) Evaluate these conditions to assess whether the plant responded as designed and then (2) determine the required actions for this condition.
A. (1) SBLC and PCIS responded as designed.
(2) Continue with actions as directed by T-117, LevelPower Control B. (1) PCIS responded as designed, but the SBLC Squib valves failed to fire.
(2) Start the B Pump using RRC 1 1.1 -2 SBLC Initiation During A Plant Event.
C. (1) SBLC responded as designed, but RWCU failed to klly isolate.
(2) Complete the isolation using GP-8B, PCIS Isolation - Groups I1 and III.
D. (1) RWCU failed to fully isolate causing SBLC to have low discharge pressure.
(2) Complete the isolation using GP-8B, PCIS Isolation - Groups I1 and 111.
NRC Question Data Sheet Choice I
Answer Kev Basis or Justification Correct:
C SBLC responded normally. The Group 11/111 Inboard isolation on SBLC initiation failed, must manually isolate the system using GP8B.
Distractors:
A SBLC responded normally, but the RWCU system did not isolate fully due to failure of the Inboard Isolation logic. T-117 actions are appropriate, but do not resolve the failed PClS isolation.
The SBLC squib lights stay lit as long as the pump is running. The valves fired normally as indicated by discharge pressure being just above reactor pressure rather than several hundred psig higher lifting the discharge reliefs as they would have been if the Squibs didnt fire.
RWCU did fail to fully isolate because the outlet valve did not go closed, however, this would not effect the discharge pressure of SBLC which is in an acceptable range anyway.
B D
Level of Knowledge HIGH Difficulty Time Allowance (minutes)
SRO 3.0 4
1 OCFR55.43.b.5 Prepared By:
Philip E. Nielsen (71 7) 456-41 22 PBAPS Regulatory Exam Author Source:
Reference@):
0 bjective:
Learning New Exam Item 0
Modified Bank Item 0
ILT Exam Bank 0
Previous NRC Exam 0
Other Exam Bank GP-8B, PClS Isolation - Groups II and Ill PLOT-5011.49 Knowledge/Ability
. WA 223002 A211 Importance:
SRO PCIS/Nuclear Steam Supply Shutoff 3.9
Peach Bottom February 2005 Initial Senior Reactor Operator NRC Examination
- 14.
Unit 2 is operating at power, near the end of cycle, with the following Safety Relief Valve (SRV) failures:
0 0
D SRV deenergized (fuses removed) due to a failed solenoid.
E SRV Bellows is confirmed failed.
L SRV Bellows is confirmed failed.
Based on the above conditions, which one of the following is correct regarding (1) continued plant operation and (2) the basis, according to Technical Specifications?
A. (1) Be in Mode 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and Mode 4 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
(2) Sufficient relief valves will NOT open on safety hction.in the event of a design MSIV closure transient.
B. (1) Be in Mode 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and dome pressure I 100 psig within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
(2) One of the ADS Safety Relief Valves will not function in the event of a LOCA.
C. (1) NO Tech Spec Actions required.
(2) Sufficient relief valves are available to open on safety function in the event of a design MSIV closure transient.
D. (1) NO Tech Spec Actions required.
(2) Sufficient relief valves are available to be manually opened in the event of a Turbine Trip Without Bypass.
NRC Question Data Sheet C
Answer Key Correct, a combination of 11 SRV/SVs remain to function in the event of a design overpressurization transient, i.e., MSlV closure.
Question # 14 SRO I
B Choice
~
No SRVs above are designated ADS valves.
I Basis or Justification D
Correct:
Tech Specs require solenoid opening only for ADS. ATWS not considered in the TS bases.
Distractors:
Level of Knowledge Difficulty Time Allowance (minutes)
HIGH 2.5 4
SRO 1 OCFR55.43.b.2 Sufficient SRV/SVs are available.
Knowledge/Ability WA 239002 PWG 2.2.25 Importance: SRO Re1 ief/Safety Valves
3.7 Source
Reference(s):
Learning 0 bjective:
Source Documentation 0
New Exam Item 0
Modified Bank Item 0
ILT Exam Bank TS 3.4.3 and bases TS 3.5.1 and bases PLOT5001A.08,9 Previous NRC Exam 2002 0 Other Exam Bank REQUIRED MATERIALS:
Tech Spec Sections 3.4.3 and 3.5.1 with NO Surveillance Requirements.
Notes and Comments Prepared By:
Philip E. Nielsen (717) 456-4122 PBAPS Regulatory Exam Author
Peach Bottom February 2005 Initial Senior Reactor Operator NRC Examination 1 5.
Given these conditions:
0 Unit 2 was operating at full power.
0 The crew is beginning a Surveillance Test for Full Load Testing of the Emergency Diesel Generator (EDG).
0 The EDG is running, ready for synchronization on the E42 Bus.
0 The E-42 Breaker Synch Switch is turned on with the Synch Scope rotating slow in the fast direction.
Under these conditions, a complete loss of off-site power occurs.
Evaluate this condition to assess (1) the status of the E-4 EDG and the E-42 Breaker and (2) the required procedural actions.
A. (1) E-4 EDG is TRIPPED, E-42 Breaker is OPEN.
(2) Restart the EDG using SO 52A.7.AY Diesel Generator Manual Emergency Start. E-42 must be manually closed after resetting the anti-pump lockout.
B. (1) E-4 EDG is TRIPPED, E-42 Breaker is OPEN.
(2) Restart the EDG using SO 52A.7.AY Diesel Generator Manual Emergency Start. E42 Breaker will close automatically when the EDG is running.
C. (1) E4 EDG is RUNNING, E-42 Breaker is OPEN.
(2) The anti-pump lockout must be manually reset using SO 52.1.B, Diesel Generator Operations before the E-42 Breaker will close.
D. (1) E-4 EDG is RUNNING, E-42 Breaker is CLOSED.
(2) Monitor and control EDG Loading during continued operation using SO 52.1.By Diesel Generator Operations.
NRC Question Data Sheet C
A Answer Key Question # 15 SRO I
These conditions will send a trip signal to the E-42 Breaker but not to the E4 EDG. Because E42 receives simultaneous trip and close signals from the dead bus condition, the breaker will receive an anti-pump lockout and must be reset manually.
E4 EDG will not receive a trip signal so it does not require restarting. The anti-pump lockout on the E42 Breaker must be reset.
I Choice B
I Basis or Justification E4 EDG will not receive a trip signal so it does not require restarting. The anti-pump lockout on the E42 Breaker must be reset before it can close.
Correct:
D Distractors:
I------
E4 EDG will be running, but the E42 Breaker can not close due to an anti-pump lockout of the breaker.
Level of Knowledge Difficulty Time Allowance (minutes)
HIGH 3.0 4
SRO 1 OCFR55.43.b.5 Source:
Reference( s):
Learning Objective:
New Exam Item 0
Modified Bank Item 0
ILT Exam Bank 0
Previous NRC Exam 0
Other Exam Bank SO 52A. 1.B, Diesel Generator Operations Precautions PLOT-5052.6f PLOT-5054.6b
~~~~~
~~
Knowledge/Ability 264000 A2.07 KIA EDGs Importance:
SRO 3.7 REQUIRED MATERIALS:
Notes and Comments Prepared By:
Philip E. Nielsen (717) 456-4122 PBAPS Regulatory Exam Author
Peach Bottom February 2005 Initial Senior Reactor Operator NRC Examination
- 16.
Unit 3 is operating in MODE 1 at full power.
Electrical Maintenance reports that during a review of a work order from the outage, it was determined that the A Recirculation Pump MG Set Drive Motor Breaker has a faulty trip coil.
The A Recirculation Pump is running normally.
Based on these conditions, Technical Specifications require the ATWS-RPT Reactor Pressure Trip capability to be restored within:
A. 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
B. 6hours.
C. 72hours.
D. 14days.
NRC Question Data Sheet Correct:
Distractors:
Answer Key Question # 16 SRO
~
A T.S. 3.3.4.1 C applies since without trip capability, both functions do NOT have trip capability maintained.
6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> is how long you have to shutdown the Recirc Pump if you do not met the required actions.
72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> corresponds to the time if only one function did not have trip ca pa bi I ity.
14 days corresponds to the time if only one channel of the logic had been lost.
B C
D I
Choice I
Basis or Justification Level of Knowledge HIGH Difficulty Time Allowance (minutes)
SRO 4.0 4
1 OCFR55.43.b. 1 I
I I
Source:
Reference(s):
Objective:
Knowledge/Ability K/A Learning I
Psvchometrics I
0 New Exam Item 0
Modified Bank Item ILT Exam Bank 0
Previous NRC Exam 0
Other Exam Bank Tech Spec 3.3.4.1 and Bases PLOT-5002.8 202001 G2.2.22 Importance: SRO Recirculation 4.1 REQUIRED MATERIALS:
Unit 3 T.S. 3.3.4.1 Notes and Comments Question was reformatted, but not significantly modified.
Prepared By:
Philip E. Nielsen (71 7) 456-41 22 PBAFS Regulatory Exam Author
Peach Bottom February 2005 Initial Senior Reactor Operator NRC Examination
- 17.
The following conditions exist on Unit 2:
100%Power 0
0 Reactor Feed Pumps (RFPs) are operating in Master Automatic A Level Select is lit solid after being manually selected by the operator AUTO/ABC Select Pushbutton is NOT lit.
Under these conditions, the reference leg for the A narrow range level transmitter (LT-52A) develops a small leak which is slightly larger than the makeup rate of its condensing chamber and Backfill input.
ACTUAL Reactor Level will (1) and the crew must (2)
A. (1) LOWER 3 to 4 inches and stabilize at a new lower value (2) select single element control using SO 6C. 1.D-2, Feedwater Level Control.
B. (1) LOWER continuously resulting in a Scram on low RPV level (2) manually control level using OT-100, Reactor Low Level.
C. (1) FUSE 3 to 4 inches and then stabilize at a new higher value (2) select AUTO/ABC using SO 6C. 1.D-2, Feedwater Level Control.
D. (1) FUSE until the Main and RFP Turbines trip on high RPV level (2) manually control level below 35 inches using OT-1 10, Reactor High Level.
NRC Question Data Sheet Level of Knowledge HIGH Answer Key Difficu I ty Time Allowance (minutes)
SRO 2.5 5
1 OCFR55.43.b.5 Question # 17 SRO Source:
Reference(s):
Learning 0 bjective:
Choice New Exam Item 0
Modified Bank Item 0
ILT Exam Bank OT-1 00, Reactor Low Level 50028.3k 0 Previous NRC Exam 0
Other Exam Bank Corfect:
Knowledge/Ability WA Distractors:
216000 A2.03 Importance: SRO Nuclear Boiler Instrumentation 3.0 B
A C
D Basis or Justification With less pressure in the reference leg, the variable leg pressure will appear to be larger causing indicated level to rise. Feedwater would respond by lowering ACTUAL level. Manual control is required per OT-100, Reactor Low Level to prevent a low-level scram.
Although ACTUAL level will lower, it will not stabilize at a new lower value as it would with a steam or feed flow transmitter failure. Placing the system in single element will not resolve the issue and the system is level dominant anvwav.
ACTUAL level will not rise and selecting AUTO/ABC will not recover level.
ACTUAL level will not rise and OT-1 10 entry is not required.
REQUIRED MATERIALS:
Notes and Comments Prepared By:
Philip E. Nielsen (717) 456-4122 PBAPS Regulatory Exam Author
~
Peach Bottom February 2005 Initial Senior Reactor Operator NRC Examination
- 18.
Given the following conditions:
Unit 2 is in MODE 5.
0 The Mode Selector Switch is in REFUEL.
The Refueling Platform is over the Spent Fuel Pool.
0 A fuel bundle had been loaded on the Main Hoist and raised out of the fuel pool storage rack.
Which one of the following activities will generate a Rod Block?
A. The Main Hoist is raised to the full up position.
B. A single Control Rod is selected, but not withdrawn.
C. The Refueling Platform is moved over the reactor vessel.
D. The Mode Selector Switch is placed in the Startup/Hot Standby position.
NRC Question Data Sheet Level of Knowledge HIGH Answer Key Difficulty Time Allowance (minutes)
SRO 2.5 3
1 OCFR55.43. b.7 Question # 18 SRO 234000 A3.02 Fuel Handling Equipment Choice Importance: SRO
3.7 Correct
Distractors:
C A
D Basis or Justification These conditions would result in a Rod Block.
Raising the Hoist to the full up position will dear a Rod Block.
Selecting the control rod without withdrawing it will NOT result in a Rod Block.
Placing the MSS in Startup/Hot Standby will NOT cause a Rod Block under these conditions.
Source Documentation Source :
Reference(s):
Learning 0 bjective:
Knowledge/Ability KIA 0
New Exam Item 0
Modified Bank Item E ILT Exam Bank 0 Previous NRC Exam 0 Other Exam Bank SO 62.7.A-2, Receipt of Rod Blocks NLSRO-0762.13 (Description of K&A, from catalog)
Ability to monitor automatic operations of the Fuel Handling Equipment System including:
Interlock ODeration REQUIRED MATERIALS:
Notes and Comments This question has been reworded since its use on the 1999 NRC exam, but is NOT significantly modified.
Prepared By:
Philip E. Nielsen (71 7) 456-41 22 PBAPS Regulatory Exam Author
Peach Bottom February 2005
~
Initial Senior Reactor Operator NRC Examination
- 19.
After returning from a week of vacation, a SRO has worked the following schedule:
Monday night, he arrives at 1800 to relieve the dayshift CRS and work a fill nightshift (1 900-0700).
After his shift ends on Tuesday morning, the CRS is required to stay to attend some required management training from 0700 to 1000.
The CRS returns at 1800 on Tuesday evening to relieve the dayshift CRS and work another full nightshift.
At shift turnover, the SRO learns that he is expected to attend a management meeting to review Equipment Operator performance on Wednesday morning from 0700 to 1100.
The SRO is not scheduled for any additional work during this week.
Use LS-AA-119, Overtime Controls (attached), to evaluate these conditions and determine what, if any, Overtime Guideline Deviations (OGDs) are required for the SRO to meet this CRS shift schedule and additional training and meeting expectations.
(An) Overtime Guideline Deviation(s) islare:
A. NOT required for the described conditions.
B. REQUIRED for the last three hours of his second nightshift.
C. REQUIRED for each of the activities held after his 12-hour CRS shift.
D. REQUIRED for the last four hours of his second nightshift and the Tuesday morning meeting.
NRC Question Data Sheet B
A Answer Key The Operator worked 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> in the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, which leaves only 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> to be worked in his next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Shift Turnover time is not considered. Although hours worked in any position are considered in the calculations, deviations are only required for working in a safety related position (meaning that the first days training time counts, but a deviation is not required for the second days meeting).
Recognizing that Shift Turnover time does not count and that deviations are not required for non-safety-related work, the candidate could believe that no deviations are reauired.
Question # 19 SRO I
C Choice The candidate could misinterpret the 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in 48 requirement to mean that the SRO needs a deviation each time he exceeds his 12-hour shift.
I Basis or Justification D
Correct:
Correctly recognizing that the SRO worked 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> in his first day, the candidate could believe that a deviation is required for all hours after eight in the second day. The candidate would be erring by not subtracting Turnover Time and by requesting a deviation for non-safety-related work on Tuesday morning.
Distractors:
Level of Knowledge Difficulty Time Allowance (minutes)
HIGH 3.0 4
SRO 1 OCFR55.43.b.l Source:
Reference(s):
0 bjective:
Learning New Exam Item 0
Modified Bank Item 0 ILT Exam Bank 0
Previous NRC Exam 0
Other Exam Bank LS-AA-119, Overtime Controls PLOT-1 800.1 2 Knowledge/Ability KIA PWG 2.1.05 Importance: SRO 3.4 Notes and Comments Following these procedural overtime limitations is required by Tech Spec Section 5.2.2, Administrative Controls, Organization, Unit Staff. This supports the required tie to 10 CFR 55.43.b.l.
Prepared By:
Philip E. Nielsen (71 7) 456-41 22 PBAPS Regulatory Exam Author
Peach Bottom February 2005 Initial Senior Reactor Operator NRC Examination
- 20.
Given the following Unit 3 Conditions:
0 Reactor Mode Switch is in Shutdown.
0 Average Reactor Coolant Temperature is 222°F.
0 Reactor Vessel Head closure bolt tensioning will be complete in 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
The Unit is currently in (1) and will be in (2) when Reactor Vessel Head closure bolt tensioning is complete.
A. (1) MODE 3 (2)MODE 3 B. (1) MODE 4 (2)MODE 3 C. (1) MODE 5 (2)MODE 3 D. (1) MODE 5 (2)MODE4
NRC Question Data Sheet Level of Knowledge Difficulty LOW 3.0
~
Answer Key Time Allowance (minutes)
SRO 4
10CFR55.43.b.l Question # 20 SRO Source:
Reference@):
Objective:
Learning Choice New Exam Item 0
Modified Bank Item 0
ILT Exam Bank 0
Previous NRC Exam 0
Other Exam Bank Tech Spec Table 1.l-1 and bases.
PLOT-1800.4 Correct:
Knowledge/Ability KIA Distractors:
PWG 2.1.22 Importance: SRO 3.3 C
A B
D
~
~
~~
Basis or Justification With the Reactor Mode Switch in Shutdown and with the head closure bolts not fully tensioned, the Unit is considered to be in MODE 5, the refueling mode. Since temp is > 212OF, it will shift to MODE 3 when the bolts are tensioned.
Due to a temperature of > 212OF, a candidate may believe that we are in MODE 3, Hot Shutdown and may believe this is not affected by tensioning the head closure bolts.
With the head closure bolts not fully tensioned, a candidate may expect that the condition would be considered MODE 4 (cold shutdown) and that it would change to MODE 3 (hot shutdown) when the bolts are tensioned to hold pressure.
The candidate may recognize that the Unit is MODE 5, and believe that it will move to MODE 4 (cold shutdown) when the bolts are tensioned.
REQUIRED MATERIALS:
Notes and Comments Prepared By:
Philip E. Nielsen (717) 456-4122 PBAPS Regulatory Exam Author
Peach Bottom February 2005
- 21.
~-
Initial Senior Reactor Operator NRC Examination Unit 2 is in MODE 1 at full power.
An applicable Tech Spec Surveillance with a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> fiequency was last performed satisfactorily at 1230 on 1/14/05.
The Limiting Condition for Operation (LCO) required actions direct that the equipment be restored to OPERABLE status in 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, or be in MODE 3 in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and MODE 4 in 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
If a plant priority on Unit 3 prevents the surveillance from be performed, track the LCO to determine the EARLIEST time Unit 2 is required to be in MODE 4.
A. By 0430 on 1/17/2005.
B. By 0630 on 1/17/2005.
C. By 1030 on 1/17/2005.
D. By 2230 on 1/17/2005.
NRC Question Data Sheet C
A Answer Key Basis or Justification Last performed at 1230 on 1/14 + 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> until due (1230 on 1/15) + 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> grace (1830 on 1/15) + 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to restore (2230 on 1/15) + 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> to MODE 4 (1030 on 1/17)
Last performed at 1230 on 1/14 + 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> until due (1230 on 1/15) + 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to restore (1630 on 1/15) + 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> to MODE 4 (0430 on 1/17)
Question # 21 SRO I
D Choice Last performed at 1230 on 1/14 + 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> until due (1230 on 1/15) + 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> grace (1830 on 1/15) + 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to restore (2230 on 1/16) +12 hours to MODE 3 (INCORRECT 1030 on 1/16) + 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> to MODE 4. (2230 on Correct:
Level of Knowledge HIGH Distractors:
Difficulty Time Allowance (minutes)
SRO 3.5 5
1 OCFR55.43.b. 1 Knowledge/Ability K/A performed at 1230 on 1/14 + 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> until due (1230 on 1/15) + 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> grace (1830 on 1/15) + 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> to MODE 4 (0630 on 1/17)
PWG 2.2.23 Importance: SRO Equipment Control 3.8 Psychometrics I
Source:
Reference(s):
Learning Objective:
Source Documentation 0 New Exam Item 0 ILT Exam Bank 0 Previous NRC Exam 0 Other Exam Bank Modified Bank Item Tech Spec 1.3, 1.4, and SR 3.02.
PLOT 1800.6 REQUIRED MATERIALS:
Notes and Comments This question was significantly modified by rewording the stem and changing all of the distractors and the correct answer.
Prepared By:
Philip E. Nielsen (717) 456-4122 PBAPS Regulatory Exam Author
Peach Bottom February 2005 Initial Senior Reactor Operator NRC Examination
- 22.
Unit 2 is in a Refueling Outage. Which one of the following Refuel Floor activities must be DIRECTLY supervised by the Fuel Handling Director (Le., Designated Alternate is NOT permitted) in accordance With FH-6C, Core Component Movement - Core Transfers?
A. Cleaning recirc jet pumps in the Vessel.
B. Loading a new fuel bundle into the Vessel.
C. Moving old LPRM strings to the Spent Fuel Pool.
D. Shuffling of irradiated fuel in the Spent Fuel Pool.
NRC Question Data Sheet Psychometrics Level of Knowledge Difficulty Time Allowance (minutes)
1 OCFR55.43.b.7 Answer Key Question # 22 SRO PWG 2.2.29 Equipment Control I
Choice Importance:
Distractors:
B A
C D
Basis or Justification New fuel into the core is a Core Alt and requires direct supervision by the Fuel Handling Director.
~
~
Cleaning jet pumps may be supervised by the Designated Alternate.
~~
~
Does not require direct supervision of FHD or DA.
Does not require direct supervision of FHD or DA.
Source:
Reference(s):
Lea m i ng 0 bject ive:
Knowledge/Ability K/A Source Documentation 0
New Exam Item Previous NRC Exam 2002 0
Modified Bank Item 0 ILT Exam Bank 0 Other Exam Bank FH-GC Section 7.0 N LSR00763.0 (Description of K&A, from catalog)
Knowledge of SRO Fuel Handling responsibilities REQUIRED MATERIALS:
Notes and Comments Prepared By:
Philip E. Nielsen (717) 456-4122 PBAPS Regulatory Exam Author
Peach Bottom February 2005 Initial Senior Reactor Operator NRC Examination
- 23.
Given the following conditions:
The Equipment Operators (EOs) need to perform a surveillance test in an area with 80 mR/hr general area radiation levels.
The job will take 60 minutes for one EO, but could be completed in only 35 minutes with two EOs.
The Radiation Protection Engineer has determined that the area radiation levels could be reduced to 8 mR/hr with shielding.
The shielding would take 30 minutes for one individual to install, but could be completed in half the time with two individuals.
All radiological precautions have been taken.
An ALARA briefing is in progress.
Considering total personnel dose only, which of the following is directed by the SRO to ensure that the job dose is maintained As Low As Reasonably Achievable (ALARA) in accordance with Rp-AA-400, ALARA Program?
A. TWO EOs perform the surveillance WITHOUT shielding.
B. ONE EO performs the surveillance WITHOUT shielding.
C. TWO individuals INSTALL shielding and then TWO EOs perform the surveillance.
D. ONE individual INSTALLS the shielding and then ONE EO performs the surveillance.
NRC Question Data Sheet D
A Answer Key 1 individual for 30 minutes in an 80 mWhr field = 40 mR exposure PLUS 1 individual for 60 minutes in an 8 mWhr field = 8 mR = 48 mR TOTAL job exposure.
2 individuals for 35 minutes in an 80 mWhr field = 93.3 mR total exposure.
Question ID# 023 SRO I
B I
Basis or Justification
~~~
1 individual for 60 minutes in an 80 mWhr field = 80 mR total exposure.
Choice C
Correct:
2 individuals for 15 minutes in an 80 mWhr field = 40 mR exposure PLUS 2 individuals for 35 minutes in an 8 mWhr field = 9.3 mR = 49.3 mR TOTAL job exposure.
Distractors:
Level of Knowledge HIGH Difficulty Time Allowance (minutes)
SRO 3.5 3
1 OCFR55.43.b.4 Source:
Reference@):
Learning Objective:
New Exam Item Modified Bank Item ILT Exam Bank 0 Previous NRC Exam 0 Other Exam Bank RP-AA-400 ALARA Program PLOT1 770.03 KIA I Radiation Controls 2.6 13.0 Knowledge/Ability PWG 2.3.02 I Importance:
Notes and Comments Prepared By:
Philip E. Nielsen (717) 456-4122 PBAPS Regulatory Exam Author
-~
~-
Peach Bottom February 2005 Initial Senior Reactor Operator NRC Examination
- 24.
An electrical problem has resulted in the loss of all Rod Position Indication (RPIS) on the full core display. A few minutes later, a manual reactor scram was performed with the following conditions:
All RPIS indicators are blank.
Reactor power is 3.0 E-2% lowering.
0 RPV level lowered to - 10 inches and is now +20 inches.
RPV pressure is 1000 psig controlled by EHC.
Drywell pressure is 0.3 psig.
Scram Air Header pressure is 0 psig.
Based on the above conditions, which one of the following is correct?
An ATWS:
A. IS in progress. Exit T-100 and then enter T-101 at RC-1.
B. IS in progress. Enter T-100 and concurrently execute T-101 RC/Q rods.
C. is NOT in progress. Exit T-100 and enter T-101 at RC-1.
D. is NOT in progress. Enter T-100 and concurrently enter GP-3, Plant Shutdown.
NRC Question Data Sheet Level of Knowledge HIGH Answer Key Difficulty Time Allowance (minutes)
SRO 2.5 3
1 OCFR55.43.b.5 Question # 24 SRO Knowledge/Ability WA Correct:
PWG 2.4.4 Importance: SRO 4.3 B
A C
D Basis or Justification Correct, Rod position is unknown and an ATWS is assumed. T-100 refers the operator to perform T-1 01 RC/Q rods concurrently since rods are assumed to be not full in. No T-101 entry conditions exist.
Do not exit T-100 in this condition.
An ATWS is in progress. Do not exit T-100.
~
~~
An ATWS is in progress. GP-3 not executed concurrently with T-100. T-100 performed and then exited to GP-3.
Source:
Reference@):
Learning Objective:
Source Documentation 0
New Exam Item 0
Modified Bank Item ILT Exam Bank T-100 Step S-17 and bases T-101 bases PLOT2100.03 0
Previous NRC Exam 0 Other Exam Bank REQUIRED MATERIALS:
Notes and Comments Minor rewording, does not count as a significantly modified question.
Prepared By:
Philip E. Nielsen (71 7) 456-4122 PBAPS Regulatory Exam Author
Peach Bottom February 2005 Initial Senior Reactor Operator NRC Examination
- 25.
Unit 3 experienced a plant transient and a reactor scram. The following conditions are present:
Reactor Level
-15 inches Reactor Power 0%
Reactor Pressure 780 psig Reactor Temperature 5 15°F Drywell Pressure Drywell Temperature 289°F Containment Radiation 6.5E+05 R/hr Torus Level 15 feet Wind Direction (from) 270 degrees Wind Speed 5 mph Dose Assessment System Inoperable Rose to 28 psig, then dropped in a rapid and unexplained fashion Use the attached portions of the PBAPS EAL Matrix (EP-AA-1007) and LimerickPeach Bottom Plant-Based PAR Flowchart (EP-AA-111, Attachment 8) to determine the appropriate Protective Action Recommendation (PAR) for these conditions.
The Emergency Director must recommend KI for the general public in evacuated areas and evacuate a radius of:
A. 2 miles in all sectors AND 5 miles in the WNW, W, and WSW sectors.
B. 2 miles in all sectors AND 5 miles in the ENE, E, and ESE sectors.
C. 5 miles in all sectors AND 10 miles in the WNW, W, and WSW sectors.
D. 5 miles in all sectors AND 10 miles in the ENE, E, and ESE sectors.
NRC Question Data Sheet D
A B
C Answer Key Question # 25 SRO Basis or Justification With all three barriers lost, it is appropriate to evacuate a radius of 5 miles and 10 miles in the downwind sectors. With the wind coming from 270 degrees, it is appropriate to evacuate ENE, E, and ESE.
With all three barriers lost, evacuating a radius of only 2 miles and 5 miles in the downwind sectors is not adequate. This answer also indicates the wrong sectors to evacuate based on a common error in table usage.
With all three barriers lost, evacuating a radius of only 2 miles and 5 miles in the downwind sectors is not adequate. This answer does contain the correct sectors to evacuate based on wind direction.
With all three barriers lost, it is appropriate to evacuate a radius of 5 miles and 10 miles in the downwind sectors. This answer indicates the wrong sectors to evacuate based on a common error in table usage.
Correct:
Level of Knowledge Difficulty Time Allowance (minutes)
HIGH 2.5 5
I Distractors:
SRO 1 OCFR55.43. b.4 Emergency Plan
4.0 Source
Reference(s):
Learning Objective:
Knowledge/Ability KIA Source Documentation 0
New Exam Item 0
Modified Bank Item 0
ILT Exam Bank 0 Previous NRC Exam Other Exam Bank - LOR Bank EP-AA-111, Emergency Classification and Protective Action Recommendations EP-AA-1007, PBAPS EAL Matrix EPP-1100 PWG 2.4.44 I Importance:
SRO (Description of K U, from catalog)
Knowledge of emergency action protective action recommendations.
REQUIRED MATERIALS:
EP-AA-1007, PBAPS EAL Matrix Pages 3-5 thru 3-1 1 EP-AA-111, Attachment 8, LimericWPeach Bottom Plant-Based PAR Flowchart Notes and Comments Prepared By:
Philip E. Nielsen (717) 4564122 PBAPS Regulatory Exam Author