ML15293A552

2015-10 Draft Written Exam

ML15293A552
Person / Time
Site:	South Texas
Issue date:	10/02/2015
From:	Vincent Gaddy Operations Branch IV
To:	South Texas
References
Download: ML15293A552 (257)

Text

Examination Outline Cross-

Reference:

Level RO 000009 Small Break LOCA / 3 Tier # 1 Knowledge of the operational implications Group # 1 of the following concepts as they apply to K/A # 000009, EK1.01 the small break LOCA:

Importance Rating 3.5 EK1.02 Use of steam tables RO Question 1 Given the following conditions:

1. RCS pressure is currently 1020 psia and is remaining stable

2. Core Exit Thermocouple (CET) Temperature is 532 F

3. RCS Temperature is lowering at a rate of 60 degrees per hour

4. Steam Generators are at the 60% NR and stable with no Main or Auxiliary Feedwater flow.

5. All ECCS equipment is operating as designed ASSUME RCS Pressure Remains stable and no other operator actions are occurring.

The RCS is currently ___(1)___ and the RCS cooldown is being controlled by ____(2)_____.

(1) (2)

A. Saturated, combination of break cooling and steaming the Steam Generators B. Subcooled, combination of break cooling and steaming the Steam Generators C. Saturated, break cooling only D. Subcooled, break cooling only

Proposed Answer: D Explanation (Optional):

Steam Tables with a RCS pressure of 1020 psia the saturation temperature of water is 547 F, therefore the RCS is currently sub-cooled. Given that the steam generators are at their required level & stable and no feed is being supplied it eliminates the possibility that the cooldown is caused by steaming and leaves the only possible choice to be break flow as the cooling mechanism.

Technical Reference(s): Steam Tables 2000 ASME Steam Tables (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: 2000 ASME Steam Tables Learning Objective: (As available)

Question Source: New X 2392 Question History: Last NRC Exam N/A Question Cognitive Level: Comprehension or Analysis Difficulty 3.0 10 CFR Part 55 Content: (CFR 41.8 / 41.10 / 45.3)

Comments:

Examination Outline Cross-

Reference:

Level RO 000015/17 RCP Malfunctions / 4 Tier # 1 Ability to operate and / or monitor the Group # 1 following as they apply to the Reactor K/A # 000015/17 AA1.22 Coolant Pump Malfunctions (Loss of RC Importance Rating 4.0 Flow)

AA1.22 RCP seal failure/malfunction Question 2 Given the following:

• Unit 1 is operating at 100% power

• Annunciator RCP 1A NO 1 SEAL LKF FLOW HI/LO alarms.

• Seal leakoff flow for RCP 1A indicates 0.8 gpm on CP-004.

Which of the following is the probable cause for the seal leakoff flow being low?

A. High seal water injection temperature B. Loss of seal injection flow C. Excessive #2 seal leakage D. Low VCT Pressure

Proposed Answer: C Explanation (Optional):

A. Incorrect; High seal water injection temperature decreases the density of the leakoff water resulting in a pressure imbalance (higher pressure drop across the inner portion of the ring) that causes the seal to open and pass more flow.

B. Incorrect; Loss of seal injection flow results in warmer liquid from the RCS to leak up and through the seal. The increased leakoff temperature, increases leakoff flow.

C. Correct; Excessive #2 seal leakoff lowers the backpressure on #1 seal causing it to close up and flow to go down.

D. Incorrect; Low VCT pressure results in less back pressure on the #1 seal, therefore a higher #1 seal leakoff flow.

Technical Reference(s): 0POP09-AN-4M7-B-1, Rev 29 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: Bank # NRC 37 Question History: Last NRC Exam 1999 Question Cognitive Level: Comprehension or Analysis Difficulty 3.0 10 CFR Part 55 Content: 55.41.7 55.45.5 55.45.6 Comments:

Parent:

Given the following:

• Unit 1 is operating at 100% power

• Annunciator RCP 1B NO 1 SEAL LKF FLOW HI/LO alarms.

• Seal leakoff flow for RCP 1B indicates 0.9 gpm on CP-004.

Which ONE of the following is the probable cause for the seal leakoff flow being low?

A. High seal water injection temperature B. Loss of seal injection flow C. Excessive #2 seal leakage D. Low VCT Pressure Answer C

RCP 1A(2A) NO 1 SEAL LKF FLOW HI/LO Probable Causes: 1) High flow:

a) Damage to the Number 1 Seal b) High seal water injection temperature c) Loss of seal water injection flow d) Number 1 Seal Ring cocked e) Instrument failure f) High Seal 1 water inlet temperature

2) Low flow:

a) Insufficient DP across the Number 1 Seal b) Excessive leakage at the Number 2 Seal c) Damage to the Number 1 Seal d) Number 1 Seal inlet clogged e) SEAL LKF ISOL FV-3154 closed f) Instrument failure

Examination Outline Cross-

Reference:

Level RO 000025 Loss of RHR System / 4 Tier # 1 Ability to determine and interpret the Group # 1 following as they apply to the Loss of K/A # 000025, AA2.07 Residual Heat Removal System:

Importance Rating 3.4 AA2.07 Pump cavitation Question 3 The Unit is in Mode 5 mid-Loop Operation with RCS level at +9 inches.

RHR Pump 1A is in service with a flow rate of 3000 gpm.

At 10:00 AM the MAB RPO reports to the Control Room that RHR 1A is making a loud oscillating swishing noise when he was in the area.

The Primary RO then notices that RHR 1A amps are swinging more than normal, and RCS hot leg level is indicating +1.5 inches.

The Unit Supervisor has announced entry into the Loss of Residual Heat Removal Off normal Procedure, 0POP04-RH-0001, Loss of Residual Heat Removal.

The NEXT operator action that the control room operators should take is to ___________.

A. Reduce RHR 1A pump flow to 1000 to 1500 gpm B. Decrease RCS temperature until the noise and amperage oscillations stop C. Stop RHR 1A and Start RHR1B D. Reduce RHR 1A pump flow to < 1000 gpm

Proposed Answer: A Explanation (Optional):

A. Correct - With RCS NR Hot Leg Level at >1 inch, procedure direction is to reduce RHR flow within limit per Addendum 2 (1000 to 1500 qpm). This range should eliminate cavitation at the given level.

B. Incorrect - This choice is not procedure based and would not help to correct the problem.

C. Incorrect - Starting an additional RHR pump when you have indications of cavitation would not be the correct action. Plus the procedure requires a level of

+6 inches prior to attempting a start or restart of an RHR Pump.

D. Incorrect - Reducing RHR flow to < 1000 gpm would risk an RHR Pump trip on low flow. (925 gpm)

Technical Reference(s):

1. 0POP04-RH-0001, Loss of Residual Heat Removal (Rev. 27)

(Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: Modified Bank # B B50457-91901-01 Question History: Last NRC Exam  : None Question Cognitive Level: Comprehension or Analysis Difficulty 2 10 CFR Part 55 Content: (CFR: 41.10)

Comments:

Parent B50457-91901-01 Requal Exam bank

The Unit is in Mode 5 in Mid Loop Operation.

RHR Pump 1A is in service with a flowrate of 3000 gpm.

At 10:00 a.m. the MAB RPO reports to the Control Room that the RHR pump 1A was making a loud and oscillating "swishing" noise when he was in the area.

The Primary RO then notices that RHR pump 1A amps are beginning to swing much more that normal, and RCS Hot Leg level is indicating (+1.5) inches.

The Unit Supervisor has announced entry into the Loss of Residual Heat Removal off normal procedure.

The next action the operators should take is to .....

A. reduce 1A RHR pump flow to < 1500 gpm B. decrease RCS temperature until the noise and amperage oscillation stop C. stop the 1A RHR Pump, and start the 1B RHR pump D. reduce 1A RHR pump flow until the noise and ampere oscillation stop Answer: A JUSTIFICATION:

MISCINFO:

A. Incorrect - This choice is not procedure based and would not help to correct the problem.

B. Incorrect - same as 'A' above.

C. Correct - With RCS NR Hot Leg Level at -1 inches, procedure direction is to reduce RHR flow within limit per Addendum 2 (1500 qpm).

D. Incorrect - This answer is similar to 'C' but, due to specific procedure direction, 'C' is a more correct choice.

REFERENCES

1. 0POP04-RH-0001, Loss Of Residual Heat Removal (Rev. 7)

FILE ID: BE025010.HHB REV. 1 DATE: 09/05/95 TIME: 3 min.

New Author: Greg Chitwood SYSTEM DESIGNATOR: RH KSA IMPORTANCE LESSON PLAN IDA NUMBERS NUMBER RO SRO NUMBER TASK# OBJ#

1. 000025EA2.07 3.4 3.7 LOT504.57 A86700 A91901

2. 000025SG12 3.3 3.5

Examination Outline Cross-

Reference:

Level RO 000026 Loss of Component Cooling Tier # 1 Water / 8 Group # 1 K/A # 000026 G2.2.4 G2.4.4 Ability to recognize abnormal Importance Rating 4.5 indications for systems operating parameters that are entry level conditions for emergency and abnormal operating Question 4 Given the following conditions:

• Rx Power is at 100% Rated Thermal Power

• CCW Surge tank level 63%

• Annunciator CCW SURGE TANK Level Lo in

• A train CCW pump in operation

• B train CCW pump in Standby Based on the above conditions the entry conditions for _____(1)______ are met and results in

____(2)_______ .

A. (1) 0POP04-CC-0001 (Component Cooling Water Leak),

(2) AUTOMATIC closure "RCDT HX 1A(2A) INL MOV-0392" (CCW to RCDT Hx isolation)

B. (1) 0POP04-CC-0001 (Component Cooling Water Leak) and 0POP04-CV-0004 (Loss of Normal Letdown),

(2) AUTOMATIC closure "LETDN ORIF HDR ISOL FV-011" (Letdown Orifice Isolation Valve)

C. (1) 0POP04-CC-0001 (Component Cooling Water Leak),

(2) AUTOMATIC closure of "SUPPLY ISOL MOV 0768 (CCW Train A isol to charging pump header)

D. (1) 0POP04-CC-0001 (Component Cooling Water Leak) and 0POP04-CV-0004 (Loss of Normal Letdown),

(2) loss of the A train charging header.

Proposed Answer: A Explanation (Optional):

A. Correct, At 64.6% level on CCW Surge tank "RCDT HX 1A(2A) INL MOV-0392" (CCW to RCDT Hx isolation)will automatically close 63% level indication on the CCW surge tank is below the required entry condition for 0POP04-CC-0001 (Component Cooling Water Leak)

B. Incorrect, "LETDN ORIF HDR ISOL FV-011" does not automatically isolate, it is manually isolated in the Component Cooling water leak procedure C. Incorrect, AUTOMATIC closure of "SUPPLY ISOL MOV 0768 (CCW Train A isol to charging pump header) does not happen until CCW surge tank level is below 61%

D. Incorrect, loss of letdown is not required until the orifice isolations are manually closed and CCP header A is not isolated until 61% on CCW surge tank Technical Reference(s): 0POP04-CC-0001 Rev 14, 0POP04-CV-0004 Rev 11 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination:

Learning Objective: (As available)

Question Source: New X 2393 Question History: Last NRC Exam None Question Cognitive Level: Comprehension or Analysis Difficulty 3 10 CFR Part 55 Content: (CFR: 41.4)

Comments:

This meets the K/A for loss of component cooling as component cooling is lost to several loads cooled by the CCW system.

Examination Outline Cross-

Reference:

Level RO 000027, Knowledge of the operational Tier # 1 implications of the following concepts as Group # 1 they apply to Pressurizer Pressure Control K/A # 000027 AK 1.02 Malfunctions:

Importance Rating 2.8 AK1.02 Expansion of liquids as temperature increases RO2.8/SRO 3.1 Question 5 Given the following conditions on Unit 1 Initial Plant conditions:

• 100% reactor Power, normal plant operation

• RCS Pressure is 2205 psig

• Backup Heaters have been manually energized With no further operator actions, what is the expected plant response?

A. RCS Pressure will rise until the PORV cycles to lower RCS pressure; PZR Level will be lower after the PORV cycles.

B. Backup heaters will turn off when one the spray valves open and cycle; PZR will remain within its normal level band.

C. Backup heaters will automatically cycle off prior to the spray valves opening; the PZR level will rise slightly.

D. Backup Heaters will remain on and the RCS pressure will be controlled at 2235 psig.

Proposed Answer: D Explanation (Optional):

A. Incorrect, RCS pressure will rise and then will lower once the spray valves open B. Incorrect, there is not interlock between turning off the backup heaters and the spray valves opening but the PZR will remain in its normal operating band C. Incorrect, if the heaters were in automatic this would be a correct response, but with the heaters in manual they would not secure.

D. Correct, this is the alignment that is used when equalizing the boron between the pressurizer and the RCS.

Technical Reference(s): 0POP04-RP-0001, Loss of Automatic Pressurizer Pressure Control (Attach if not previously provided)

(Including version/revision number)

Proposed references to be provided to applicants during examination:

Learning Objective: (As available)

Question Source: New 2394 X Question History: Last NRC Exam Question Cognitive Level: Comprehension or Analysis Difficulty 3 10 CFR Part 55 Content: 55.41.8, 55.41.10 55.45.3 Comments:

Examination Outline Cross-

Reference:

Level RO Knowledge of the interrelations between Tier # 1 the, and the following an ATWS: Group # 1 EK2.06 Breakers, relays, and disconnects K/A # 000029 EK 2.06 Importance Rating 2.9 Question 6 An operator action of 0POP05-EO-FRS1, Response to Nuclear Power Generation - ATWS, is to Ensure 480V LC 1K1 (2K1) and 1L1 (2L1) feeder breakers open.

This step will de-energize power to the Rod Drive MG Set:

A. Motors, Opening only one of the breakers should be sufficient to cause a reactor trip.

B. Motors, Both breakers must be opened to cause a reactor trip.

C. By Opening Load Center breaker to Rod Drive Motor-Generator #1 and #2, Opening only ONE of the breakers should be sufficient to cause a reactor trip.

D. By Opening Load Center breaker to Rod Drive Motor-Generator #1 and #2, BOTH breakers must be opened to cause a reactor trip.

Proposed Answer: B Explanation (Optional):

A. Incorrect, while opening the 480 volt breakers will de-energize the motor it requires both breakers to be opened to cause the Rods to trip.

B. Correct C. Incorrect, opening the 480 volt breakers interrupts power to the Motor and the Load center Breakers will lose power but will not open due to this step they may open due to under voltage condition D. Incorrect, opening the 480 volt breakers interrupts power to the Motor.

Technical Reference(s): 0POP05-EO-FRS1 rev 17, Response to Nuclear Power Generation - ATWS (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination:

Learning Objective: (As available)

Question Source: Bank # 1577 Question History: Last NRC Exam Not used Question Cognitive Level: Comprehension or Analysis Difficulty 2 10 CFR Part 55 Content: 55.41.6 Comments:

Distractors C and D in parent were not credible due to the generators do not have feeder breakers

Bank 1577 An operator action of 0POP05-EO-FRS1, Response to Nuclear Power Generation - ATWS, is to Ensure 480V LC 1K1 (2K1) and 1L1 (2L1) feeder breakers open.

This step will de-energize power to the Rod Drive MG Set:

A. motors. Opening only one of the breakers should be sufficient to cause a reactor trip.

B. motors. Both breakers must be opened to cause a reactor trip.

C. generators. Opening only one of the breakers should be sufficient to cause a reactor trip.

D. generators. Both breakers must be opened to cause a reactor trip.

Answer B

Examination Outline Cross-

Reference:

Level RO Knowledge of the reasons for the following Tier # 1 responses as they apply to the SGTR: Group # 1 EK3.02 Prevention of secondary PORV K/A # 000038 EK3.02 cycling Importance Rating 4.4 Question 7 The control room operators are responding to the symptom of a SGTR, Control Board indications indicate that the 1D SG is the affected SG.

Step 3 of 0POP05-EO-EO30, Steam Generator Tube Rupture states: ADJUST ruptured SG(s)

PORV controller setpoint to BETWEEN 1260 PSIG AND 1265 PSIG.

1D Steam Generator is now reading ~ 1260 psig The PORV setpoint lowered to between 1260 and 1265 psig on the affected SG to ..

A. Stabilize the ruptured SG pressure and level to prevent an uncontrolled cooldown of the RCS.

B. Prevent steam generator overpressure due to overfilling of the ruptured steam generator.

C. Prevent an unmonitored release by keeping the PORV from lifting.

D. Minimize atmospheric releases.

Proposed Answer: D Explanation (Optional):

A. Incorrect, Lowering the set point would not prevent an uncontrolled RCS cooldown of the RCS, Plausible because if the there were no other SGs available feeding a intact SG with a tube rupture is a possible procedure flow path.

B. Incorrect, There is no indication in the stem that the affected SG is overfilled, Plasuable because feeding a SG with a tube rupture that is faulted can lead to an uncontrolled cooldown.

C. Incorrect, If the PORV did lift the release would be monitored D. Correct Technical Reference(s): 0POP05-EO-EO30, Steam Generator Tube Rupture LP 504.15. Rev 10 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination:

Learning Objective: LOT504.15, Objective: 92408 Question Source:

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Difficulty 2 10 CFR Part 55 Content: 55.41.5, 55.41.10 55.43.6, 55.43.13 Comments:

From E-3 Background document for step 3 of 0POP05-EO-EO30 Isolation of the ruptured steam generator(s) effectively minimizes release of radioactivity from this generator. In addition, isolation is necessary to establish a pressure differential between the ruptured and non-ruptured steam generators in order to cool the RCS and stop primary to secondary leakage.

Examination Outline Cross-

Reference:

Level RO 000040 (W/E12) Steam Line Rupture - Tier # 1 Excessive Heat Transfer / 4 Group # 1 Ability to operate and / or monitor the K/A # 000040 AA1.15 following as they apply to the Steam Line Importance Rating 3.9 Rupture:

AA1.15 T-ave. protection indicators Question 8 Consider a Main Steam Line Break and its associated cooldown.

Which of the following sets of initial plant conditions would result in the GREATEST challenge to Shutdown Margin?

Rx PWR (%) Tavg (°F) Burnup (MWD/M TU)

A. 100 592 18,500 B. 0 567 18,500 C. 100 592 150 D. 0 567 150

Proposed Answer: B Explanation (Optional):

A conservative analysis of the potential consequences of an excessive increase in secondary steam flow incident includes the following conservative assumptions.

• EOL no-load value for the

• Shutdown margin

• Moderator Temperature Coefficient Technical Reference(s): LOT501.16, R1, UFSAR Figure 15.1-11 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination:

Learning Objective: (As available)

Question Source: Bank # NRC 214 Question History: Last NRC Exam N/A used on Audit (2001)

Question Cognitive Level: Memory or Fundamental Knowledge Difficulty 2 10 CFR Part 55 Content: 55.41.5 Comments:

No changes were made to the parent question

Major Steam Line Break (continued)

Types of Accidents Analyzed The following cases have been analyzed in determining the core power and RCS transients:

• Complete severance of a pipe, with the plant initially at no-load conditions, full reactor coolant flow and offsite power available.

• The previous case with loss of offsite power (LOOP) simultaneous with the steam line break and initiation of the SI signal.

LOOP results in reactor coolant pump (RCP) coastdown.

Since the SGs are provided with integral flow restrictors with a 1.4 ft2 throat area, any rupture with a break area greater than 1.4 ft2, regardless of location, would have the same effect on the NSSS as the 1.4 ft2 break.

Assumptions UFSAR Figure 15.1-11 Assumptions used in the analysis are shown below.

Assumption Description Initial/Nominal Initial reactor power, T ave , and RCS pressure are assumed to be Conditions at their hot-shutdown nominal values.

• The reactor is shutdown with all rods inserted except for the most reactive RCCA which is stuck in the full out position.

Limiting A conservative analysis of the potential consequences of an Parameters excessive increase in secondary steam flow incident includes the following conservative assumptions.

• EOL no-load value for the Shutdown margin Moderator Temperature Coefficient Note: UFSAR Figure 15.1-11 shows the expected K eff responses for the temperature and pressure effects on the moderator temperature coefficient.

Examination Outline Cross-

Reference:

Level RO 000054 Loss of Main Feedwater / 4 Tier # 1 Ability to determine and interpret the Group # 1 following as they apply to the Loss of Main K/A # 000054 AA 2.03 Feedwater (MFW):

Importance Rating 4.1 AA2.03 Conditions and reasons for AFW pump startup Question 9 Given the following plant Conditions:

Rx Startup in progress Rx Power 4%

The Crew has just completed Addendum 10, Transferring Feed for AFW to MFW of 0POP03-ZG-005 Plant Startup to 100%.

SGFP-14 Startup Feed Pump is the only SGFP operating at this time Subsequently:

SGFP-14 trips for an unknown reason.

What is the INITIAL status of the AFW system?

A. AFW pumps running due to a AMSAC actuation signal.

B. AFW pump running due to an immediate automatic Reactor Trip signal.

C. AFW pumps are in standby and will start when the second Steam Generator meets the 2 of 4 logic for 20% level.

D. AFW pumps are in standby and will start when the First Steam Generator meets the 2 of 4 logic for 20% level.

Proposed Answer: D Explanation (Optional):

A. Incorrect but plausible, AMSAC initiated by 3 of 4 SG NR level < 15% [ channels III and IV - LT-518, 528, 537, & 547] when armed by 2 of 2 turbine high load (PT505 and 506) signals (~ 224 psig or ~ 30% turbine power) (corresponding to 40% Reactor power)

B. Incorrect but Plausible, Loss of all feed does not automatically start the AFW system and there is no immediate reactor trip.

C. Incorrect but plausible, AFW is auto started when the first SG reaches meets the 2 of 4 logic for Lo Lo level of 20% not the second D. Correct Technical Reference(s): 0POP03-ZG-005 Plant Startup to 100%., LOT202.01.HO1 Rev. 10 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination:

Learning Objective: (As available)

Question Source: New 2397 X Question History: Last NRC Exam Question Cognitive Level: Comprehension or Analysis Difficulty 3 10 CFR Part 55 Content: 55.41.4 Comments:

Copied from LOT202.01.HO1 Rev. 10 Trip Setpoints The direct trip setpoints applicable to the Steam Generator are for LO-LO level and HI-HI level.

These setpoints are 20% for LO-LO level and 87.5% for HI-HI level.

The LO-LO level setpoint provides a reactor trip and a Auxiliary Feedwater actuation signal for this level indication. The circuitry logic is (2/4) on any Steam Generator. The purposes of this trip and actuation is to protect the heat sink for the reactor, and prevent uncovering the S/G tubes (prevent thermal shock to the tubes).

Examination Outline Cross-

Reference:

Level RO 000055 Station Blackout / 6 Tier # 1 Knowledge of the operational implications Group # 1 of the following concepts as they apply to K/A # 055 G 2.4.1 the Station Blackout Importance Rating 4.6 2.4.1 Knowledge of EOP entry conditions and immediate action steps Question 10 Unit 2 has experinced a Reactor Trip. The Primary RO is performing the Immediate Actions of 0POP05-EO-EO00, Reactor Trip or Safety Injection. Steps 1 and 2 have been performed SAT and the following conditions exist for Step 3:

4.16KV ESF busses are NOT energized 480V ESF LCs are NOT energized 480V ESF MCCs are NOT energized No Emergency Diesel Generators are running Which of the following is the next correct action for the Primary RO to perform of the Immediate actions of 0POP05-EO-EO00, Reactor Trip or Safety Injection?

A. RESTORE power to all AC ESF busses prior to continuing in 0POP05-EO-EO00, REACTOR TRIP OR SAFETY INJECTION B. Immediately go to 0POP05-EO-EC01, LOSS OF ALL AC POWER RECOVERY WITHOUT SI REQUIRED, This procedure will establish natural circulation and attempt to recover the AC ESF busses.

C. RESTORE power to at least one AC ESF bus by EMERGENCY STARTING STBY DG then continue with remaining Immediate actions of 0POP05-EO-EO00.

D. Immediately go to 0POP05-EO-EC00, LOSS OF ALL AC POWER This procedure will attempt to restore power to the AC ESF busses.

Proposed Answer: C Explanation (Optional):

A. Incorrect, It is not necessary to restore all AC ESF busses prior to continuing with the Immediate actions of 0POP05-EO-EO00. This is plausible because the applicant may misunderstand the step that directs the operator to restore power to at least one AC ESF bus and restoring power to both would be optimal.

B. Incorrect, there is no immediate transfer to 0POP05-EO-EC01, The distractor is plausible because the applicant may eventually get to this procedure after transfer to 0POP05-EO-EC00 if power was not able to be restored C. Correct, Immediate action step 3 Response not obtained step a of 0POP05-EO-0000 requires this action.

D. Incorrect, The applicant would go to this procedure if Immediate action step 3 Response not obtained step a of 0POP05-EO-0000 is not successful in starting the STBY DG.

Technical Reference(s): OPOP05-EO-0000 Rev 23 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination:

Learning Objective: (As available)

Question Source:

New 2398 X Question History: Last NRC Exam None Question Cognitive Level: Comprehension or Analysis Difficulty 3 10 CFR Part 55 Content: 55.43.5, 55.41.10 55.45.13 Comments:

Examination Outline Cross-

Reference:

Level RO 000056 Loss of Off-site Power / 6 Tier # 1 Ability to operate and / or monitor the Group # 1 following as they apply to K/A # 000056 AA1.06 the Loss of Offsite Power:

Importance Rating 3.6 AA1.06 Safety injection pump Question 11 Given the following plant conditions:

• The plant was at full power when a loss of offsite power occurs

• Pressurizer pressure is 1850 psig

• Containment Pressure is 8 psig

• Offsite power has not been restored

• Emergency Diesels have just received there start signal For the above conditions, which one of the following actions will automatically occur?

A. HHSI pumps start 11 seconds after the Emergency Diesels have received there start signal B. HHSI pumps start 7 seconds after ESF Busses are energized C. HHSI pumps start ~ 17 seconds after the ESF busses are re-energized D. HHSI pumps start 6 seconds after the Emergency Diesel generators receive there start signal

Proposed Answer: B Explanation (Optional):

A. Incorrect, with a loss of offsite power all loads on the ESF busses will load shed and the diesel generator will re-energize the bus within ~10 seconds of the actuation. Mode III load sequence will begin 1 second after the diesel generator breakers have closed in repowering the bus. The HHSI pumps will start ~ 6 seconds later B. Correct, once the ESF bus is re-energized there is a 1 second delay for the mode III sequence to begin and the HHSI pumps will start 6 seconds later or 7 seconds after the ESF bus is re-energized.

C. Incorrect, once the ESF busses are energized (takes 10 seconds) an additional one second mode III sequence occurs and then the HHSI pumps will start in 6 seconds, (Total of 17 seconds to start the HHSU pumps from the time that the Emergency Diesels receive the start signal to re-energize the bus)

D. Incorrect, this is plausible because there is a 6 second delay after the mode III sequence has started to start the HHSI pumps, however the 6 seconds is after the bus has been re-energized and the mode III load sequence has started.

Technical Reference(s): LOT201.22 Rev. 15 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Difficulty 2 10 CFR Part 55 Content: 55.41.7 55.45.6, 55.45.5 Comments:

1.ESF Actuation 1.1 Electrical Power Distribution 1.1.1 Normal Operations A. Electrical power normally supplied from main turbine through auxiliary transformers via 3 separate, redundant ESF power trains.

B. Protected vital power supplied from:

1. Preferred Power - One or more circuits from the off-site transmission network.

2. Standby Power - 3 electrically and mechanically separate emergency diesel generators.

3. Diesel generators designed to accept a sequential loading of all assigned ESF equipment within 10 sec.

after receipt of an SI actuation signal.

1.1.2 ESF loading sequence: SI coincident with site blackout, Mode III A. All equipment operation on vital buses is terminated on loss of power by load shed signal.

B. Diesel generators start and connect to the vital buses within 10 seconds; vital power restored.

C. ESF loading sequence commences on DG breaker(s) closure.

D. Mode III sequence is the same as Mode I after the load center breakers are sequenced on at 1 second.

Mode III - Safety Injection and Loss of Offsite Power Times given are times after DG output breaker has closed.

Equipment Time A. 480 VAC Busses 1 sec B. HHSI pump 6 sec C LHSI pump 10 sec D. Cnmt spray pump

• 15 sec E. RCFCs 15 sec

F. CCW pump 20 sec G. ECW pump 25 sec H. AFW pump 30 sec I. CR HVAC 35 sec J. Ess. Chilled water pump 35 sec K. Cnmt spray pump

• 40 sec L. Ess. Chiller 240 sec

Examination Outline Cross-

Reference:

Level RO 000057 Loss of Vital AC Inst. Bus / 6 Tier # 1 Knowledge of the reasons for the following Group # 1 responses as they apply to K/A # 000057 AK3.01 the Loss of Vital AC Instrument Bus:

Importance Rating 4.1 AK3.01 Actions contained in EOP for loss of vital ac electrical instrument bus Question 12 Given the following:

• Unit 1 is performing a Mode 3 cooldown

• RCS is at 520 °F and 1800 psig

• A loss of power occurs on 120 VAC vital distribution panel DP- 1201

• The crew enters 0POP04-VA-0001, Loss of 120 VAC Class Vital Distribution

• A Plant Operator is directed to locally open the 1A SG PORV for temperature control 0POP04-VA-0001, cautions the Plant Operator to NOT open the SG PORV more than 50%.

Which ONE of the following is the reason for this caution?

A. Prevent excessive cooldown with limited RCS instrumentation available.

B. Minimize the mass loss out of the 1A SG with limited AFW capability.

C. Ensure the ability to reclose the 1A SG PORV following opening.

D. Limit the amount of positive reactivity due to cooldown.

Proposed Answer: C Explanation (Optional):

A. INCORRECT - Not the correct basis for the SG PORV stroke limit. Some RCS instrumentation will be de-energized, but not to the extent that monitoring capabilities will be jeopardized.

B. INCORRECT - Not the correct basis for the SG PORV stroke limit. Automatic control of the AFW regulating valve will be lost, but manual control is still available.

C. CORRECT - This is the correct basis for the SG PORV stroke limit. There is only sufficient stored energy in the SG PORV accumulators for only one and one half full strokes of the valve. The caution limits the stroke of the SG PORV to ensure that the valve can be closed.

D. INCORRECT - Not the correct basis for the SG PORV stroke limit. A low steam line pressure SI exists, but under the given conditions it would have already been blocked by procedure.

Technical Reference(s): 0POP04-VA-0001 Rev 30, Addendum 9 Caution Step 1 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: 92109 Question Source: Bank NRC 28 (Note changes or attach parent)

Question History: Last NRC Exam 1999 Question Cognitive Level: Comprehension or Analysis Difficulty 3 10 CFR Part 55 Content: 55.41.5, 55.41.10 55.45.6, 55.45.13 Comments:

Changed distractor D from (Avoid an inadvertent safety injection actuation on low steam line pressure) because it was not credible as the SI signal would have already been blocked by procedure Note from 0POP4-VA-0001 Rev 30 appendix 9 NOTE

There is sufficient stored energy in the PORV hydraulic unit accumulators for only one and one half strokes. SG PORVs should NOT be opened GREATER THAN 50%.

IF the S G P O RV wa s ope n a nd powe r is los t to hydra ulic pum p, THEN the a m ount of available stroke is reduced and the SG PORV available stroke COULD be limited to only one half stroke.

NRC Bank Question 28 Given the following:

• Unit 1 is performing a Mode 3 cooldown

• RCS is at 520 °F and 1800 psig

• A loss of power occurs on 120 VAC vital distribution panel DP-001

• The crew enters 0POP04-VA-0001, Loss of 120 VAC Class Vital Distribution

• A Plant Operator is directed to locally open the 1A SG PORV for temperature control 0POP04-VA-0001, cautions the Plant Operator to NOT open the SG PORV more than 50%.

Which ONE of the following is the reason for this caution?

A. Prevent excessive cooldown with limited RCS instrumentation available.

B. Minimize the mass loss out of the 1A SG with limited AFW capability.

C. Ensure the ability to reclose the 1A SG PORV following opening.

D. Avoid an inadvertent safety injection actuation on low steam line pressure.

Examination Outline Cross-

Reference:

Level RO SRO 000058 Loss of DC Power / 6 Tier # 1 Knowledge of the reasons for the following Group # 1 responses as they apply to K/A # 000058 AK3.02 the Loss of DC Power:

Importance Rating 4.0 4.2 AK3.02 Actions contained in EOP for loss of dc power Question 13 An electrical fault has occurred on Unit 1 that de-energized Train A Class 1E 125 VDC Power.

OP0P04-DJ-001, Loss of Class 1E 125 VDC Power, Addendum 1 is in progress. Step 7 reads as follows:

What is the reason for placing the control switches in the required positions?

A. This action immediately re-positions the components to their normal shutdown position after they repositioned due to the loss of the bus.

B. This action re-positions the components to their normal operating alignment when power is restored.

C. This action will ensure the components will not inadvertently reposition when power is restored.

D. This action will ensure that the components will reposition to their normal shutdown alignment when power is restored.

Proposed Answer: C Explanation (Optional):

A. Incorrect, there is no power to the relay racks and the valves will not reposition B. Incorrect, This is not the normal operating position for these components C. CORRECT per the Note in Addendum 1 just prior to step 7 OP0P04-DJ-001, Loss Of Class 1E 125 VDC Power, Addendum 1 D. Incorrect, the intent is to not allow the components to reposition when power is restored Technical Reference(s): OP0P04-DJ-001, Loss Of Class 1E 125 VDC Power, Addendum 1, Note on page 17 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective:

Question Source: New X Question History: Last NRC Exam N/A Question Cognitive Level: Comprehension or Analysis Difficulty 3 10 CFR Part 55 Content: 55.41.5, 55.41.10 55.45.1, 55.45.6 Comments:

OP0P04-DJ-001, Loss Of Class 1E 125 VDC Power, Addendum 1, Note on page 17

Examination Outline Cross-

Reference:

Level RO 000062 Loss of Nuclear Svc Water / 4 Tier # 1 Ability to determine and interpret the Group # 1 following as they apply to the Loss of K/A # 000062 AA2.01 Nuclear Service Water (SWS):

Importance Rating 2.9 AA2.01 Location of leak in the SWS Question14 Unit 1 is operating at 100% power when a leak develops in the Component Cooling Water System. 0POP04-CC-0001, Loss of Component Cooling Water, is entered.

Level in the CCW surge tank is 63% and decreasing slowly.

Which of the following components is a potential source of the leak?

A. Letdown Heat Exchanger (Hx)

B. Reactor Coolant Pump 1C motor air cooler C. Reactor Coolant Drain Tank Hx D. Excess Letdown Hx

Proposed Answer: B Explanation (Optional):

A. Incorrect; CCW to the Letdown Hx is isolated on the first level isolation (@64.6%)

B. Correct; CCW to and from all the RCPs is isolated on the second level isolation

(@61.5%), which has not been reached yet.

C. Incorrect; CCW to the RCDT Hx is isolated on the first level isolation (@64.6%)

D. Incorrect; CCW to the Excess Letdown Hx is isolated on the first level isolation

(@64.6%)

Technical Reference(s): 0POP04-CC-0001, Rev 3, step 1; 9F05017, Rev 19; 9F05020, Rev 16; 9F05021, Rev 12 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: Bank # NRC 1 Question History: Last NRC Exam N/A Question Cognitive Level: Comprehension or Analysis Difficulty 3 10 CFR Part 55 Content: 55.41.4 55.45.13 Comments:

Parent:

Unit 1 is operating at 100% power when a leak develops in the Component Cooling Water System. 0POP04-CC-0001, Loss of Component Cooling Water, is entered.

• Level in the CCW surge tank is 63% and decreasing slowly.

Which ONE of the following components is a potential source of the leak?

A. Letdown Heat Exchanger (Hx)

B. Reactor Coolant Pump 1C motor air cooler C. Reactor Coolant Drain Tank Hx D. Excess Letdown Hx Correct answer B From Addendum 1 of 0POP4-CC-0001 Rev 15

1. WHEN CCW Surge Tank level is less than 64.6%, THEN the following valves will automatically close:

"NNS LO ADS INL IS O L MO V -0235" (CCW to common supply header isolation)

"NNS LO ADS INL IS O L MO V -0236" (CCW to common supply header isolation)

"BRANCH IS O L MO V -0297" (CCW to RCDT Hx and Excess LD Hx isolation)

"RCDT HX 1A(2A) INL MO V -0392" (CCW to RCDT Hx isolation)

"EXCES S LETDOWN HX 1A(2A) INL MO V -0393" (CCW to Excess LD isol)

2. WHEN CCW Surge Tank level is less than 61.5%, THEN the following valves will automatically close:

Charging System Components "S UP P LY IS O L MO V -0768"(CCW Train A isol to charging pump header)

"RET IS O L MO V -0772"(CCW return header isolation from charging pumps to Train A)

"S UP P LY X -CONN FV-4656"(CCW to charging pump A crosstie)

"RETURN X-CONN FV-4657"(CCW return from charging pump A crosstie)

Examination Outline Cross-

Reference:

Level RO W/E04 LOCA Outside Containment / 3 Tier # 1 2.1.28 Knowledge of the purpose and Group # 1 function of major system components and K/A # W/E04 2.1.28 controls.

Importance Rating 3.2 Question15 Given the following:

• Unit 1 is in Mode 3

• RCS pressure 800 psig

• Pressurizer level is lowering rapidly

• Containment sump levels are NOT rising The CRS has entered 0POP04-RC-0006, Shutdown LOCA. One major action accomplished by this procedure is to ____(1)____ in order to _____(2)______.

A. (1) Isolate letdown, (2) protect the Reactor Core B. (1) Maintain Pressurizer level, (2) Minimize the amount of contaminated water generated C. (1) Secure RCPs, (2) prevent damaging the motors.

D. (1) Stabilize RCS temperature and pressure, (2) Prepare for an orderly cooldown in accordance with the normal operating procedures.

Proposed Answer: A Explanation (Optional):

A. Correct, Listed in 0POP04-RC-0006 , Shutdown LOCA Rev 15 B. Incorrect, while minimizing waste is a good thing it is not the concern here C. Incorrect, RCP operation would be secured based on possible damage to the RCP seals. Credible because the student might believe that any containment isolation could involve CCW to the RCPs.

D. Incorrect, stabilization for transfer to Normal Operating procedures is not the correct course; the off normal procedure directs the cooldown.

Technical Reference(s): 0POP04-RC-0006, Shutdown LOCA Rev 17 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: New 2401 X Question History: Last NRC Exam N/A Question Cognitive Level: Comprehension or Analysis Difficulty 3 10 CFR Part 55 Content: 55.41.10 Comments:

Examination Outline Cross-

Reference:

Level RO W/E11 Loss of Emergency Coolant Tier # 1 Recirc. / 4 Group # 1 Knowledge of the operational implications K/A # W/E11 EK1.1 of the following concepts as they apply to Importance Rating 3.7 the (Loss of Emergency Coolant Recirculation)

EK1.1 Components, capacity, and function of emergency systems.

Question16 A large break loss of coolant accident has occurred in Unit 2.

The following conditions exist:

• RWST level is 32,500 gallons (6%) and lowering.

• HHSI pumps have power available

• LHSI pumps have power available

• HHSI mini flow valves have power indication

• LHSI mini flow valves have power indication

• RWST to SI suction header has power indication

• Cold leg injection valves have power available

• Containment sump to SI suction header valves have no light indication

• Attempts made to regain power to the Containment Sump SI suction Header valves have NOT been successful Per Procedure 0POP05-E0-EC11 Loss of Emergency Coolant Recirculation what actions should be taken?

A. The HHSI and LHSI pumps should be throttled and makeup to the RWST should be initiated.

B. Secure all but HHSI and LHSI pump and continue to use the remaining volume of the RWST.

C. The HHSI and LHSI pumps are required to be secured until RWST level is restored to >

32,500 gallons (6%) then restarted.

D. The HHSI and LHSI pumps are to be placed in pull to lock and makeup from the VCT via a CCP should be used for inventory makeup until recirculation can be established or directed by TSC.

Proposed Answer: D Explanation (Optional):

A. Incorrect, while it seems reasonable to add water to the RWST there is no procedural guidance to commence makeup to the RWST in this situation B. Incorrect, there is no procedural guidance to perform this action, the 6% level directs you to secure feeding from the RWST.

C. Incorrect, 6% RWST level is the required level to stop the pumps that are taking suction on the RWST but there is no procedural guidance to restart the pumps after level has been raised to > 6%, this action would be directed from the TSC D. Correct, per 0POP05-E0-EC11 Loss of Emergency Coolant Recirculation, steps 29 and 30 direct the operator to place the pumps taking suction on the RWST in Pull to Lock and initiate charging from the VCT and make adds to that tank to maintain level > 3% to prevent auto swap over to the RWST Technical Reference(s): 0POP05-E0-EC11 Rev 18, loss of emergency coolant recirculation (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination:

Learning Objective: (As available)

Question Source: New 2402 X Question History: Last NRC Exam N/A Question Cognitive Level: Comprehension or Analysis Difficulty 3 10 CFR Part 55 Content: 55.41.8, 55.41.10 55.45.3 Comments:

Reference Examination Outline Cross-

Reference:

Level RO W/E05 Inadequate Heat Transfer - Loss Tier # 1 of Secondary Heat Sink / 4 Group # 1 Knowledge of the interrelations between K/A # W/E05 EK2.2 the (Loss of Secondary Heat Sink) and the Importance Rating 3.9 following:

EK2.2 Facilitys heat removal systems, including primary coolant, emergency coolant, the decay heat removal systems, and relations between the proper operation of these systems to the operation of the facility.

Question 17 The following plant conditions exist:

• Unit 1 reactor trip and SI have occurred.

• ALL SG levels are BELOW the narrow range.

• 1B AFW pump is OOS.

• Due to other AFW malfunctions, total AFW flow AVAILABLE is 350 gpm.

Which ONE of the following statements describes the basis for stopping the RCPs under these conditions?

A. Reduces reactor coolant inventory loss by reducing seal leak off.

B. Minimizes the possibility of a tube rupture as cooler AFW is injecting into the SGs.

C. Conserves SG secondary inventory by reducing heat input to the RCS.

D. Minimizes the time for the PORV to open and establish Feed and Bleed Cooling.

Proposed Answer: C Explanation (Optional):

A. Incorrect, securing the RCPs will not conserve inventory by reducing leak off B. Incorrect, This is not a reason for securing the RCPs having cold water impact with a dry SG is a concern but not for the step given C. Correct D. Incorrect, the purpose of securing the RCPs early in the procedure is to extend the time available to the operators to recover feedwater and not have to enter feed and bleed cooling Technical Reference(s): FR-H.1 Background Document, 14800044QG Rev D (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination:

Learning Objective: (As available)

Question Source: Bank NRC 787 (Note changes or attach parent)

Question History: Last NRC Exam From NRC exam Bank not previously used Question Cognitive Level: Comprehension or Analysis Difficulty 3 10 CFR Part 55 Content: 55.41.14 Comments:

Reference:

Parent NRC 787 The following plant conditions exist:

• Unit 1 reactor trip and SI have occurred.

• ALL SG levels are BELOW the narrow range.

• 1B AFW pump is OOS.

• Total AFW flow AVAILABLE is 350 gpm Which ONE of the following statements describes the basis for stopping the RCPs under these conditions?

A. Reduces reactor coolant inventory loss by reducing seal leak off.

B. Conserves SG secondary inventory by reducing heat input to the RCS.

C. Minimizes the possibility of a tube rupture as AFW is restored to the SGs.

D. Increases safety injection flow by decreasing RCS cold leg pressure.

Proposed Answer: B

Examination Outline Cross-

Reference:

Level RO 000077 Generator Voltage and Electric Tier # 1 Grid Disturbances / 6 Group # 1 Knowledge of the interrelations between K/A # 000077 AK2.0.1 Generator Voltage and Electric Grid Importance Rating 3.1 Disturbances and the following:

AK 2.0 1 Motors Question 18 A major power outage has occurred in South Texas, Several power plants have tripped off line and switchyard voltage has lowered due to the ongoing Electric Grid Disturbance.

Load Tap Changers have failed to bring voltage back to normal in the Unit.

Unit 1 experiences a Reactor Trip due to a Large Break Loss of Coolant Accident.

All required ECCS components started as designed.

The Emergency Diesels are not connected to any bus Compared to starting at normal grid voltage the current used by ECCS pumps is __________

and the potential for Pump Motor damages is more likely due to __________.

A. Lower, Slower start times resulting in bearing damage B. Lower, Higher Temperatures in the motor windings C. Higher, Faster start times resulting in bearing damage D. Higher, Higher Temperatures in motor windings

Proposed Answer: D Explanation (Optional):

Lower voltage results in higher current flow to operate motors Power = Voltage

• Current therefore with the power of the pumps remaining the same and a lower voltage the current will need to be higher The R of the Motor is constant, the heat generated in the motor will go up based on the I^2R losses. This higher level of heat is more likely to break down the insulation in the motor windings.

This concept makes D the only correct choice for the given conditions Technical Reference(s): 0POP04-AE-0005, Offsite Power System Degraded Voltage REV 9 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: New 2403 X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Difficulty 2 10 CFR Part 55 Content: 55.41.4, 55.41.5, 55.41.7, 55.41.10 55.45.8 Comments:

Reference:

Examination Outline Cross-

Reference:

Level RO 000033 Loss of Intermediate Range NI / 7 Tier # 1 Knowledge of the reasons for the following Group # 2 responses as they apply to the Loss of K/A # 000033 AK3.02 Intermediate Range Nuclear Importance Rating 3.6 Instrumentation:

AK3.02 Guidance contained in EOP for Question19 A reactor startup is in progress and power is being escalated to place the turbine on line.

Reactor Power is indicating 0% when NI-35 fails low off scale With NI-35 failed low off scale the power escalation should be ____________ .

A. Stopped below 10% power until NI-35 is restored to perform a channel check between the IR and the PR Nuclear instruments B. Stopped and the reactor should be shut down until NI-35 is returned to operable status C. Continued as the PR instruments are on scale and you have good indication of reactor power.

D. Continued until reactor power is at 15% and then perform a channel check with the remaining intermediate range NI

Proposed Answer: A Explanation (Optional):

A. Correct 0POP04-NI-0001, Nuclear Instrument Malfunctions addendum 2 step 3 directs the not to exceed 10% power until the inoperable channel is restored.

B. Incorrect, there is no requirement to trip unless the IR instrument has failed high C. Incorrect, this violates the requirements of 0POP04-NI-001 D. Incorrect, there is no guidance to perform a channel check with only one IR instrument.

Technical Reference(s): 0POP04-NI-0001, Nuclear Instrument Malfunctions (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: New X Question History: Last NRC Exam N/A Question Cognitive Level: Comprehension or Analysis Difficulty 3 10 CFR Part 55 Content: 55.41.5, 55.41.10 55.45.6, 55.45.13 Comments:

Examination Outline Cross-

Reference:

Level RO 000036 (BW/A08) Fuel Handling Accident Tier # 1

/ 8 Ability to operate and / or monitor the Group # 2 following as they apply to the Fuel K/A # 000036 AA1.03 Handling Incidents:

Importance Rating 3.5 AA1.03 Reactor building containment evacuation alarm enable switch.

Question 20 Unit 2 is performing a core reload at the end of a refueling outage when a malfunction of the fuel bridge occurs allowing a fuel bundle to drop on top of several previously installed fuel assemblies.

The fuel bridge operator notices a large quantity of bubbles coming up from the area where the fuel assembly has landed. The fuel bridge radiation monitor is in alarm and the bridge operators have notified the control room.

The following radiation monitors are in HIGH alarm:

RT-8011 RCB Atmosphere RT-8012 RCB Purge Ventilation Monitor RT-8013 RCB Purge Ventilation Monitor RT-8099 Refueling Machine Area Monitor As a control room operator the containment evacuation alarm will (1) and the containment purge system (2).

A. Automatically sound due to RT-8011 in HIGH alarm, should have isolated based on a Containment Ventilation Isolation signal from RT-8099 and either RT-8012 or RT-8013.

B. Need to be manually actuated due to RT-8011 in high alarm, should have isolated based on a Containment Ventilation Isolation signal from RT-8012 and RT-8013.

C. Automatically sound due to RT-8099 in high alarm, will remain running as the purge system is needed to mitigate the alarming condition.

D. Need to be manually actuated due to RT-8012 and RT-8013 in HIGH alarm, will need to be manually secured in response to alarming Radiation monitors RT-8012 and RT-8013.

Proposed Answer: B Explanation (Optional):

A. Incorrect, Containment Evacuation Alarm must be Initiated per 0POP04-RA-001, Containment Ventilation Isolation signal is derived from RT-8012 and RT-8013 not RT-8099.

B. Correct, 0POP04-RA-001 requires you to initiate the reactor building evacuation alarm and Containment Building Ventilation System RT-8012 & 8013 -High radiation in the RCB Purge System Exhaust sends a signal to the Solid State Protection System (SSPS) for Containment Ventilation Isolation (CVI).

C. Incorrect, Alarm does not automatically sound D. Incorrect, Containment Building Ventilation System RT-8012 & 8013 -High radiation in the RCB Purge System Exhaust sends a signal to the Solid State Protection System (SSPS) for Containment Ventilation Isolation (CVI).

Technical Reference(s): 0POP04-RA-0001, Radiation Monitoring System Alarm Response (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: New X Question History: Last NRC Exam N/A Question Cognitive Level: Comprehension or Analysis Difficulty 3 10 CFR Part 55 Content: 55.41.11 Comments:

References from 0POP04-RA-0001, Radiation Monitoring System Alarm Response

Examination Outline Cross-

Reference:

Level RO 000037 Steam Generator Tube Leak / 3 Tier # 1 Ability to operate and / or monitor the Group # 2 following as they apply to the Steam K/A # 000037 AA1.06 Generator Tube Leak:

Importance Rating 3.8 AA1.06 Main steam line rad monitor meters Question 21 The control room has the following indication at 100% power RT-8130B, Primary to Secondary Leak Rate Monitor, indicates a leak rate of 2000 gpd The actions for a steam generator tube leak have been completed and the plant has been shut down for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and is cooling down on the intact steam generators preparing to transition to RHR.

Based on the above plant conditions what should the reading for RT-8130B, Primary to Secondary Leak Rate Monitor be?

A. Higher, with the unit off line the steam is not flowing as fast past the monitors radiation is exposed to the monitor for a longer period of time making the indications read higher B. Lower, with the unit off line there is not as many radioactive particles in the steam and the monitor will be reading lower C. Higher, as time has passed the contamination in the main steam line has risen making the monitor read higher D. Lower, with the reactor shutdown the production of N-16 has stopped and the monitor will read lower.

Proposed Answer: D Explanation (Optional):

A. Incorrect, the monitor is a N-16 monitor and without the reactor operating N-16 population will not be present 60 min after shutdown B. Incorrect, this monitor reads N-16 and will be affected by the N-16 radiation, not radioactive particles C. Incorrect, the contamination of the steam lines will not result in a higher reading D. Correct, RT-8130 is the N-16 monitor and with the reactor shut down the production of N-16 has lowered significantly (~0) and the N-16 that was present while at power has decayed away.

Technical Reference(s): LOT202.02.HO.01 Rev.10, LOT202.41H0.01 Rev.15 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: New 2406 X Question History: Last NRC Exam N/A Question Cognitive Level: Comprehension or Analysis Difficulty 3 10 CFR Part 55 Content: 55.41.7 55.45.5 55.45.6 Comments:

References:

N16 Radiation monitors (RT-8130, 8131, 8132, 8133) are installed to monitor primary to secondary leakage.

Detector enclosed in SS sleeve and placed on top of each steam line above the non-1E chillers.

N16 Processing Unit Cabinet (ZLP-950) located just inside Cold Chemistry Lab (TGB 29').

Part of Gen-Aid System (enhanced main generator monitoring).

Steam generator leakage monitoring screens assessable from ICS (Integrated Computer and Gen-Aid Computers.

The processing unit uses a steam generator specific algorithm to calculate and display the activity linked to the N-16 measurement window, that is the photons striking the detector with between 4.5 Mev and 7 Mev of energy. This calculation is weighted by a reactor power value (0% to 120%). The final value produced by the processor is gallons per day of primary to secondary leakage. The processing unit generally provides three digital outputs and one analog 4 to 20mA output. The digital outputs are alarms: alert, high, and high-high. The analog output is gallons of primary to secondary leakage per day.

Examination Outline Cross-

Reference:

Level RO 000051 Loss of Condenser Vacuum / 4 Tier # 1 Ability to determine and interpret the Group # 2 following as they apply to K/A # 000051 AA2.02 the Loss of Condenser Vacuum:

Importance Rating 3.9 AA2.02 Conditions requiring reactor and/or turbine trip.

Question 22 Unit 1 Coming out of a refueling outage, the Generator output breaker is closed and reactor power is currently at 25% holding power while I& C trouble shoot a problem with the condenser vacuum instrumentation.

During the troubleshooting actual condenser vacuum drops to 20.5 inches HG and stabilizes.

What is the expected plant response?

Assume all automatic actions occur as designed with no operator action A. The Turbine will Trip, The Reactor will Trip, MFW pumps will trip due to the reactor trip B. The Turbine will Trip, The Reactor not Trip and MFW pumps will remain running C. The Turbine will not Trip, The Reactor will Trip and the MFW pumps will Trip D. The Turbine will not Trip, the Reactor will not Trip and MFW pumps will remain running

Proposed Answer: B Explanation (Optional):

A. Incorrect, the trip set point for Unit 1 is 21 HGV, plausible because below 10% power the reactor will not trip when a turbine trip occurs B. Correct C. Incorrect, the MFW pumps will not trip on low vacuum under these conditions SGFP trip setpoint is 14 inches HG D. Incorrect, turbine trip setpoint on Unit 1 is 21 HGV Plausible because there is a step to manually trip turbine at 22 in the loss of condenser vacuum procedure.

Technical Reference(s): LOT202.03.HO.01 Rev. 13, LOT202.14.HO.01 Rev.12 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: New 2407 X Question History: Last NRC Exam N/A Question Cognitive Level: Comprehension or Analysis Difficulty 3 10 CFR Part 55 Content: 55.41.6 Comments:

LOT202.03.HO.01 Rev. 13 Turbine Trips and Setpoints Electrical overspeed trip ---------------- 110% rated speed (1980 rpm)

Mechanical overspeed trip ------------- 111% rated speed (1998 rpm)

Condenser low-vacuum trip ------------21.0 Inches HG Excessive thrust-bearing wear trip -- + 40 mils (from reference)

Low bearing oil pressure trip ---------- 6.0 psig Steam generator Hi-Hi level ----------- 87.5%

Stator Cooling Water Low D/P -------- 19 PSID/40 sec Generator trip Reactor trip Safety injection

AMSAC Manual turbine trip from control room Manual turbine trip at turbine LOT202.14.HO.01 Rev.12 MFWP Low vacuum trip - 14" hg vacuum, decreasing, or 3 psig, increasing. When the turbine is latched and vacuum is less than 20" Hg, the low vacuum trip is automatically set at 3 psig.

When the vacuum increases to 20" Hg, the low vacuum trip is automatically reset to 14" Hg vacuum, decreasing.

Examination Outline Cross-

Reference:

Level RO 000059 Accidental Liquid RadWaste Rel. / Tier # 1 9 Group # 2 K/A # 000059 2.1.32 2.1.32 Ability to explain and apply all Importance Rating 3.4 system limits and precautions.

Question 23 Plant Operators are performing a transfer of the Floor Drain Tank to the Waste Monitor Tank per 0POP02-WL-0003, Floor Drain Tank Operation.

In accordance with 0POP02-WL-0003, Floor Drain Tank Operation, which of the following is a condition that could occur if procedural steps are performed improperly?

A. Contamination of non-contaminated systems B. Exceeding STP airborne limits C. Contamination of personnel D. Exceeding STP personnel dose limits

Proposed Answer: A Explanation (Optional):

A. CORRECT: During this evolution if improper valve lineup is performed the non radioactive systems could be contaminated by the Floor Drain Tank.

B. INCORRECT: Not a condition cautioned by the procedure and performing this evolution will not cause airborne limits to be exceeded.

C. INCORRECT: Not a condition cautioned by the procedure and since this is a closed system personnel will not become contaminated.

D. INCORRECT: Not a condition cautioned by the procedure and personnel dose limits will not be exceeded while performing this evolution.

Technical Reference(s): LOT 203.1, 0POP02-WL-0003 Rev 11 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: T20311 Question Source: Bank # NRC 470 Question History: Last NRC Exam NRC Class 16 Question Cognitive Level: Memory or Fundamental Knowledge Difficulty 2 10 CFR Part 55 Content: 55.41.11 Comments:

Distractor B and D do not go together as you can exceed your personnel dose limit and not exceed the airborne limits, and airborne limits can be exceeded without exceeding the dose limit Candidates must understand that when operating these portions of the LWPS the potential for contamination of non-radioactive systems exist. Several years ago contamination of a clean system actually occurred at STP, while operating the LWPS a improper valve lineup caused contamination of the Aux Steam System.

0POP02-WL-0003 Rev 11 4.0 Notes and Precautions 4.1 The maximum allowed filter differential pressure is 35 psid.

4.2 Changes in flow rates or pressures SHOULD be performed gradually to prevent filter damage.

4.3 Any activity that requires entry or access to a Radiologically Controlled Area (RCA), requires a Radiation Work Permit.

4.4 Improper performance of steps in this procedure could result in cross-contamination of non-contaminated systems (Reference 2.2.4).

4.5 The normal demineralizer alignment is LWPS Aux Demin upstream and WE Cond Demin downstream.

4.6 The lineup for some valves listed in Lineup 1, Valve Lineup, is performed per 0POP02-XM-0001, Mechanical Auxiliary Building Valve Pit Lineup. The valves lined up per 0POP02-XM-0001 are listed in Lineup 1, Valve Lineup, for reference. When performing Lineup 1, Valve Lineup, these valves do not have to be lined up. Credit may be taken for the last performance of a particular room lineup performed per 0POP02-XM-0001 or this procedure.

ORIGINAL QUESTION:

Plant Operators are performing a transfer of the Floor Drain Tank to the Waste Monitor Tank per 0POP02-WL-0003, Floor Drain Tank Operation.

Which ONE of the following is a condition that the procedure cautions could occur if procedural steps are performed improperly?

A. Contamination of non-contaminated systems B. Exceeding STP airborne limits C. Contamination of personnel D. Exceeding STP personnel dose limits

Examination Outline Cross-

Reference:

Level RO 000060 Accidental Gaseous Radwaste Tier # 1 Rel. / 9 Group # 2 Knowledge of the operational implications K/A # 000060 AK1.02 of the following concepts as they apply to Importance Rating 2.5 Accidental Gaseous Radwaste Release:

AK1.02 Biological effects on humans of the various types of radiation, exposure levels that are acceptable for personnel in a nuclear reactor power plant; the units used for radiation intensity measurements and for radiation exposure levels.

Question 24 An accidental leak in the waste gas system has occurred. Two operators are required to enter an area of the plant to isolate the waste gas leak. RP has determined that the area the team will be in is considered a high airborne area with an airborne concentration of 1 DAC/HR. The general area that the team will be in is 50 mrem/hr.

It will take the team 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to complete the necessary valve alignments to properly isolate the leak. What is the CEDE that the two workers will receive for the task and what is the likely biological effects of this entry?

A. 210 mrem CEDE, some small increase in the probability in biological both Stochastic and Non-Stochastic effects.

B. 204 mrem CEDE, some small increase in the probability in biological both Stochastic and Non-Stochastic effects.

C. 210 mrem CEDE, significant increase in the probability in biological both Stochastic and Non-Stochastic effects D. 204 mrem CEDE, significant increase in the probability in biological both Stochastic and Non-Stochastic effects

Proposed Answer: A Explanation (Optional):

A. Correct, general area equals 50 mrem/hr with 4 man hours in exposed area equals 200 mrem exposure, 1 DAC equals an exposure of 2.5 mren/hr with 4 manhours of exposure equals 10 mrem. Therefore by summing the exposure received 210 mrem is received by the team. An exposure of 210 mrem is below the occupational limits and while all radiation exposure has some potential to raise the probability of biological effects it should be minimal.

B. Incorrect see A above, the 204 mrem is plausible if the candidate assumed that 1 DAC/hr is equivalent to 1 mrem per hour.

C. Incorrect see A above D. Incorrect see A and B Above Technical Reference(s): 10 CFR 20.1004, 10 CFR 20.1204 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: New 2408 X Question History: Last NRC Exam N/A Question Cognitive Level: Comprehension or Analysis Difficulty 3 10 CFR Part 55 Content: 55.41.8 55.41.10 55.45.3 Difficulty: 3 Comments:

10 CFR 20.1204 (h)(1) In order to calculate the committed effective dose equivalent, the licensee may assume that the inhalation of one ALI, or an exposure of 2,000 DAC-hours, results in a committed effective dose equivalent of 5 rems (0.05 Sv) for radionuclides that have their ALIs or DACs based on the committed effective dose equivalent.

10 CFR 20.1004

Rem is the special unit of any of the quantities expressed as dose equivalent. The dose equivalent in rems is equal to the absorbed dose in rads multiplied by the quality factor (1 rem=0.01 sievert).

Examination Outline Cross-

Reference:

Level RO APE 061 AK2 Knowledge of the Tier # 1 interrelations between the Area Radiation Group # 2 Monitoring (ARM) System Alarms and the K/A # AK2.01 following:

Importance Rating 2.5 AK2.01 Detectors at each ARM system Question 25 Each detector in the Area Radiation Monitoring (ARM) System has an RM-80 Microprocessor Unit associated with it that is located in the vicinity of its associated detector.

Which of the below correctly lists functions of the RM-80 Microprocessor?

1. allow the detector to be source checked.

2. provide ALERT/HIGH alarm signals to the Rad Monitoring Console (RM-11) in the Control Room.

3. automatically actuate plant equipment (e.g. fans, valves, etc.) on an ALERT or HIGH alarm condition.

4. indicate local radiation levels from the detector.

A. 1, 2, 3 B. 1, 2, 4 C. 1, 3, 4 D. 2, 3, 4

Proposed Answer: B Explanation (Optional): There are no automatic functions associated with the ARM system.

Technical Reference(s): LOT 202.41 Rev 15, 0POP04-RA-0001 Rev 30 Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: Bank # 1685 Question History: Last NRC Exam Audit exam 2009 Question Cognitive Level: Memory or Fundamental Knowledge Difficulty 2 Question Level of Difficulty: 2 10 CFR Part 55 Content: 55.41.7 55.45.7 Comments:

Examination Outline Cross-

Reference:

Level RO W/E02 SI Termination - Knowledge of the Tier # 1 operational implications of the following Group # 2 concepts as they apply to the (SI K/A # W/E02, EK1.2 Termination)

Importance Rating 3.4 EK1.2 Normal, abnormal and emergency operating procedures associated with (SI Termination).

Question 26 In which of the following procedures is RCS pressure included in the SI termination criteria?

A. 0POP05-EO-EO20, Faulted Steam Generator Isolation B. 0POP05-EO-EC11, Loss of Emergency Coolant Recirculation C. 0POP05-EO-EC33, SGTR Without Pressurizer Pressure Control D. 0POP05-EO-FRP1, Response to Imminent Pressurized Thermal Shock Condition

Proposed Answer: A Explanation (Optional):

A. CORRECT. Faulted Steam Generator Isolation procedure, step 6, requires RCS pressure greater than 1745 psig and stable or rising to terminate SI.

B. INCORRECT. Loss of Emergency Coolant Recirculation procedure, step 16, does not require RCS pressure meet certain values to terminate SI.

C. INCORRECT. SGTR Without Pressurizer Pressure Control procedure, step 7, does not require RCS pressure meet certain values to terminate SI.

D. INCORRECT. Response to Imminent Pressurized Thermal Shock Condition procedure, step 6, does not require RCS pressure meet certain values to terminate SI.

Technical Reference(s): 0POP05-EO-EO20, Faulted Steam Generator Isolation, Rev 11 0POP05-EO-EC11, Loss of Emergency Coolant Recirculation, Rev 18 0POP05-EO-EC33, SGTR Without Pressurizer Pressure Control, Rev 17 0POP05-EO-FRP1, Response to Imminent Pressurized Thermal Shock Condition, Rev 12 Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: New Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Question Level of Difficulty: 3 10 CFR Part 55 Content: 55.41.8, 41.10 55.45.3 Comments:

Examination Outline Cross-

Reference:

Level RO W/E 13 Steam Generator Over-pressure - Tier # 1 Knowledge of the interrelations between Group # 2 the (Steam Generator Overpressure) and K/A # W/E 13, EK2.2 the following:

Importance Rating 3.0 EK2.2 Facilitys heat removal systems, including primary coolant, emergency coolant, the decay heat removal systems, and relations between the proper operation of these systems to the operation of the facility.

Question 27 Given the following:

• Unit 1 is operating at 100% power.

• RCS Tavg is 592 °F.

• RCS T is 60 °F.

If RCS Tavg is allowed to rise to 602 °F, which one of the following should occur?

A. The reactor will trip on high pressurizer pressure B. RCS Subcooling will lower and go below 35 °F.

C. The Steam Dumps will automatically open.

D. SG PORVs will automatically open.

Proposed Answer: D Explanation (Optional):

A. INCORRECT: If the temperature of the pressurizer increases 10 °F as does Thot, saturation pressure is approximately 2350 psig, which is below the high pressure trip setpoint of 2380 psig B. INCORRECT: RCS subcooling will lower, but it's already below 35 °F in the intial condtions. (Tsat at NOP is approximately 650 °F)

C. INCORRECT: The Steam Dumps would open if they received an arming signal, but there is no arming signal present (turbine trip or load reduction).

D. CORRECT: With a Tavg of 602 °F, this would create a Tcold of approx. 572 °F. Tcold determines SG pressure and the saturation pressure for 572 °F is 1230 psig which is above the SG PORV setpoints.

Technical Reference(s): LOT 204.01, Integrated Plant Operations, Rev. 2 Proposed references to be provided to applicants during examination: Saturated steam temperature table Learning Objective: (As available)

Question Source: Bank # 1913 Question History: Last NRC Exam 2010 Question Cognitive Level: Comprehension or Analysis Question Level of Difficulty: 3 10 CFR Part 55 Content: 55.41.7 55.45.7 Comments:

Examination Outline Cross-

Reference:

Level RO 003 Reactor Coolant Pump - Ability to Tier # 2 predict and/or monitor changes in Group # 1 parameters (to prevent exceeding design K/A # 003, A1.03 limits) associated with operating the RCPS Importance Rating 2.6 controls including:

A1.03 RCP motor stator winding temperatures Question 28 According to procedure 0POP04-RC-0002, Reactor Coolant Pump Off Normal, which of the following valid conditions would require tripping the reactor and stopping the affected reactor coolant pump:

A. Motor Upper or Lower Radial Bearing Temp Equal to 175°F B. Lower Seal Water Bearing Temp Equal to 220°F C. Motor Stator Winding Temp equal to 320°F D. Number 1 Seal DP equal to 250 PSID

Proposed Answer: C Explanation (Optional):

The following are the requirements for tripping the reactor and stopping a reactor coolant pump A. INCORRECT. The RCP must be tripped when the Motor Upper or Lower Radial Bearing Temp is greater than or equal to 195°F B. INCORRECT. The RCP must be tripped when the Lower Seal Water Bearing Temp is greater than or equal to 230°F C. CORRECT. The RCP must be tripped when the Motor Stator Winding Temp is greater than or equal to 310°F D. INCORRECT. The RCP must be tripped when the Number 1 Seal DP is less than 220 PSID Technical Reference(s): 0POP04-RC-0002, Reactor Coolant Pump Off Normal, Rev 34 Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: New Question History: Last NRC Exam N/A Question Level of Difficulty: 2 Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 55.41.5 55.45.5 Comments:

Examination Outline Cross-

Reference:

Level RO 004 Chemical and Volume Control Tier # 2 Knowledge of the effect of a loss or Group # 1 malfunction on the following CVCS K/A # 004 K 6.17 components:

Importance Rating 4.4 K6.17 Flow paths for emergency boration Question 29 Conditions have occurred while responding to a Reactor Trip which requires Emergency Boration be initiated. While attempting to open MOV 218, ALT BORATION ISOL, the valve indicated intermediate and then all indication was lost. Per 0POP04-CV-0003, Emergency Boration for the given conditions what is the Emergency Boration flow path that will be established?

A. GOTO to Addendum 1 of 0POP04-CV-003 Emergency Boration from RWST and align the system by ensuring OCIV-MOV-025 is Open, Charging is aligned to inject via normal or alternate path, Charging Pump Running and RWST Isolation MOV-112C is open B. Preferred Emergency Boration Flowpath by ensuring OCIV-MOV-025 is Open, Charging is aligned to inject via normal or alternate path, Charging Pump Running and starting a Boric Acid Transfer Pump.

C. GOTO to Normal Boration through FCV 0110 A, per 0POP04-CV-003 Emergency Boration, Utilize Form 1 of 0POP02-CV-0001, Makeup to the Reactor Coolant System to continuality add Boron to the RCS.

D. Emergency Boration through "1(2)-CV-0226 BORIC ACID TANK 1A CHARGING PUMP SUCTION BORATION VALVE and 1(2)-CV-0333 BORIC ACID TANK 1A CHARGING PUMP SUCTION ISOLATION VALVE "

Proposed Answer: D Explanation (Optional):

A. Incorrect, per 0POP04-CV-0003, Emergency Boration if the preferred method of emergency boration is unavailable (Boric Acid Tank un available or not operable it directs you to Addendum 1, Emergency Boration which would direct emergency boration via the RWST From the RWST B. Incorrect, while MOV 218 may be partially open there is no way for the operator to know if the valve is open or closed based on the information that was given to him in the stem.

C. Incorrect, 0POP04-CV-003 does not direct the operator to transfer to Normal Boration via FCV 0110 A, this alignment adds water to the batch controller and then to the VCT, under emergency boration conditions the VCT is isolated per procedure. This is plasuable as 0POP04-CV-003 does direct the operator to initiate normal boration while manual actions are taken to manually open CV-0333 and CV-0226 then secure the normal boration path.

D. Correct, With the Boric Acid tanks available 0POP04-CV-003 will direct the operator to align normal boration flowpath while manual actions are taken to align suction to the CCP directly from the Boric Acid tanks.

Technical Reference(s): 0POP04-CV-0003 REV 14, Emergency Boration, 0POP02-CV-0001 REV 45, Makeup to the Reactor Coolant System (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: New X Question History: Last NRC Exam N/A Question Cognitive Level: Comprehension or Analysis Difficulty 3 10 CFR Part 55 Content: 55.41.8 Comments:

The procedure first attempts to initiate emergency boration by pumping from the Boric Acid Tanks using a Boric Acid pump through MOV-0218 to the suction of the charging pumps. If the Boric Acid Transfer Pumps are not available, then the Gravity feed method is used. If the Boric Acid Storage Tanks are not available, the RWST may be used as the source of boric acid.

0POP04-CV-0003, Emergency Boration

Steps 1-4 are met Step 5 cannot be completed so perform the RNO Step 5 RNO

• Align Normal boron injection flowpath

• Align the CCP suction directly to the Boric Acid tank

• Secure normal injection flow path

Examination Outline Cross-

Reference:

Level RO 005 Residual Heat Removal - Ability to (a) Tier # 2 predict the impacts of the following Group # 1 malfunctions or operations on the RHRS, K/A # 005 A2.01 and (b) based on those predictions, use Importance Rating 2.7 procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

A2.01 Failure modes for pressure, flow, pump motor amps, motor temperature, and tank level instrumentation Question 30 Given the following plant conditions:

• The plant is in Mode 4 cooling down to Mode 5

• All six RHR Pump suction MOVs have been energized

• The crew is preparing to place RHR in service

• Sensing line blockage has caused RCS wide range pressure transmitter PT-405 to stabilize at 700 psig

• RCS pressure has been verified to be 330 psig Which of the following correctly describes the effect of the sensing line blockage on the listed RHR suction MOVs?

A. Both RH-MOV-0060A and RH-MOV-0060C will open normally AND Both RH-MOV-0061A and RH-MOV-0061C will have to be open from the ASP.

B. Both RH-MOV-0060A and RH-MOV-0060C will have to be open from the ASP AND Both RH-MOV-0061A and RH-MOV-0061C will open normally.

C. Both RH-MOV-0061A and RH-MOV-0060C will open normally AND Both RH-MOV-0060A and RH-MOV-0061C will have to be open from the ASP.

D. Both RH-MOV-0061A and RH-MOV-0060C will have to be open from the ASP AND Both RH-MOV-0060A and RH-MOV-0061C will open normally.

Proposed Answer: C Explanation (Optional):

A. INCORRECT: RH-MOV-0060A will NOT open normally and RH-MOV-0061A will open normally.

B. INCORRECT: RH-MOV-0060C will open normally and RH-MOV-0061C will NOT open normally.

C. CORRECT: PT-405 signal transmits to RH-MOV-0060A and RH-MOV-0061C thus these valves will NOT open normally (they will need to be opened from the ASP), but RH-MOV-0061A and RH-MOV-0060C will open normally.

D. INCORRECT: RH-MOV-0061A will open normally and RH-MOV-0060A will NOT open normally.

Technical Reference(s): 0POP02-RH-0001, Residual Heat Removal System Operation, Rev 64; Lesson Plan Lot 201.09, Residual Heat Removal System, Rev 11 Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: Modified Bank # 1739 Question History: Last NRC Exam 2009 Question Cognitive Level: Comprehension or Analysis Question Level of Difficulty: 3 10 CFR Part 55 Content: 55.41.5 55.43.5 55.45.3/13 Comments:

Parent question below

Examination Outline Cross-

Reference:

Level RO 006 Emergency Core Cooling - Tier # 2 Knowledge of the operational implications Group # 1 of the following concepts as they apply to K/A # 006, K5.11 ECCS:

Importance Rating 2.5 K5.11 Basic heat transfer equation Question 31 Given the following:

• A Large Break Loss of Coolant Accident has occurred.

• The ECCS is operating in the Cold Leg Recirculation mode.

Based on standard heat transfer relationships, which of the following describes the INITIAL effects of INCREASED CCW temperature to the RHR Heat Exchanger during these plant conditions?

T across the RHR Heat Exchanger tubes will A. RISE resulting in LOWER heat transfer rate in the RHR Heat Exchanger.

B. LOWER resulting in LOWER heat transfer rate in the RHR Heat Exchanger.

C. RISE resulting in HIGHER heat transfer rate in the RHR Heat Exchanger.

D. LOWER resulting in HIGHER heat transfer rate in the RHR Heat Exchanger.

Proposed Answer: B Explanation (Optional):

A. INCORRECT. Higher tube inlet temperature results in T across the HX going down.

B. CORRECT. Based on Q=UA T, if T goes down, heat transfer rate also go down.

C. INCORRECT. Higher tube inlet temperature results in T across the HX going down.

D. INCORRECT. Based on Q=UA T, if T goes down, heat transfer rate must also go down.

Technical Reference(s): LOT 204.01, Integrated Plant Operations Part 2, Rev 2 Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: Bank # 1871 Question History: Last NRC Exam 2009 Question Cognitive Level: Comprehension or Analysis Question Level of Difficulty: 3 10 CFR Part 55 Content: 55.41.5 55.45.7 Comments:

Examination Outline Cross-

Reference:

Level RO 007 Pressurizer Relief/Quench Tank Tier # 2 Ability to monitor automatic operation of Group # 1 the PRTS, including: K/A # 007 A3.01 A3.01 Components which discharge to the Importance Rating 2.7 PRT.

Question 32 The following Unit 2 conditions exist:

• A faulted steam generator (outside containment) resulted in a reactor trip and safety injection

• Operators have isolated the faulted steam generator

• RCS pressure is currently 1800 psig and INCREASING

• PRT PRESS HI alarm illuminates

• PRT level and pressure are INCREASING slowly Which ONE of the following is the cause of the PRT level and pressure increase?

A. Reactor Coolant Pump seal leakoff flow is returning to the PRT.

B. A Pressurizer Power Operated Relief Valve is open.

C. Normal letdown flow is diverted to the PRT.

D. Loss of instrument air to containment fails open the Reactor Make Up Water valve to the PRT, SPRAY ISOL FV-3650.

Proposed Answer: A Explanation (Optional):

A. Correct B. Incorrect at current RCS pressure the PORV is not open C. Incorrect Letdown does not divert to the PRT D. Incorrect Technical Reference(s): LOT201.04 Rev. 8 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: none Learning Objective: (As available)

Question Source: Bank # NRC 160 Question History: Last NRC Exam 2003 Question Cognitive Level: Comprehension or Analysis Difficulty 3 10 CFR Part 55 Content: 55.41.7 55.45.3 Comments:

From Lesson Plan LOT201.04 Rev. 8 1.1.3 PRT receives relief from other systems in addition to discharge from PORV's and safeties:

A. RHR pump discharge - A, B, and C B. HHSI discharge (3)

C. LHSI discharge (3) (RHR heat exchanger bypass line)

D. CVCS seal return E. CVCS letdown F. Also accommodates reactor vessel head vent

Examination Outline Cross-

Reference:

Level RO 008 Component Cooling Water - Ability to Tier # 2 manually operate and/or monitor in the Group # 2 control room: K/A # 008, A4.05 Importance Rating 2.7 A4.05 Normal CCW-header total flow rate and the flow rates to the individual components cooled by the CCWS Question 33 With the plant operating at 100%, the Component Cooling Water Surge Tank Level is indicating 63.5%. Which of the following components would see a reduction in Component Cooling Water flow?

A. Spent Fuel Pool Heat Exchangers B. Reactor Coolant Pump Motor Air and Oil Coolers C. Letdown Heat Exchanger D. Centrifugal Charging Pump Oil Coolers

Proposed Answer: C Explanation (Optional):

Applicant must know what isolates on a CCW Surge Tank Lo-Lo Level at 64.6%.

A. INCORRECT. CCW to the Spent Fuel Pool Heat Exchangers does not isolate on a CCW Surge Tank Lo-Lo Level B. INCORRECT. CCW to the Reactor Coolant Pump Motor Air and Oil Coolers does not isolate on a CCW Surge Tank Lo-Lo Level C. CORRECT. CCW isolates to the Letdown Heat Exchanger on a CCW Surge Tank Lo-Lo Level.

D. INCORRECT. CCW to the Centrifugal Charging Pump Oil Coolers does not isolate on a CCW Surge Tank Lo-Lo Level Technical Reference(s): Lesson Plan Lot 201.12, Component Cooling Water System Rev 13; 0POP04-CC-0001, Component Cooling Water System Leak Rev 15 Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: New Question History: Last NRC Exam N/A Question Cognitive Level: Comprehension or Analysis Question Level of Difficulty: 3 10 CFR Part 55 Content: 55.41.7 55.45.5 Comments:

Examination Outline Cross-

Reference:

Level RO SRO Knowledge of bus power supplies to the Tier # 2 following: Group # 1 K2.01 PZR heaters K/A # 010/K2.01 Importance Rating 3.0 Question 34 Which of the following Load Centers provide power for the Pressurizer Backup Heater Group B?

A. E1A1 B. E1B1 C. E1B2 D. E1C1

Proposed Answer: D Explanation (Optional):

A. INCORRECT. E1A1 supplies power for the A Group Pressurizer Backup Heaters.

B. INCORRECT. E1B1 does not supply power to any Pressurizer Backup Heaters.

C. INCORRECT. E1B2 does not supply power to any Pressurizer Backup Heaters.

D. CORRECT. E1C1 supplies power for the B Group Pressurizer Backup Heaters.

Technical Reference(s): LOT201.36, 4.16KV, 480V ESF (Class 1E) AC Power Distribution, Rev 36 Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: New 2372 Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Question Level of Difficulty: 2 10 CFR Part 55 Content: 55.41.7 Comments:

Examination Outline Cross-

Reference:

Level RO 012 Reactor Protection - Knowledge of Tier # 2 RPS design feature(s) and/or interlock(s) Group # 1 which provide for the following: K/A # 012, K4.06 Importance Rating 3.2 K4.06 Automatic or manual enable/disable of RPS trips Question 35 Which of the following trip signals is blocked with reactor power at 15%?

A. Pressurizer low pressure B. Pressurizer high level C. Low flow in a single loop D. Reactor coolant pump undervoltage

Proposed Answer: C Explanation (Optional):

A. INCORRECT. The Pressurizer low pressure is blocked at less than 10% power.

B. INCORRECT. The Pressurizer high level is blocked at less than 10% power.

C. CORRECT. Low flow in a single loop is blocked at less than 40% power.

D. INCORRECT. Reactor coolant pump undervoltage is blocked at less than 10% power.

Technical Reference(s): Lesson Plan LOT 201.20, Rev 17 Proposed references to be provided to applicants during examination: None Question Source: New Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Question Level of Difficulty: 2 10 CFR Part 55 Content: 55.41.7 Comments:

Examination Outline Cross-

Reference:

Level RO 013 Engineered Safety Features Actuation Tier # 2

- Ability to monitor automatic operation of Group # 1 the ESFAS including: K/A # 013/A3.01 Importance Rating 3.7 A3.01 Input channels and logic Question 36 Ten minutes after a LOCA, containment pressure indicates the following:

• PT-934 = 9.5 psig

• PT-935 = 9.6 psig

• PT-936 = 9.3 psig

• PT-937 = 9.4 psig Which of the following describes the response of the Containment Spray System 10 seconds after containment pressure reaches the listed values, assuming bistables actuate at their exact setpoint and no ESF systems have been reset?

Containment Spray has automatically actuated; A. But only the pumps have started.

B. The pumps have started and their discharge valves have opened.

C. But only the pump discharge valves have opened.

D. But the pumps have not started and the discharge valves have not opened.

Proposed Answer: B Explanation (Optional):

A. INCORRECT: The valves directly open from the actuation signal B. CORRECT: Containment Spray actuates on a 2/4 logic at 9.5 psig. The valves directly open from the actuation signal. The pumps will start providing there is still a Sequencer signal present which locks in a 40 seconds after the sequencer started. The sequencer started on the initial Safety Injection. (RCB pressure at 3.0 psig.

C. INCORRECT: The pumps will also start because the sequencer has not been reset.

D. INCORRECT: The valves open directly from the actuation signal and the pumps will start from the sequencer.

Technical Reference(s): LOT201.11, Containment Spray System, Rev 14 LOT201.20, Solid State Protection System, Rev 17 LOT202.22, Engineering Safety Features, Rev 15 Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: Bank # 1359 Question History: Last NRC Exam 2009 Question Cognitive Level: Comprehension or Analysis Question Level of Difficulty: 3 10 CFR Part 55 Content: 55.41.7 55.45.5 Comments:

Examination Outline Cross-

Reference:

Level RO 022 Knowledge of the effect that a loss or Tier # 2 malfunction of the CSS Will have on the Group # 1 following: K/A # 022/K3.02 Importance Rating 3.0 K3.02 Containment instrumentation readings Question 37 Given the following containment history:

Time Containment Containment Containment Pressure Radiation Integrated dose 0815 2 psig 9.0 E4 R/hr 9.0 E3 Rad 0830 4 psig 1.0 E5 R/hr 2.7 E4 Rad 0845 6 psig 9.6 E6 R/hr 2.7 E5 Rad 0900 8 psig 2.0 E7 R/hr 7.7 E6 Rad Which of the following describes the EARLIEST time at which adverse containment should have been declared?

A. 0815 B. 0830 C. 0845 D. 0900

Proposed Answer: B Explanation (Optional):

A. INCORRECT. Adverse containment is either containment pressure greater than or equal to 5 psig, containment radiation level greater than or equal to 105 R/h, or containment integrated dose greater than or equal to 106 Rad.

B. CORRECT. Adverse containment is either containment pressure greater than or equal to 5 psig, containment radiation level greater than or equal to 105 R/h, or containment integrated dose greater than or equal to 106 Rad. Since containment radiation level equals 105 R/h, this is adverse containment.

C. INCORRECT. Adverse containment is either containment pressure greater than or equal to 5 psig, containment radiation level greater than or equal to 105 R/h, or containment integrated dose greater than or equal to 106 Rad. Since containment pressure is greater than 5 psig, and containment radiation level is greater than 105 R/h, this is adverse containment, however, adverse containment conditions had already been met.

D. INCORRECT. Adverse containment is either containment pressure greater than or equal to 5 psig, containment radiation level greater than or equal to 105 R/h, or containment integrated dose greater than or equal to 106 Rad. Since containment pressure is greater than 5 psig, containment radiation level is greater than 105 R/h, and containment integrated dose is greater than 106 Rad, this is adverse containment, however, adverse containment conditions had already been met.

Technical Reference(s): 0POP01-ZA-0018, Emergency Operating Procedure User's Guide, Rev. 21 Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: New Question History: Last NRC Exam N/A Question Cognitive Level: Comprehension or Analysis Question Level of Difficulty: 3 10 CFR Part 55 Content: 55.41.7 55.45.6 Comments:

Examination Outline Cross-

Reference:

Level RO 026 Containment Spray Tier # 2 Knowledge of CSS design feature(s) Group # 1 and/or interlock(s) which provide for the K/A # 026 K4.08 following:

Importance Rating 4.1 K4.08 Automatic swapover to containment sump suction for recirculation phase after Question 38 Unit 2 experienced a LOCA and the crew has just transitioned to 0POP05-EO-ES13, Transfer To Cold Leg Recirculation. The Reactor Operator notes and reports that Train A DID NOT automatically swap suction to the containment sump.

Which of the following conditions would cause the failure? (Consider each condition separately)

A. RWST Lo-Lo Alarm - ON B. STATUS LAMPBOX 5M2-3: AUTO RECIRC NOT RESET TRAIN A - OFF C. Train A, STATUS MONITORING PANEL 1M25: BYP/INOP RWST OUTL MOV-0001A -

ON D. Train A, HHSI Pump 1A MIN FLOW ISOL MOV-0011A - OPEN; MOV-0012A -

CLOSED

Proposed Answer: B Explanation (Optional):

A. Incorrect; each train receives a single level input and with an SI present should initiate and Auto Recirculation for its train. As presented RWST Train A level (LI-931) should have caused auto swap over (75,000 gallons); Trains B & C did by inference and LI-932 indicates higher than LI-931 B. Correct; this lampbox should come on with SI present and remain on until SI is reset and the SI AUTO RECIRC TRAIN A RESET pushbutton is depressed. There are no procedural actions in EO00 or EO10 to have reset Auto Recirculation. Indicates malfunction of SSPS Train A actuation output which will prevent auto swap over on lo-lo RWST level.

C. Incorrect; MOV-0001A Bypass/Inop (due to loss of control power or not fully open) does not prevent auto swap over operation from occurring.

D. Incorrect; Each SI pump has two recirc valves in series. The interlocking scheme for Auto Recirculation to occur requires 1/2 of the valves to go closed for each pump in a train to close. Therefore this condition would not prevent the swap from occurring.

Technical Reference(s): 9-Z-42114 Rev 14, 9-Z-42001 Rev 7, 9-Z-42004 Rev 5, 9-Z41813 Rev 8, 9-Z-42002 Rev7 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: LOT 201.10, 29419 Question Source: Bank # NRC 123 Question History: Last NRC Exam From STP NRC Bank not used on previous NRC exam Question Cognitive Level: Comprehension or Analysis Difficulty 2 10 CFR Part 55 Content: 55.41.9 Comments:

Parent NRC 123 Unit 2 experienced a LOCA and the crew has just transitioned to 0POP05-EO-ES13, Transfer To Cold Leg Recirculation. The Reactor Operator notes and reports that Train A DID NOT automatically swap suction to the containment sump.

Which ONE of the following indications would justify the failure? (consider each condition separately)

A. STATUS MONITORING PANEL 1M25: BYP/INOP RWST OUTL MOV-0001A - ON B. STATUS LAMPBOX 5M2-3: AUTO RECIRC NOT RESET TRAIN A - OFF C. RWST Lo-Lo Alarm - ON D. HHSI Pump 1A MIN FLOW ISOL MOV-0011A - OPEN; MOV-0012A - CLOSED Correct Answer B 1.1.1 RWST LO-LO/Empty A. LO-LO level

1. Initiates auto-recirc (Switchover)

2. Logic is train specific

3. One channel of RWST LO-LO level coincident with its train of S.I. Actuation will initiate auto switchover on that train of ECCS

4. Need all three levels LO-LO and all 3 Train SI signals to get all 3 trains to switch over automatically 1.2 ESF Status Monitoring 1.2.1 BYP INOP lights when any upper sugar cube lights.

1.2.2 SI lights if have a SI.

1.2.3 F/ACT lights if any center sugar cube lit (signifies that component NOT in its SI position after SI actuation).

1.2.4 AUTO RECIRC lights when RWST level reaches the setpoint for switching the pumps suction to the recirc sump.

1.2.5 F/ACT lights if any bottom sugar cube lit.

Examination Outline Cross-

Reference:

Level RO 039 Main and Reheat Steam System - Tier # 2 Knowledge of the operational implications Group # 1 of the following concepts as they apply to K/A # 039, K5.05 the MRSS:

Importance Rating 2.7 K5.05 Bases for RCS cooldown limits Question 39 Given the following:

• Unit 1 is in Mode 3

• Steam Dumps are in Steam Pressure Mode controlling at 1185 psig.

• All RCPs are running

• All Steam Dump Valves fail open causing the RCS to cool at >100 ºF/hr.

Which of the following correctly describes the MINIMUM operator action/s that would ensure all Steam Dumps are closed and the reason cooldown limits are established?

A. Place EITHER Steam Dump Interlock Selector Switch to BYPASS INTERLOCK.

Excessive cooldown can result in non-ductile failure of the Reactor Vessel.

B. Place BOTH Steam Dump Interlock Selector Switches to BYPASS INTERLOCK.

Excessive cooldown can result in ductile failure of the Reactor Vessel.

C. Place EITHER Steam Dump Interlock Selector Switch to OFF/RESET. Excessive cooldown can result in non-ductile failure of the Reactor Vessel.

D. Place BOTH Steam Dump Interlock Selector Switches to OFF/RESET. Excessive cooldown can result in ductile failure of the Reactor Vessel.

Proposed Answer: C Explanation (Optional):

A. INCORRECT. The 'Bypass Interlock' position of the Steam Dump Selector Switches will not close the steam dumps.

B. INCORRECT. The 'Bypass Interlock' position of the Steam Dump Selector Switches will not close the steam dumps. Also, cooldown limits are based on non-ductile failure (brittle) failure, not ductile failure.

C. CORRECT. The Steam Dump Interlock Selector Switches are redundant control devices such that either one being positioned to 'Off/Reset' will remove the control signal from all steam dumps, closing them. Cooldown limits are based on non-ductile failure (brittle) failure.

D. INCORRECT. Either Selector Sw. is sufficient to close the Steam Dumps, both are not needed. Also, cooldown limits are based on non-ductile failure (brittle) failure, not ductile failure.

Technical Reference(s): LOT 202.09, Steam Dump System, Rev 13; LOT102.61 handout page 25 Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: Bank # 1675 Question History: Last NRC Exam 2007 Question Cognitive Level: Comprehension or Analysis Question Level of Difficulty: 3 10 CFR Part 55 Content: 55.41.5 55.45.7 Comments:

Examination Outline Cross-

Reference:

Level RO 059 Ability to predict and/or monitor Tier # 2 changes in parameters (to prevent Group # 1 exceeding design limits) associated with K/A # 059, A1.03 operating the MFW controls including:

Importance Rating 2.7 A1.03 Power level restrictions for operation of MFW pumps and valves Question 40 Given the following:

• Unit 1 is operating at 100% power.

• SGFPT #13 is to be removed from service for pump maintenance.

• The Shift Manager has decided to run the Startup Feedpump (SUFP) during this time to remain at 100% power.

In accordance with 0POP02-FW-0002, S.G.F.P. Turbine, what additional action should be taken to account for SGFPT # 13 being removed from service?

A. A third FW Booster Pump must be started to reduce the load on the remaining two SGFPTs.

B. A third FW Booster Pump must be started to ensure SUFP flow matches the flow from the secured SGFPT.

C. The SGFP Master Speed Controller must be placed in Manual to help keep the remaining two SGFPTs below 5400 rpm.

D. The SGFP Master Speed Controller must be placed in Manual to match each of the remaining SGFPT flows with the SUFP flow.

Proposed Answer: A Explanation (Optional):

A. CORRECT. When using the S/U SGFP to replace a SGFP at 100% power, then a third FW Booster pump will be required to reduce the load on the remaining two SGFPS.

B. INCORRECT. When using the S/U SGFP to replace a SGFP at 100% power, then a third FW Booster pump will be required to reduce the load on the remaining two SGFPS.

C. INCORRECT. If placing Main Feedwater flow on the S/U SGFP from a running SGFPT then it MAY be necessary to take manual control of the SGFP MASTER SPEED CONTROLLER OR Main or Low Power Feedwater Regulating Valves to maintain stable SG Water Levels; at the discretion of the Unit Supervisor or Shift Manager.

D. INCORRECT. If placing Main Feedwater flow on the S/U SGFP from a running SGFPT then it MAY be necessary to take manual control of the SGFP MASTER SPEED CONTROLLER OR Main or Low Power Feedwater Regulating Valves to maintain stable SG Water Levels; at the discretion of the Unit Supervisor or Shift Manager.

Technical Reference(s): 0POP02-FW-0002, S.G.F.P. Turbine, Rev. 66 Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: Bank # 1668 Question History: Last NRC Exam Never Question Cognitive Level: Comprehension or Analysis Question Level of Difficulty: 3 10 CFR Part 55 Content: 55.41.5 55.45.5 Comments:

Examination Outline Cross-

Reference:

Level RO 061 Auxiliary/Emergency Feedwater Tier # 2 Ability to (a) predict the impacts of the Group # 1 following malfunctions or operations on K/A # 061 A2.07 the AFW; and (b) based on those Importance Rating 3.4 predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

A2.07 Air or MOV failure.

Question 41 Unit 1 has tripped from 100% power due to a complete loss of IA.

All systems have responded as designed The Turbine Driven Auxiliary Pump (AFW PUMP 14) initially started and then tripped for unknown reason.

Attempts to reset AFW Pump 14 have been unsuccessful.

The other AFW pumps are operating as designed.

In order to reestablish AFW flow to the Steam Generator 1D, the control room operator must A. Use control switches located on CP006 to open Cross connect valve for the Turbine Driven Auxiliary Pump (AFW Pump 14) and one additional Cross Connect Valve for an additional AFW Pump.

B. Use control switches located on CP006 to open Cross connect valve for the Turbine Driven Auxiliary Pump (AFW Pump 14)

C. Dispatch an operator to manually open Cross Connect for the Turbine Driven Auxiliary Pump (AFW Pump 14)

D. Dispatch an operator to manually open Cross Connect for the Turbine Driven Auxiliary Pump (AFW Pump 14) and one additional Cross Connect Valve from an additional AFW Pump.

Proposed Answer: D Explanation (Optional):

A. Incorrect, on a loss of Instrument air the cross connect valves fail closed and there is no motive air to reposition remotely B. Incorrect, See A but plausible because two valves are required to be opened C. Incorrect, 2 valves are required to cross connect the AFW system D. Correct Technical Reference(s): LOT202.28 Rev. 10 (AFW Lesson plan), 0POP05-EO-FRH5 Rev 7, RESPONSE TO STEAM GENERATOR LOW LEVEL (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Question Source:

New 2410 X Question History: Last NRC Exam N/A Question Cognitive Level: Comprehension or Analysis Difficulty 3 10 CFR Part 55 Content: 55.41.5 55.43.5 55.45.3, 55.45.13 Comments:

From lesson plan LOT202.28 Rev. 10 (AFW Lesson plan)

Manual operation A. Handwheel provided on valve operator to allow manual operation if required due to failure (i.e., loss of power or air pressure)

B. Manual control allows operator to open valve and establish flow from any pump to any SG C. Handwheel also allows valve to be gagged shut D. Opening valve manually

1. Rotating handwheel in open direction moves handwheel screw upward

2. Handwheel screw contacts jam nuts (Trains A, B, and C) or stem lock nut (Train D)

3. Forces valve open against spring pressure RNO from 0POP05-EO-FRH5 Rev 7, RESPONSE TO STEAM GENERATOR LOW LEVEL Step 4 RNO

c. IF any AFW pump fails to start, THEN:

1) RESET all SG LO-LO level AFW actuations.

2) ENSURE applicable AFW regulating valve CLOSED.

3) OPEN applicable AFW cross-connects.

4) CONTROL AFW flow to LESS THAN 675 GPM per AFW pump.

Examination Outline Cross-

Reference:

Level RO 062 AC Electrical Distribution Tier # 2 Group # 1 2.4.3 Ability to identify post-accident K/A # 062, 2.4.3 instrumentation Importance Rating 3.7 Question 42 A loss of which Class 1E vital 120 VAC power distribution channel will result in a loss of train C accident monitoring described in TRM 3.3.3.6?

A. I B. II C. III D. IV

Proposed Answer: D Explanation:

Per 0POP02-AE-0004, 120 VAC ESF Vital Distribution Power Supplies, Addendum 3, Loss of Distribution Panel DP 002 Failures, a loss of DP 002 will de-energize train C containment hydrogen monitoring. Per powerpoint for LOT 201.38, Channel IV of Class 1E Vital 120 VAC provides power to DP 002. Also LOT 201.27 states the monitor power supply as hydrogen monitor CM-AIT-4105-DP002. Per TRM 3.3.3.6, the only listed accident instrument is containment hydrogen monitoring. The other options are plausible if the applicant does not remember which channel supplies the C Containment Hydrogen Monitor. Because it requires the applicant to determine what is required accident monitoring in order to answer the question, it meets the K/A.

Technical Reference(s): 0POP02-AE-0004 Addendum 3 Rev. 58, LOT 201.38 Rev.

13, LOT 201.27 Rev. 11, TRM 3.3.3.6 Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: New Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Question Level of Difficulty: 4 10 CFR Part 55 Content: 55.41.7

Examination Outline Cross-

Reference:

Level RO 063 - DC Electrical Distribution Tier # 2 Group # 1 2.2.39 Knowledge of less than or equal to K/A # 063, 2.2.39 one hour Technical Specification action Importance Rating 3.9 statements for systems.

Question 43 The plant is in Mode 3.

Both battery chargers for Channel I are inoperable. Both battery chargers for Channel II are also inoperable.

Per Technical Specification 3.8.2.1, within one hour restore, at a minimum, _____(1)____ to operable status on ______(2)_________ .

A. One battery charger; EITHER Channel I OR Channel II B. One battery charger; BOTH Channel I AND Channel II C. Both battery chargers; EITHER Channel I OR Channel II D. Both battery chargers; BOTH Channel I AND Channel II

Proposed Answer: A Explanation (Optional):

Per TS 3.8.2.1.d, with more than 1 channel with no battery chargers operable, one battery charger must be restored to operable on at least three channels within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. Both Channels III and IV are not mentioned and are thus implied to be operable. Therefore, one battery charger on either Channel I or Channel II must be restored to operable within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> per the TS.

Because this asks for knowledge of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or less TS, it is considered to be RO level knowledge per the SRO guidance document. The other options are plausible if the applicant cannot remember how many chargers are required to be operable on how many channels to exit the one hour action statement.

Technical Reference(s): TS 3.8.2.1.d Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: New Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Question Level of Difficulty: 3 10 CFR Part 55 Content: 55.41.10 Comments:

Examination Outline Cross-

Reference:

Level RO 064 Emergency Diesel Generator - Tier # 2 Knowledge of the effect of a loss or Group # 1 malfunction of the following will have on K/A # 064 K6.08 the ED/G system:

Importance Rating 3.2 K6.08 Fuel oil storage tanks Question 44 With an ESF DG in a standby condition, which of the following would cause a DG TROUBLE alarm?

A. One DG Starting Air Receiver pressure 180 psig, the other DG Starting Air Receiver Pressure 200 psig B. DG Fuel Oil Storage Tank level 60,000 gallons C. DG Jacket Water temperature 125°F D. DG Lube Oil temperature 125°F

Proposed Answer: B Explanation (Optional):

A. INCORRECT. In order for the DG TROUBLE alarm to come in starting air receiver pressure on at least one of the starting air receivers must be below 175 psig.

B. CORRECT. The DG Trouble alarm will come in if DG fuel oil storage tank level is below 61,460 gallons or above 64,360 gallons.

C. INCORRECT. In order for the DG TROUBLE alarm to come in, DG Jacket Water temperature must be less than 100°F or above 190°F.

D. INCORRECT. In order for the DG TROUBLE alarm to come in, DG Lube Oil temperature must be less than 100°F or above 185°F.

Technical Reference(s): 0POP09-AN-03M3, Annunciator Lampbox 3M03 Response Instructions, Rev 32 Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: New 2374 Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Question Level of Difficulty: 3 10 CFR Part 55 Content: 55.41.7 55.45.7 Comments:

Examination Outline Cross-

Reference:

Level RO 073 Process Radiation Monitoring - Tier # 2 Knowledge of the physical connections Group # 2 and/or cause effect relationships between K/A # 073 K1.01 the PRM system and the following Importance Rating 3.6 systems:

K1.01 Those systems served by PRMs Question 45 Which of the following is a control function of only the Reactor Containment Building Ventilation System effluent radiation monitors, RT-8012 & 8013?

Sends a signal to A. The Gaseous Waste Processing System shutdown circuitry to close the intake and exhaust valves.

B. The Solid State Protection System for Containment Ventilation Isolation C. Initiate Control Room/Electrical Auxiliary Building emergency ventilation.

D. Initiate Fuel Handling Building exhaust filtration.

Proposed Answer: B Explanation (Optional):

From UFSAR Table 7.3-3, one of the inputs to a containment isolation signal is high radiation signal and is derived from 1 of the two Class 1E RCB Purge Isolation monitors. The other distractors are plausible as there are PRMs that perform those functions.

Technical Reference(s):

Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: Bank # 1327 Question History: Last NRC Exam 2005 Question Cognitive Level: Memory or Fundamental Knowledge Question Level of Difficulty: 2 10 CFR Part 55 Content: 55.41.7 Comments:

Examination Outline Cross-

Reference:

Level RO 076 Service Water - Ability to predict Tier # 2 and/or monitor changes in parameters (to Group # 1 prevent exceeding design limits) K/A # 076, A1.02 associated with operating the SWS Importance Rating 2.6 controls including:

A1.02 Reactor and turbine building closed cooling water temperatures Question 46 Unit 1 is at 15% power and commencing to raise turbine load at 10%/hr.

Which of the following describes how the temperature of the components cooled by the Closed Loop Auxiliary Cooling Water (CL-ACW) System will be controlled?

A. The Open Loop Auxiliary Cooling Water (OL-ACW) System TCV (Temperature Control Valve) on the outlet of the SW/FW Heat Exchanger will modulate open to maintain CL-ACW temperature.

B. The Closed Loop Auxiliary Cooling Water (CL-ACW) System TCV on the outlet of the SW/FW Heat Exchanger will modulate open to maintain CL-ACW temperature.

C. The individual component TCVs will modulate open to maintain component temperature.

D. The TGB Watch will manually throttle CL-ACW from the SW/FW Heat Exchanger to maintain CL-ACW temperature.

Proposed Answer: C Explanation (Optional):

A. INCORRECT: There is not a TCV on the Open Loop outlet of the Hx. There is full cooling water flow to the Hx at all times.

B. INCORRECT: Closed loop cooling flow is not modulated on the outlet of the Hx, there is full system flow through the Hx at all times.

C. CORRECT: Since system temperature is not maintained, each cooled component has a TCV on it's CL-ACW cooling outlet line.

D. INCORRECT: CL-ACW system temperature is not controlled, there is full flow cooling flow thru the Hx.

Technical Reference(s): Lesson Plan 202.24, Rev 7 Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: Bank # 1083 Question History: Last NRC Exam 2010 Question Cognitive Level: Memory or Fundamental Knowledge Question Level of Difficulty: 3 10 CFR Part 55 Content: 55.41.5 55.45.5 Comments:

Examination Outline Cross-

Reference:

Level RO 078 Instrument Air Tier # 2 K3 Knowledge of the effect that a loss Group # 1 or malfunction of the IAS will have on K/A # 078 K3.02 the following:

Importance Rating 3.4 K3.02 Systems having pneumatic valves and controls.

Question 47 Unit 2 is in Mode 5 maintaining 160 to 180 degrees in the RCS on the A RHR System The containment instrument air isolation valve malfunctions and goes shut.

What are the plant affects based on this failure?

A. The RHR temperature control valves and Bypass Valve Close and a plant heat up will occur.

B. The RHR Temperature Control Valves Close and the Bypass Valve Opens and a plant heat up will occur C. The RHR Temperature Control Valves and Bypass Valves Remain in the current position and no changes in RCS temperature occur.

D. The RHR Temperature Control Valves Open and the Bypass Valve close and a plant cool down occurs

Proposed Answer: D Explanation (Optional):

Technical Reference(s):RHR design basis document 5R169MB1021 Rev. 0 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: New X Question History: Last NRC Exam N/A Question Cognitive Level: Comprehension or Analysis Difficulty 3 10 CFR Part 55 Content: 55.41.7 55.45.6 Comments:

2.INTEGRATED OPERATION (LOT only)

Using the above information, it is possible to determine what the system response will be to different situations by analyzing given conditions and applying system automatic actions. For example, the following actions have taken place:

• The plant is in mode 5, maintaining 170ºF - 190º A RHR.

• The containment Instrument Air isolation valve goes shut.

• What are the plant effects with regard to RHR?

Without Instrument Air to the temperature control valves in RHR, the temperature control valves go full open and the bypass go full shut, initiating max cooldown.

Examination Outline Cross-

Reference:

Level RO 103 Containment - Knowledge of Tier # 2 containment system design feature(s) Group # 2 and/or interlock(s) which provide for the K/A # 103, K4.06 following:

Importance Rating 3.1 K4.06 Containment isolation system Question 48 A main steam line break inside containment has occurred causing the containment pressure to rise to 7.5 psig.

Which of the following valves should be open?

A. FV-4493 CCW TO RCPs CONT. ISOL. VALVE B. FV-3936 RWST TO SFPCCS ISOLATION VALVE C. FV-0011 LETDOWN ORIFICE HEADER ISOLATION VALVE D. FV-9776 RCB SUPPLEMNTARY PURGE CONT. ISOL. VALVE

Proposed Answer: A Explanation (Optional):

A. CORRECT. FV-4493 would close on a Containment Isolation Phase B signal, which actuates on a Containment Spray Signal on HI-3 Containment Pressure (). That signal would not be valid.

B. INCORRECT. FV-3936 closes on a Safety Injection Signal, which actuates on HI-1 containment pressure.

C. INCORRECT. FV-0011 closes on a Containment Isolation Phase A signal, which is valid on the Safety Injection Signal.

D. INCORRECT. FV-9776 closes on a Containment Ventilation signal, which is valid on the Safety Injection Signal.

Technical Reference(s): UFSAR, Chapter 7.3 Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: New Question History: Last NRC Exam N/A Question Cognitive Level: Comprehension or Analysis Question Level of Difficulty: 3 10 CFR Part 55 Content: 55.41.7 Comments:

Examination Outline Cross-

Reference:

Level RO 008 Component Cooling Water Tier # 2 Group # 1 2.4.11 Knowledge of abnormal condition K/A # 008, 2.4.11 procedures Importance Rating 4.0 Question 49 With Unit 1 at 100% power, a 0.5 gpm leak occurs in the A RCP Thermal Barrier Heat Exchanger.

Later the CCW SURGE TK LVL LO annunciator is received due to CCW leakage at the RCDT heat exchanger and CCW surge tank levels continue to lower.

At this point, which procedure should be entered and why?

A. 0POP04-RC-0002, Reactor Coolant Pump Off Normal, to shutdown the failed reactor coolant pump.

B. 0POP04-CC-0001, Component Cooling Water System Leak, to isolate/ensure isolation of CCW to the RCP Thermal Barrier Heat Exchanger.

C. 0POP04-CC-0001, Component Cooling Water System Leak, to isolate/ensure isolation of CCW to the RCDT Heat Exchanger.

D. 0POP05EOEO00, Reactor Trip or Safety Injection, to respond to a reactor trip due to manually tripping the reactor in response to loss of CCW surge tank level.

Proposed Answer: C Explanation (Optional):

A. INCORRECT. This procedure would only be entered if there is a Thermal Barrier Heat Exchanger High Temperature Alarm.

B. INCORRECT. While this procedure is entered, it does not direct isolating CCW to the RCP Thermal Barrier Heat Exchangers C. CORRECT. The entry conditions for this procedure is due to the CCW Surge Tank Low Level alarm, and the procedure will isolation CCW to the RCDT heat Exchanger.

D. INCORRCT. A reactor trip would only occur if CCW flow to the RCP Thermal Barrier Heat Exchanger was less than 30 gpm.

Technical Reference(s): 0POP04-CC-0001, Component Cooling Water System Leak, Rev 15 Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: Bank # 1503 Question History: Last NRC Exam Never Question Cognitive Level: Comprehension or Analysis Question Level of Difficulty: 3 10 CFR Part 55 Content: 55.41.10 55.43.5 55.45.13 Comments:

PARENT QUESTION:

With Unit 2 at rated thermal power, a 0.5 gpm leak develops in the RCP A thermal barrier heat exchanger.

Later, the CCW SURGE TK LVL LO annunciator is received due to CCW leakage at the RCDT heat exchanger, and CCW surge tank levels continues to lower.

At this point, the Unit Supervisor should enter procedure:

A. POP04-RC-0002, Reactor Coolant Pump Off Normal, to shutdown the failed RCP B. POP04-CC-0001, Loss of Component Cooling Water, to isolate/ensure isolation of CCW to the RCDT heat exchanger C. POP04-CC-0001, Loss of Component Cooling Water, to isolate/ensure isolation of CCW to the RCP A thermal barrier heat exchanger D. POP05-EO-EO00, Reactor Trip or Safety Injection, after tripping the reactor due to loss of CCW surge tank level

Examination Outline Cross-

Reference:

Level RO Tier # 2 026 Containment Spray Group # 1 Knowledge of CSS design feature(s) and/or interlock(s) which provide for the following: K/A # 026 K3.04 (CFR: 41.7) Importance Rating 4.1 K4.08 Automatic swapover to containment sump suction for recirculation phase after LOCA (RWST low-low level alarm)

Question 50 Unit 2 was operating at 100% power when a Unit Trip occurs due to a large break loss of coolant accident inside containment. Containment pressure has exceeded 15 psig. All of the ECCS equipment has operated as designed.

Subsequently:

• RWST level has lowered to 75,000 gallons (14%).

• Auto swap over initiated.

• ALL valves repositioned as expected with the exception of the Mini flow recirculation valves on Train A HHSI and LHSI pumps.

• Operators closed ALL the RWST suction valves per 0POP05-EO-ES13, Transfer to Cold Leg Recirculation.

What is the plant response to this configuration?

Containment water levels will A. continue to rise as the remaining water in the RWST drains into the sump due to the Train A Mini Flow Recirculation valve position.

B. lower due to the recirculation of the water back into the RCS and the RWST will remain at a constant level.

C. remain constant and the RWST level will lower as water is being removed from the tank due to the Train A HHSI and LHSI pumps still removing water via the Train A Mini Flow Recirculation valves.

D. lower slowly and the RWST level will rise due to Train A Mini Flow Recirculation valve position.

Proposed Answer: D Explanation (Optional):

A. Incorrect, the mini flow valves will cause some of the water in the sump to be diverted into the RWST B. Incorrect, the containment sump level will lower but not due to the water being diverted to the RCS C. Incorrect, RWST level will actually rise, no suction source is available to the ECCS pumps from the RWST D. Correct, the ECCS pumps will divert a portion of the flow from the sump to the RWST.

Technical Reference(s): 14800032QG, Rev C ( Step Basis for 0POP05-ES13 Rev 10, Transfer to Cold Leg Recirculation), 0POP05-ES13 Rev 10, Transfer to Cold Leg Recirculation, 0POP05-EO-EO10, Loss of Reactor or Secondary Coolant (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: New 2412 X Question History: Last NRC Exam N/A Question Cognitive Level: Comprehension or Analysis Difficulty 3 10 CFR Part 55 Content: 55.41.7 55.45.6 Comments:

Examination Outline Cross-

Reference:

Level RO 063 - DC Electrical Distribution System Tier # 2 Knowledge of the physical connections Group # 1 and/or cause/effect relationships between K/A # 063, K1.03 the DC electrical system and the following Importance Rating 2.9 systems:

K1.03 Battery charger and battery Question 51 Given the following:

• Maintenance is being performed on E1C11 battery

• The battery breaker is open to allow maintenance to jumper out a cell.

• An overcurrent lockout causes 4 KV ESF Bus E1C to de-energize.

Based on these conditions, Class 1E 120 VAC Distribution Panel DP-1204 will remain:

A. Energized through its associated Inverter/Rectifier.

B. Energized through its associated Voltage Regulating Transformer.

C. De-energized until the E1C11 battery breaker is closed locally.

D. De-energized until ESF Diesel Generator #13 output breaker automatically closes.

Proposed Answer: C Explanation (Optional):

A. INCORRECT. With the given conditions, both the AC and DC supplies to the inverter/rectifier are de-energized.

B. INCORRECT. With the given conditions, the supply to the voltage regulating transformer is de-energized.

C. CORRECT. Closing the battery breaker will energize the DP panel through its inverter.

D. INCORRECT. With the given conditions, the diesel output breaker will not close.

Technical Reference(s): LOT 201.38, Class 1E Vital 120 VAC, Rev 13 Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: Bank # 1677 Question History: Last NRC Exam 2007 Question Cognitive Level: Memory or Fundamental Knowledge Question Level of Difficulty: 3 10 CFR Part 55 Content: 55.41.2 through 55.41.9 55.45.7, 55.45.8 Comments:

PARENT QUESTION:

Given the following:

• The E1C11 battery breaker is open to allow maintenance to jumper out a cell.

• An overcurrent lockout causes 4 KV ESF Bus E1C to de-energize.

Based on these conditions, Class 1E 120 VAC Distribution Panel DP-1204 will remain:

A. energized through its associated Inverter/Rectifier.

B. energized through its associated Voltage Regulating Transformer.

C. de-energized until the E1C11 battery breaker is closed locally.

D. de-energized until ESF Diesel Generator #13 output breaker automatically closes.

Examination Outline Cross-

Reference:

Level RO 064 Emergency Diesel Generator - Ability Tier # 2 to (a) predict the impacts of the following Group # 1 malfunctions or operations on the ED/G K/A # 064 A2.08 system; and (b) based on those Importance Rating 2.7 predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

A2.08 Consequences of opening/closing breaker between buses (VARS, out-of-phase, voltage)

Question 52 Given the following conditions:

• Emergency Diesel Generator (EDG) 12 is being paralleled to Train B Bus.

• Emergency Diesel Generator output breaker is closed with EDG voltage slightly less than the Train B Bus.

Which of the following:

1) Identifies the impact on the Emergency Diesel if voltage remains less than the Train B Bus?

2) What action should be taken?

A. 1) EDG VAR meter will move in the negative (-) VAR (LEAD-IN) direction.

2) Place the Generator Voltage Adjust Switch in the RAISE position to ensure a slightly positive VAR load.

B. 1) EDG VAR meter will move in the positive (+) VAR (LAG-OUT) direction.

2) Place the Generator Voltage Adjust Switch in the LOWER position to ensure a slightly positive VAR load.

C. 1) EDG VAR meter will move in the negative (-) VAR (LEAD-IN) direction.

2) Place the Generator Voltage Adjust Switch in the LOWER position to ensure a slightly positive VAR load.

D. 1) EDG VAR meter will move in the positive (+) VAR (LAG-OUT) direction.

2) Place the Generator Voltage Adjust Switch in the RAISE position to ensure a slightly positive VAR load.

Proposed Answer: A Explanation (Optional):

A. CORRECT. With EDG voltage less than bus voltage when the breaker is closed, a negative VAR load will be absorbed into the Emergency Diesel Generator. The Generator Voltage Adjust Switch is placed in RAISE to increase generator terminal voltage and zero out the VAR load.

B. INCORRECT. Plausible because this would be the correct action if generator voltage were higher than Safeguards Bus voltage when the breaker was closed and it was desired to zero out the VAR load.

C. INCORRECT. Plausible because the VAR meter will move in the negative direction, however, placing the Generator Voltage Adjust Switch in lower will cause more VARs to be absorbed into the Generator.

D. INCORRECT. Plausible because the action is correct, however, positive VARs would only be created if EDG voltage were higher than Safeguards Bus voltage when the breaker was closed.

Technical Reference(s): 0POP02-DG-0002, Emergency Diesel Generator 12(22),

Rev 66, LOT201.39; Emergency Diesel Generator, Rev 15; 0PSP03-DG-0001, Standby Diesel 11(21) Operability Test, Rev 50 Proposed references to be provided to applicants during examination: None Question Source: New 2376 Question History: Last NRC Exam N/A Question Cognitive Level: Comprehension or Analysis Question Level of Difficulty: 3 10 CFR Part 55 Content: 55.41.5 55.43.5 55.45.3/13 Comments:

Examination Outline Cross-

Reference:

Level RO 076 Service Water - Ability to manually Tier # 2 operate and/or monitor in the control Group # 1 room: K/A # 076 A4.01 Importance Rating 2.9 A4.01 SWS Pumps Question 53 With the plant in Mode 1 and a Train of Essential Cooling Water (ECW) Pump switches in the following positions:

• Controlroom Handswitch - AUTO

• Transfer Switch - CONT RM

• ECW/CCW Train Selector Switch - STANDBY Which of the following is NOT an auto start signal for an ECW Pump?

A. Auto start of the same-train ESF DG.

B. SI actuation signal.

C. ECW pressure in the other two ECW Trains 25 psig.

D. CCW header pressure 75 psig.

Proposed Answer: A Explanation (Optional):

A. CORRECT. This is NOT an auto start signal for an ECW pump.

B. INCORRECT. An ESF Sequencer start signal will start the ECW pumps after a time delay (Modes 1, 2, and 3).

C & D. INCORRECT. With the control switches for a non-running Train in the "AUTO" and "CONT RM" positions and the associated ECW/CCW Train Selector Switch in the "STANDBY" position, the ECW/CCW pumps in that Train will automatically start and annunciator "CCW TRAIN AUTO START" will be actuated on the train's ESF Control Panel after a 15 second time delay if either of two conditions occur: ECW pressure in the other two ECW Trains goes below 30 PSIG, or CCW common header pressure goes below 76 PSIG.

Technical Reference(s): LOT201.13, Essential Cooling Water, Rev 7, slide 40 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: Modified 1329 Question History: Last NRC Exam 2005 Question Cognitive Level: Memory or Fundamental Knowledge Question Level of Difficulty: 2 10 CFR Part 55 Content: 55.41.7 55.45.5 to 55.45.8 Comments:

Original Question:

Given the following:

• Unit 1 is operating at 5% power

• Train A ECW pump handswitch is in AUTO

• CCW/ECW Train A MODE SEL SWITCH is in STANDBY Under which of the following conditions would the operator expect the Train A ECW Pump to automatically start?

1. Safety Injection (SI) actuation signal (MODE 1)

2. Normal start of ESF DG #11

3. CCW Common Header pressure is 82 psig

4. Train B and C ECW Pump discharge pressures are 28 psig A. 1 and 2 B. 1 and 4 C. 3 and 4 D. 2 and 3

Examination Outline Cross-

Reference:

Level RO SRO 078 Instrument Air - Ability to monitor Tier # 2 automatic operation of the IAS, including: Group # 1 K/A # 078 A3.01 A3.01 Air pressure Importance Rating 3.1 Question 54 The Unit 1 Instrument Air (IA) Compressors are in remote control from the Master Control (MC)

Computer.

The Instrument Air Compressors are aligned as follows:

• IA Compressor 11 is the first compressor

• IA Compressor 12 is the second compressor

• IA Compressor 13 is the third compressor Current system pressure is at approximately 120 psi with all three compressors IDLING.

Subsequently Service Air usage causes air pressure at the outlet of the air dryers to drop to 114 psi, then slowly rise back and stabilize at 122 psi.

Assuming the Service Air usage lasted less than 30 minutes and the IA Compressors LOAD and IDLE at their exact setpoints, which describes the condition of the IA Compressors after pressure stabilized at 122 psi?

IA Compressor 11 IA Compressor 12 IA Compressor 13 A. LOADED LOADED LOADED B. LOADED LOADED IDLE C. LOADED IDLE IDLE D. IDLE IDLE IDLE

Proposed Answer: B Explanation (Optional):

A. INCORRECT. The pressure does not decrease below 113 psi for the third compressor to load.

B. CORRECT. Pressure gets to 114 psi, so the third compressor does not load. The pressure does not increase above the idling setpoint for the second compressor, so it remains loaded.

C. INCORRECT. The pressure does not increase above for 125 psi for the second compressor to idle.

D. INCORRECT. The pressure does not increase above for 127 psi for the second compressor to idle.

Technical Reference(s): 0POP02-IA-0003, Instrument Air System Operation, Revision 29 Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: New 2378 Question History: Last NRC Exam N/A Question Cognitive Level: Comprehension or Analysis Question Level of Difficulty: 4 10 CFR Part 55 Content: 55.41.7 55.45.5 Comments:

Examination Outline Cross-

Reference:

Level RO SRO 005 Residual Heat Removal - Knowledge Tier # 2 of bus power supplies to the following: Group # 1 K/A # 005 K2.01 K2.01 RHR pumps Importance Rating 3.0 Question 55 Which of the following describes the power supplies for the RHR Pumps:

A. RHR Pump 1A - Load Center E1A1 RHR Pump 1B - Load Center E1B1 RHR Pump 1C - Load Center E1C1 B. RHR Pump 1A - Load Center E1A2 RHR Pump 1B - Load Center E1B2 RHR Pump 1C - Load Center E1C2 C. RHR Pump 1A - Load Center E1A1 RHR Pump 1B - Load Center E1B1 RHR Pump 1C - Load Center E1C2 D. RHR Pump 1A - Load Center E1A2 RHR Pump 1B - Load Center E1B2 RHR Pump 1C - Load Center E1C1

Proposed Answer: D Explanation (Optional):

A. INCORRECT. The normal power supply for RHR Pump 1A is E1A2, and the normal power supply for RHR Pump 1B is E1B2.

B. INCORRECT. The normal power supply for RHR Pump 1C is E1C1.

C. INCORRECT. The normal power supply for RHR Pump 1A is E1A2, and the normal power supply for RHR Pump 1B is E1B2, and the normal power supply for RHR Pump 1C is E1C1.

D. CORRECT. These are the correct pump power supplies.

Technical Reference(s): LOT 201.36, Class 1E 4.16/480 VAC Elec. Dist, Rev 8 Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: New 2379 Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Question Level of Difficulty: 3 10 CFR Part 55 Content: 55.41.7 Comments:

Examination Outline Cross-

Reference:

Level RO 011 Pressurizer Level Control - Tier # 2 Knowledge of bus power supplies to the Group # 2 following: K/A # 011/K2.01 Importance Rating 3.1 K2.01 Charging pumps Question 56 Which of the following describes the power supplies for the Unit 2 Centrifugal Charging Pumps (CCPs)?

CCP 2A CCP 2B A. Class 1E 4.16KV E2A Class 1E 4.16KV E2B B. Class 1E 4.16KV E2A Class 1E 4.16KV E2C C. Class 1E 4.16KV E2C Class 1E 4.16KV E2B D. Class 1E 4.16KV E2C Class 1E 4.16KV E2A

Proposed Answer: D Explanation (Optional):

A. INCORRECT. CCP 1A is powered from E1C (E2C) and CCP 1B is powered from E1A (E2A)

B. INCORRECT. CCP 1A is powered from E1C (E2C) and CCP 1B is powered from E1A (E2A)

C. INCORRECT. CCP 1B is powered from E1A (E2A)

D. CORRECT. CCP 1A is powered from E1C (E2C) and CCP 1B is powered from E1A (E2A)

Technical Reference(s): LOT201.06, Chemical and Volume Control System (CVCS), Rev 14 Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: New 2380 Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Question Level of Difficulty: 2 10 CFR Part 55 Content: 55.41.7 Comments:

Examination Outline Cross-

Reference:

Level RO 015 Nuclear Instrumentation - Knowledge Tier # 2 of NIS design feature(s) and/or interlock(s) Group # 2 provide for the following: K/A # 015/K4.07 Importance Rating 3.7 K4.07 Permissives Question 57 Which of the following is correct regarding the P-6 Bistable?

A. Both Intermediate Range NIs must go above 10-10 amps to automatically block Source Range NIs.

B. Both Intermediate Range NIs must go above 10-10 amps to allow a manual block of the Intermediate Range trip.

C. Both Intermediate Range NIs must go below 10-10 amps to automatically unblock Source Range Nis D. Both Intermediate Range NIs must go below 10-10 amps to allow a manual unblock of the Intermediate Range trip.

Proposed Answer: C Explanation (Optional):

A. INCORRECT. Credible because the Excore NIs do provide some automatic blocks but blocking the Source Range NIs is not one of them.

B. INCORRECT. Credible because the Intermediate Range trip is manually blocked but not until Power Range NIs go above 10% power.

C. CORRECT. Both Intermediate Range NIs must go below 10 to the minus tenth amps to automatically unblock the Source Range NIs.

D. INCORRCT. Credible because the Intermediate Range trip is unblocked but not until Power Range NIs go below 10% power.

Technical Reference(s): LOT201.16, Excore Nuclear Instrumentation, Rev. 13 Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: New 2381 Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Question Level of Difficulty: 2 10 CFR Part 55 Content: 55.41.7 Comments:

Examination Outline Cross-

Reference:

Level RO 029 Containment Purge - Ability to (a) Tier # 2 predict the impacts of the following Group # 2 malfunctions or operations on the K/A # 029, A2.01 Containment Purge System; and (b)

Importance Rating 2.9 based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

A2.01 Maintenance or other activity taking place inside containment.

Question 58 Given the following:

• Plant is at 100% power steady state operation.

• Preparations for performing a containment purge are in progress.

• Noble gas concentration inside the RCB is 5.3E-04 Ci/cc.

Which of the following identifies the procedure that should be used for the purge AND the actions that should be taken to prevent the actuation of an ESF Containment Ventilation Isolation (CVI) during the containment purge?

A. 0POP02-HC-0002, NORMAL CONTAINMENT PURGE; Increase the High alarm setpoint of RT-8012 & 8013 (RCB Purge Monitors).

B. 0POP02-HC-0002, NORMAL CONTAINMENT PURGE; Increase the High alarm setpoint on RT-8011 (Containment atmosphere radiation monitor).

C. 0POP02-HC-0003, SUPPLEMENTARY CONTAINMENT PURGE; Increase the High alarm setpoint of RT-8012 & 8013 (RCB Purge Monitors).

D. 0POP02-HC-0003, SUPPLEMENTARY CONTAINMENT PURGE; Increase the High alarm setpoint on RT-8011 (Containment atmosphere radiation monitor).

Proposed Answer: C Explanation (Optional):

A. INCORRECT. Normal Containment Purge is only when in modes 5, 6, or defueled.

Supplementary Purge may be used in any mode.

B. INCORRECT. Normal Containment Purge is only when in modes 5, 6, or defueled.

Supplementary Purge may be used in any mode.

C. CORRECT. To prevent the ESF actuation, RT-8012 and RT-8013 isolation setpoints may be raised for the supplementary purge, provided the setpoints are returned to 5.00 E-04 Ci/cc following the purge.

D. INCORRECT. There is no provision to allow for raising the alarm setpoint on RT-8011, only RT-8012 and RT-8013.

Technical Reference(s): 0POP02-HC-0002, Normal Containment Purge, Rev 26 0POP02-HC-0003, Supplementary Containment Purge, Rev 25 Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: Bank # 1281 Question History: Last NRC Exam 2005 Question Cognitive Level: Comprehension or Analysis Question Level of Difficulty: 3 10 CFR Part 55 Content: 55.41.5 55.45.3, 55.45.13 Comments:

Examination Outline Cross-

Reference:

Level RO 033 Spent Fuel Pool Cooling Tier # 2 Ability to monitor automatic operation of Group # 2 the Spent Fuel Pool Cooling System K/A # 033 A3.02 including:

Importance Rating 2.9 A3.02 Spent fuel leak or rupture Question 59 Spent Fuel Pool Cooling Pump A is in service providing cooling to the Spent Fuel Pool (SFP).

A large leak in the spent fuel cooling system piping occurs.

SFP level is 63 feet and lowering For the above conditions in the Spent Fuel Cooling system what engineering design feature helps to ensure that adequate level is maintained in the Spent Fuel Pool?

A. The SFP Cooling Pumps automatically trip < 63 feet B. The suction lines of the SFP Cooling Pumps are equipped with a vacuum breaker C. The return line into the SFP from the SFP Cooling System contains a hole (siphon breaker) in the pipe located approximately 6 below normal level D. The discharge lines of the SFP cooling pumps are physically located above the minimum required level in the SFP

Proposed Answer: C Explanation (Optional):

A. Incorrect This is a manual action that is taken at < 63 feet not an engineering design there is no automatic trip B. Incorrect, there is no vacuum break in the suction line for the Spent fuel pool cooling pumps C. Correct, the return line does contain a siphon break to prevent siphon action from taking place in the event of a leak in the spent fuel cooling system D. Incorrect, the discharge lines discharge below the minimum level Technical Reference(s): STPEGS USFAR (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Difficulty 2 10 CFR Part 55 Content: 55.41.7 55.45.9 Comments:

From USFAR 3.1.2.6.2.2 Spent Fuel Handling and Storage - Evaluation Against Criterion 61 - Irradiated fuel is stored underwater in spent fuel storage racks located at the bottom of the spent fuel pool.

Spent fuel pool water is circulated through the Spent Fuel Pool Cooling and Cleanup System (SFPCCS) to maintain fuel pool water temperature, purity, water clarity, and water level. The spent fuel storage racks preclude accidental criticality (see Evaluation Against Criterion 62).

Reliable decay heat removal is provided by the closed-loop SFPCCS, which consists of two cooling trains, two purification trains, a surface skimmer loop, and required piping, valves and instrumentation. Water is drawn from the spent fuel pool by the spent fuel pool pumps, is pumped through the tube side of the heat exchangers and is returned. Each suction line, which is protected by a strainer, is located at an elevation 4 ft below the normal water level, while the return line terminates in a sparger pipe at the bottom of the Spent Fuel Pool and contains an

antisiphon hole near the surface of the water to prevent gravity drainage. The SFPCCS is designed to remove the amount of decay heat produced by the number of spent fuel assemblies that are stored following refueling. Each train is capable of removing 100 percent of the normal maximum design heat load and 50 percent of the abnormal maximum design heat load. Table 9.1-1 gives the peak SFP temperatures calculated for various fuel heat load and SFP cooling configurations. System piping is arranged so that failure of any pipeline cannot drain the spent fuel pool or the in-Containment temporary storage area below a depth of approximately 23 ft of water over the top of the stored spent fuel assemblies. A minimum depth of approximately 13 ft of water over the top of an array of 193 (full core) assemblies with 42 hours4.861111e-4 days <br />0.0117 hours <br />6.944444e-5 weeks <br />1.5981e-5 months <br /> of decay is required to limit radiation from the assemblies to 2.5 mR/hr. or less.

Examination Outline Cross-

Reference:

Level RO 035 Steam Generator - Knowledge of the Tier # 2 effect of a loss or malfunction on the Group # 2 following will have on the S/GS: K/A # 035/K6.03 Importance Rating 2.6 K6.03 S/G level detector Question 60 Unit 1 is at 100% power when the controlling level channel for C Steam Generator fails to 0%.

Which of the following correctly describes the plant response assuming no operator action?

Feedwater flow to C Steam Generator will A. rise until a feedwater isolation will occurs.

B. lower until the reactor trips on Lo-Lo Steam Generator Level.

C. initially rise, then will lower resulting in a steam generator level stabilizing higher than before.

D. initially lower, then will rise resulting in a steam generator level stabilizing lower than before.

Proposed Answer: A Explanation (Optional):

A. CORRECT. A level error output signal is produced (176.25%), that now causes the MFRV to open. The increased flow produces a flow error, but it cannot overcome the level error output signal. Therefore the level continues to increase until a Feedwater Isolation Actuation occurs (2/4 S/G levels > 87.5%).

B. INCORRECT. Steam Generator Level increases in this failure, not decreases.

C. INCORRECT: FW flow will increase as indicated, but will not go back down because the controlling channel is failed low and will not change with actual level.

D. INCORRECT. When the controlling channel fails to 0% a level error is created that indicates the actual level is too low. This causes the MFRV to open, increasing FW flow, not decreasing it as indicated.

Technical Reference(s): LOT202.15, Steam Generator Water Level Control, Rev 9 Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: New 2382 Question History: Last NRC Exam N/A Question Cognitive Level: Comprehension or Analysis Question Level of Difficulty: 3 10 CFR Part 55 Content: 55.41.7 55.45.7 Comments:

Examination Outline Cross-

Reference:

Level RO 041 Steam Dump/Turbine Bypass Tier # 2 Control - Ability to predict and/or monitor Group # 2 changes in parameters (to prevent K/A # 041, A1.02 exceeding design limits) associated with Importance Rating 3.1 operating the SDS controls including:

A1.02 Steam pressure Question 61 A normal cooldown is being performed on Unit 2 with the following conditions:

• RCS temperature is 555 °F and decreasing at 50 °F/hr

• RCS pressure is 2000 psig and decreasing

• Cooldown is being performed using the Steam Dumps in AUTO

• All RCPs are in operation Which of the following would explain an increase in main steam pressure?

A. Steam header Pressure PT-557 failed low B. Steam header Pressure PT-557 failed high C. Turbine Impulse Pressure PT-505 failed low D. Turbine Impulse Pressure PT-505 failed high

Proposed Answer: A Explanation (Optional):

A. CORRECT. If PT-557 fails low, then the process variable will be less than the setpoint and the dump valves will close causing the plant to heat up and main steam pressure to rise.

B. INCORRECT. If PT-557 fails high, then the process variable will be greater than the setpoint and dump valves would open causing a greater cooldown and main steam pressure to decrease.

C. INCORRECT. Although PT-505 has an input into steam dumps in the Tave mode, in this condition the steam dumps are in steam pressure mode and a failure of PT-505 would have no impact on the steam dumps.

D. INCORRECT. Although PT-505 has an input into steam dumps in the Tave mode, in this condition the steam dumps are in steam pressure mode and a failure of PT-505 would have no impact on the steam dumps.

Technical Reference(s): LOT202.09, Steam Dump System, Rev 13 Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: Bank # 92 Question History: Last NRC Exam 2009 Question Cognitive Level: Comprehension or Analysis Question Level of Difficulty: 3 10 CFR Part 55 Content: 55.41.5 55.45.5 Comments:

Examination Outline Cross-

Reference:

Level RO 045 Main Turbine Generator Tier # 2 Knowledge of the effect that a loss or Group # 2 malfunction of the MT/G system will have K/A # 045 K3.01 on the following:

Importance Rating 2.9 K3.01 Remainder of the plant Question 62 Given the following:

• Unit 1 is operating at 48% power

• An electrical malfunction causes the Generator Field Breaker (41M) to trip Which of the following correctly describes the effect on the plant?

A. The Main Generator Output Breakers will open; the Turbine will continue to operate with no load. The Reactor will continue to operate B. The Main Generator Output Breakers will remain closed, a Turbine trip will occur, resulting in a reactor trip.

C. The Main Generator Output Breakers will open, a Turbine trip will occur. The Reactor will continue to operate.

D. The Main Generator Output Breakers will remain closed; the Turbine will continue to operate with no load. The Reactor will remain on line.

Proposed Answer: C Explanation (Optional):

A. Incorrect, the Turbine has tripped, Loss of field will cause a turbine trip B. Incorrect, 48% is below the P-9 setpoint of 50%, the Reactor will not trip, Main Generator Output breakers trip open on Loss of field C. Correct D. Incorrect, the Turbine has tripped, due to 86 lockout relay, Plasauble as the Reactor will not trip.

Technical Reference(s): LP.NO.:LOT202.17 Rev 10 (Main Generator),

5Z109Z42111 Reactor Trip Signals, LP.NO.: LOT202.03 Rev 13 (Main Turbine)

(Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Difficulty 2 10 CFR Part 55 Content: 55.41.7 55.45.6 Comments:

From Lesson Plan LP.NO.:LOT202.17 Rev 10 (Main Generator) 40-1/G1 and 40-2/G1 Generator Loss of Field Relays Located in CP010 40-1/G1 will trip the Turbine Trip Lockout Relay (86TT1) immediately after sensing a loss of field if Voltage Unbalance Relay 60-2/G1 is not tripped.

From Lesson Plan LP.NO.: LOT202.03 Rev 13 (Main Turbine) 4.6 Turbine Trips and Setpoints

4.6.1 Reactor Trip - P-16 4.6.2 Steam generator high high level - P-14, 87.5% Narrow range s/g level 4.6.3 Turbine overspeed, electrical - 110% of rated speed or 1980 rpm 4.6.4 Turbine overspeed, mechanical - 111% of rated speed or 1998 rpm 4.6.5 Thrust bearing wear excessive - + 40 mils from zero setpoint 4.6.6 Low condenser vacuum - 21.0 In HG vacuum 4.6.7 Low turbine bearing oil pressure - 6 psig 4.6.8 Low Stator Cooling Water Differential Pressure - 19 psid with a 40 second time delay 4.6.9 AMSAC Actuation 4.6.10 Electrical Equipment Protection - generator trip, 86 lockout relay 4.6.11 Manual trip locally at turbine 4.6.12 Manual trip from the control room (HS-6317A) 4.6.13 Safety Injection

Examination Outline Cross-

Reference:

Level RO 056 Condensate - Knowledge of the Tier # 2 physical connections and/or cause-effect Group # 2 relationships between the Condensate K/A # 056, K1.03 System and the following systems:

Importance Rating 2.6 K1.03 MFW Question 63 Which of the following actions automatically occur if both Feedwater Deaerator Storage Tank levels are at 92%?

A. Trip of the running Condensate Pumps B. Trip of the running Feed Water Booster Pumps C. Closing of the Deaerator Vent Condenser Temperature Control Valve (TV-7413)

D. Closing of the Condensate Block Valves to deaerator (MOV-574 and 575)

Proposed Answer: D Explanation (Optional):

The following are the only automatic actions in the event of Feedwater Deaerator Storage Tak HI-HI level:

1) Deaerator Hi Level Dump Valves LV-7175 and LV-7175A opened at 86% level.

2) FW Heater 11A & B Drain Valves to Deaerator LV-7242 and LV-7245 close.

3) Extraction Steam Non Return Valve 1-XV-2263 [ES-0041] and 2263A [ES-0183] close.

4) Extraction Steam Non Return Valve 1-XV-2260 [ES-0040] and 2260A [ES-0184] close.

5) Condensate Block Valves MOV-0574 and MOV-0575 close.

6) Deaerator Extraction Steam Block Valves MOV-0069 and MOV-0068 close.

A. Plausible as a tripped condensate pump could be the cause of the Hi-Hi Level.

B. Plausible as a tripped feed water booster pump could be the cause of the Hi-Hi Level.

C. Plausible as a closed temperature control valve could be the cause of the Hi-Hi Level.

Technical Reference(s): 0POP09-AN-07M3, Annunciator Lampbox 1(2)-7M03 Response Instructions Rev. 78 Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: Bank # 484 Question History: Last NRC Exam 1995 Question Cognitive Level: Memory or Fundamental Knowledge Question Level of Difficulty: 2 10 CFR Part 55 Content: 55.41.6 Comments:

Examination Outline Cross-

Reference:

Level RO 075 Circulating Water - Ability to (a) Tier # 2 predict the impacts of the following Group # 2 malfunctions or operations on the K/A # 075 A2.02 circulating water system; and (b) based on Importance Rating 2.5 those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Question 64 Given the following:

• Unit 1 is operating at 100% power.

• Circulating Water Pumps (CWP) 11, 12, and 13 are operating.

• CWP 14 is tagged out of service for maintenance.

Subsequently:

• CWP 11 trips due to an unknown reason.

Following the pump trip, the following parameters are noted:

• Condenser Vacuum - 22 inches Hg

• CWP 13 Motor Stator Temperature - 185 °F

• CWP 13 Motor Bearing Temperature - 185 °F In accordance with 0POP04-CW-0001, Loss of Circulating Water Flow, which of the following would be the first expected action?

A. Start all available Condenser Air Removal Pumps B. Secure CWP 13 C. Initiate a Main Turbine Load Reduction Per Addendum 2, Turbine Load Reduction, to restore Main Condenser Vacuum D. Trip the reactor, trip the turbine, and go to 0POP05-EO-EO00, Reactor Trip or Safety Injection.

Proposed Answer: A Explanation (Optional):

A. CORRECT. With power greater than 80%, and condenser vacuum between 21 and 24 inches Hg, the correct action is to start all available Condenser Air Removal Pumps B. INCORRECT. This is the correct step if the Circulating Water Pump Motor Stator temperature is greater than 250 °F or if the Circulating Water Pump Motor Bearing temperature is greater than 200 °F C. INCORRECT. This is the correct step if the Main Condenser Vacuum is NOT within Permissible Without Limit Zone Of Addendum 4, Main Turbine Exhaust Pressure Limitations.

D. INCORRECT. This is the correct step if condenser vacuum is less than 21 inches Hg.

Technical Reference(s): 0POP04-CW-0001, Loss of Circulating Water Flow, Rev 15 Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: New 2383 Question History: Last NRC Exam N/A Question Cognitive Level: Comprehension or Analysis Question Level of Difficulty: 3 10 CFR Part 55 Content: 55.41.5 55.43.5 55.45.3, 45.13 Comments:

Examination Outline Cross-

Reference:

Level RO SRO 086 Fire Protection - Ability to monitor Tier # 2 automatic operation of the Fire Protection Group # 2 System Including K/A # A3.01 A3.01 Starting Mechanisms of fire water Importance Rating 2.9 Question 65 All three Diesel Fire Pump controller switches are in AUTO. A transient in the fire water system occurs and fire water pressure decreases to 100 psi for approximately 10 seconds, before increasing to 128 psi, 13 seconds after the transient started. Assuming the jockey pump is running in auto, which of the following describes the condition of the fire pumps:

A. No Fire Pumps are running B. Only Fire Pump #1 is running C. Only Fire Pumps #1 and #2 are running D. Fire Pumps #1, #2 and #3 are running

Proposed Answer: B Explanation (Optional):

A. INCORRECT. Fire Pump #1 will be running.

B. CORRECT. Fire Pump #1 automatically starts when system pressure decreases below 130 psi for more than 5 seconds.

C. INCORRECT. Fire Pump #2 will not automatically start unless system pressure is below 120 psi for more than 15 seconds.

D. INCORRECT. Fire Pump #2 will not automatically start unless system pressure is below 120 psi for more than 15 seconds.

Technical Reference(s): LOT201.29, Fire Protection, Rev 6, Sections 2.3 and 2.4 Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: New 2384 Question History: Last NRC Exam N/A Question Cognitive Level: Comprehension or Analysis Question Level of Difficulty: 4 10 CFR Part 55 Content: 55.41.7 55.45.5 Comments:

Examination Outline Cross-

Reference:

Level RO 2.1 Conduct of Operations Tier # 3 Group # 1 2.1.18 Ability to make accurate, clear and K/A # 2.1.18 concise logs, records, status boards, and Importance Rating 3.6 reports.

Question 66 In accordance with Conduct of Operations Chapter 4, Reports and Notifications, which of the following plant events require notification to the NRC resident inspector?

A. Taking the reactor critical B. Turbine Trip below P-9 C. ESF Actuation D. Event having adverse effect beyond the site boundary

Proposed Answer: D Explanation (Optional):

All answers are credible because each event requires some sort of notification within the plant structure.

A. INCORRECT. Taking the reactor critical does not require notification to the NRC resident inspector.

B. INCORRECT. A turbine trip on its own (below P-9) does not require notification to the NRC resident inspector.

C. INCORRECT. ESF actuation does not require notification to the NRC resident inspector.

D. CORRECT. An event having an adverse effect beyond the site boundary does require notification to the NRC resident inspector, especially since it would require offsite notifications.

Technical Reference(s): Conduct of Operations, Chapter 4, Reports and Notification, Rev 17 Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: New 2385 Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge X Question Level of Difficulty: 3 10 CFR Part 55 Content: 55.41.10 55.43.12 55.45.13 Comments:

Examination Outline Cross-

Reference:

Level RO 2.1.21 Ability to obtain and verify Tier # 3 controlled procedure copy. Group # 1 K/A # 2.1.21 Importance Rating 3.1 Question 67 An evolution is expected to continue for several days. The procedure is labeled continual use.

What is the maximum amount of time that can elapse before an in-progress procedure must be Verified/Validated to be current per the requirements of 0PGP03-ZA-0010, Performing and Verifying Station Activities?

A. Each shift B. Every 7 days C. Every 14 days D. Only prior to start of the Job

Proposed Answer: A Explanation (Optional):

A. Correct, 0PGP03-ZA-0010, Performing and Verifying Station Activities directs continuous use procedures to be verified PRIOR TO USE B. Incorrect, creditable based on common sense and normal frequency of evolutions conducted at power plants C. Incorrect, creditable based on common sense and normal frequency of evolutions conducted at power plants D. Incorrect, creditable based on common sense and normal frequency of evolutions conducted at power plants Technical Reference(s): 0PGP03-ZA-0010, Performing and Verifying Station Activities (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination:

Learning Objective: (As available)

Question Source: New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Difficulty 2 10 CFR Part 55 Content: 55.41.12 55.45.10, 55.45.13 Comments:

0PGP03-ZA-0010, Performing and Verifying Station Activities PRIOR TO USE (with respect to verifying PROCEDURES in use are current): For continual use operational PROCEDURE sets, prior to use means each shift. For all other PROCEDURES, prior to use means within one shift prior to initiating PROCEDURE steps.

Examination Outline Cross-

Reference:

Level RO 2.1 Conduct of Operations Tier # 3 Group # 1 2.1.3 Knowledge of shift turnover practices K/A # 2.1.3 Importance Rating 3.7 Question 68 As the primary operator, according to procedure 0POP01-ZQ-0022, Plant Operations Shift Routines, which of the following activities MUST be performed prior to assuming the shift?

A. Safety Function Checklist B. Operator Burden Report C. Surveillance Schedule D. Radiation Monitoring Status

Proposed Answer: C Explanation (Optional):

A. INCORRECT: According to form 6, to be completed as early in the shift as practical.

B. INCORRECT: According to form 6, to be completed as early in the shift as practical.

C. CORRECT. According to form 6, to be completed prior to assuming the shift.

D. INCORRECT. According to form 6, to be completed as early in the shift as practical.

Technical Reference(s): 0POP01-ZQ-0022, Plant Operations Shift Routines, Rev 72 Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Question Level of Difficulty: 2 10 CFR Part 55 Content: 55.41.10 55.45.13 Comments:

Examination Outline Cross-

Reference:

Level RO 2.2 Equipment Control Tier # 3 Group # 2 2.2.13 Knowledge of tagging and K/A # 2.2.13 clearance procedures.

Importance Rating 4.1 Question 69 Which of the following is correct regarding breaker racking tags in accordance with Procedure 0PGP03-ZO-ECO1A, Equipment Clearance Order Instructions?

A. A breaker racking tag may not be placed over a danger, test or caution tag B. While a breaker racking tag is attached to a control room handswitch the operation of that switch is permitted C. The name of the person in the field controlling the evolution shall be written on the tag D. An Operator may go to the breaker, inspect the area and remove the breaker racking tag. The Operator will then report the status to the Shift Manager/Unit Supervisor

Proposed Answer: C Explanation (Optional):

According to Procedure 0PGP03-ZO-ECO1A, Equipment Clearance Order Instructions, Addendum 5, Rev 23 A. INCORRECT. A breaker tag MAY be placed over a danger, test, or caution tag B. INCORRECT. While a breaker racking tag is attached to a control room handswitch the operation of that switch is PROHIBITED C. CORRECT. The name of the person in the field controlling the evolution shall be written on the tag D. INCORRECT. The Shift Manger or Unit Supervisor or Unit Supervisor must authorize the Operator to remove the tag after the status has been reported.

Technical Reference(s): Procedure 0PGP03-ZO-ECO1A, Equipment Clearance Order Instructions, Revision 23 Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: New 2387 Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Question Level of Difficulty: 2 10 CFR Part 55 Content: 55.41.10 55.43.2 55.45.13 Comments:

Examination Outline Cross-

Reference:

Level RO 2.2 Equipment Control Tier # 3 Group # 2 2.2.22 Knowledge of limiting conditions for K/A # 2.2.22 operations and safety limits Importance Rating 4.0 Question 70 With fuel burnup of 50,000 MWD/MTU, what is the Technical Specification requirement for MAXIMUM peak fuel centerline temperature?

A. < 5080 °F B. < 5022 °F C. < 4964 °F D. < 4790 °F

Proposed Answer: D Explanation (Optional):

The maximum fuel centerline temperature allowed by Technical Specification is 5080 °F decreasing by 58 °F for every 10,000 MWD/MTU of burnup.

A. INCORRECT. Plausible if application does not know there is a reduction for core burnup.

B. INCORRECT. Plausible if applicant believes the reduction is 58 °F for every 50,000 MWD/MTU C. INCORRECT. Plausible if applicant believes the reduction is 58 °F for every 25,000 MWD/MTU D. CORRECT. With a reduction of 290 °F Technical Reference(s): Technical Specification 2.1.1 Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: New 2388 Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Question Level of Difficulty: 4 10 CFR Part 55 Content: 55.41.5 55.43.2 55.45.2 Comments:

Examination Outline Cross-

Reference:

Level RO 2.3.12 Knowledge of radiological safety Tier # 3 principles pertaining to licensed operator Group # 3 duties, such as containment entry K/A # 2.3.12 requirements, fuel handling Importance Rating 3.2 responsibilities, access to locked high-radiation areas, aligning filters, etc.

Question 71 The following Unit 2 conditions exist:

• During normal plant operations you are directed to perform a valve alignment on CCP A.

when you reach the door for CCP A you notice a radiation sign on the door.

• The radiation sign on the door reads DANGER - HIGH RADIATION AREA.

Which of the following are REQUIRED for entry into the area in accordance with 0PGP03-ZR-0051, Radiological Access Controls/Standards?

1. Be assigned to an RWP that permits entry.

2. Be assigned an individual monitoring device (TLD).

3. Be issued an Electronic Personal Dosimeter (EPD).

4. Be made knowledgeable of the radiological conditions in the area(s) to be accessed.

5. Be aware of additional Radiation Protection controls established by the RWP or RP instructions.

6. Obtain Key or be allowed access by the RP continually guarding the access point.

7. Obtain continuous RP Coverage

8. ALARA Review Committee SHALL approve the entry, before entry is made.

A. 1, 2, 3, 4, 5 only.

B. 1, 2, 3, 4, 5, 6 C. 1, 2, 3, 4, 5, 7 D. 1, 2, 3, 4, 5, 8

Proposed Answer: A Explanation (Optional):

A. Correct, IAW 0PGP03-ZR-0051, Radiological Access Controls/Standards section 6.5 items 1 through 5 are the requirements for entry into a high radiation area B. Incorrect, I tem 6 is not required for high radiation area access but is required for entry into a LOCKED high radiation area.

C. Incorrect, Item 7 is not required for entry into a high radiation area but is required for entry into a LOCKED high radiation area D. Incorrect, Item 8 in not required for entry into a high radiation area but is a requirement for a Very High Radiation Area.

Technical Reference(s): 0PGP03-ZR-0051 rev.29, Radiological Access Controls/Standards (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Difficulty 2 10 CFR Part 55 Content: 55.41.12 55.43.9 55.45.10 Comments:

0PGP03-ZR-0051 rev.29, Radiological Access Controls/Standards 6.5 Access Control for High Radiation Areas (HRA) 6.5.1 Personnel entering high radiation areas SHALL be:

6.5.1.1 Assigned to an RWP that permits entry to HRA, LHRA or VHRA.

6.5.1.2 Assigned an individual monitoring device (TLD). (10CFR20.1502).

6.5.1.3 Issued an Electronic Personal Dosimeter (EPD).

6.5.1.4 Made knowledgeable of the radiological conditions in the area(s) to be accessed.

6.5.1.5 Aware of additional Radiation Protection controls established by the RWP or RP instructions.

Examination Outline Cross-

Reference:

Level RO 2.3.4 Knowledge of radiation exposure Tier # 3 limits under normal or emergency Group # 2 conditions. K/A # G2.3.4 Importance Rating 3.2 Question 72 Area Radiation Monitor RE-8052 in the In-Core Instrumentation Room is alarming HIGH at 1000 mrem/hr. Local surveys have confirmed that this radiation level is accurate.

John, due to his expertise, will need to enter the area to conduct repairs.

• He was a contractor until he was hired by STP a few months ago.

• He is 52 years old with a lifetime TEDE of 40 rem.

• He has accumulated 3.6 rem TEDE this year, of which 360 mrem was received at STP.

The federal exposure limits imposed by 10CFR Part 20 require that Jim's stay time for the job must NOT exceed _________ without authorization for a Planned Special Exposure.

A. 0 minutes, John is already in excess of 2 REM this calendar year and special authorization would be required.

B. 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and 38.4 minutes.

C. 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and 2.4 minutes.

D. 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and 24 minutes,

Proposed Answer: D Explanation (Optional):

A. Incorrect, Plausible because student could believe that the 2 Rem limit is in affect for John and not just the 360 mrem he received at STP B. Incorrect, Student could perform the calculation based on 2000 mrem STP limit with 360 mrem already received, (2000-360)/1000 mrem/hour= 1.64 hours7.407407e-4 days <br />0.0178 hours <br />1.058201e-4 weeks <br />2.4352e-5 months <br /> or 98.4 min C. Incorrect, Student could add the 360 mrem received at STP to the total dose he has received to date and then calculate the stay time D. Correct, John has 1400 mrem left of the 5 rem federal limit therefor 1400/1000 = 1.4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or 84 min Technical Reference(s):

(Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination:

Learning Objective: (As available)

Question Source: Modified Bank # 111 Question History: Last NRC Exam N/A Question Cognitive Level: Comprehension or Analysis Question Level of Difficulty: 3 10 CFR Part 55 Content: 55.41.12 55.43.4 55.45.10 Comments:

Parent Question Area Radiation Monitor RE-8052 in the In-Core Instrumentation Room is alarming HIGH at 1200 mrem/hr. Local surveys have confirmed that this radiation level is accurate.

Jim Neutron, due to his expertise, will need to enter the area to conduct repairs.

• He was a contractor until he was hired by STP a few months ago.

• He is 52 years old with a lifetime TEDE of 40 rem.

• He has accumulated 3.6 rem TEDE this year, of which 360 mrem was received at STP.

The federal exposure limits imposed by 10CFR Part 20 require that Jim's stay time for the job must NOT exceed _________ without authorization for a Planned Special Exposure.

A. 0 minutes B. 20 minutes C. 57 minutes D. 70 minutes Exam Bank No.: 111 RO Outline Number:

K/A Catalog Number:061.G2.3.4 Tier: 1 Group/Category: 2 RO 2.5 10CFR

Reference:

55.41(b)()

Knowledge of radiation exposure limits and contamination control, including permissible levels in excess of those authorized.

STP Lesson: Objective Number:

Reference:

10 CFR 20.1201 Attached

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified from:

Distractor Justification A:

B:

C:

D: CORRECT: To exceed the federal limit of 5 rem per year a Planned Special Exposure must be authorized. Jim has 1400 mrem left; so he must be limited to 70 minutes (1400 mrem).

Examination Outline Cross-

Reference:

Level RO SRO 2.4.14 Knowledge of general guidelines Tier # 3 for EOP usage. Group # 4 K/A # 2.4.14 Importance Rating 3.8 Question 73 Place the following operations procedures in order of hierarchy.

1. Normal Operating Procedures (OP)

2. Off Normal Operating Procedures (ONP)

3. Annunciator Response Procedure (ARP)

4. Emergency Response Procedure (EOP)

A. 3, 4, 2, 1 B. 4, 2, 1, 3 C. 4, 2, 3, 1 D. 2, 4, 3, 1

Proposed Answer: C Explanation (Optional):

A. Incorrect, plausible because the ARPs often alert the operators of events when they occur and are often the first procedure that is referenced in an event, the remaining choices are in correct order B. Incorrect, plausible because the order of EOP,ONP, is correct and an operator could easily believe that the OPs are overriding of ARPs as during the course of normal operations a OP may cause an alarm that is referenced later C. Correct D. Incorrect, plausible because often a ONP is in progress first when conditions degrade requiring entry into an EOP all others are correct Technical Reference(s): 0POP01-ZA-0024 Rev 1 page 4 of 5 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: New Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Difficulty 2 Question Level of Difficulty: 2 10 CFR Part 55 Content: 55.41.10 55.43.5 55.45.13 Comments:

Examination Outline Cross-

Reference:

Level RO 2.4.16 Knowledge of EOP implementation Tier # 3 hierarchy and coordination with other Group # 4 support procedures or guidelines such as, K/A # 2.4.16 operating procedures, abnormal operating Importance Rating 3.5 procedures, and severe accident management guidelines.

Question 74 A small steam generator tube leak (15 gpm) has occurred on the A SG. The unit is reducing power to take the unit off line. During the power reduction an unexpected Unit trip has occurred.

The following conditions are noted immediately following the reactor trip.

• Reactor Power less than 35 E-6 and lowering

• SUR is negative

• RCS Sub-Cooling is 37 F

• All SGs are at 12% and slowly lowering with A SG NR level lowering at a slower rate than the other SGs

• Containment temperatures and pressures are normal

• Containment Radiation is 2 E+2 R/Hr

• Pressurizer level is 75%

• RVWL indicates 100% on plenum

• Main Feed Pumps have tripped

• No AFW is running

• SI has not initiated What Critical safety Function(s) is/are currently not met? ____(1)_____

During the performance of EO E000 Reactor Trip and SI actuation the CRS directs you to monitor critical safety functions and announces that he is transitioning to _______(2)________ .

(1) (2)

A. Heat Sink 0POP05-EO-FHR1 Loss of Heat Sink B. Core Cooling, Heat Sink 0POP05-EO-E030 Steam Generator Tube Rupture C. Heat Sink 0POP05-EO-E030 Steam Generator Tube Rupture D. Core Cooling, Heat Sink 0POP05-EO-FHR1 Loss of Heat Sink

Proposed Answer: A Explanation (Optional):

A. Correct Heat Sink is not met due to AFW flow less than 576 gpm total, Loss of Heat sink procedure is required B. Incorrect, Core cooling is currently met based on plant conditions having adequate (> 35 Degrees) SCM and normal post trip conditions, Transfer to the SGTR procedure is not required until after the critical safety function has been addressed C. Incorrect, The Loss of heat sink is correct but while a SGTR is occurring the critical safety function would be a higher priority and addressed first D. Incorrect, Core cooling is currently met based on plant conditions having adequate (>35 Degrees) SCM and normal post trip conditions Technical Reference(s):

• 0POP01-ZA-0018 Rev 19 Emergency Operating Procedure Users Guide

• 0POP05-EO-E000 Rev 20, Reactor Trip or Safety Injection

• 0POP05-EO-FO01 Rev 1, Subcriticality Critical Safety Function Status Tree

• 0POP05-EO-FO02 Rev 2, Core Cooling Critical Safety Function Status Tree

• 0POP05-EO-FO03 Rev 6, Heat Sink Critical Safety Function Status Tree

• 0POP05-EO-FO04 Rev 4, Integrity Critical Safety Function Status Tree

• 0POP05-EO-FO05 Rev 1, Containment Critical Safety Function Status Tree

• 0POP05-EO-FO06 Rev 1, Inventory Critical Safety Function Status Tree (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: New 2418 X Question History: Last NRC Exam N/A Question Cognitive Level: Comprehension or Analysis Difficulty 3 10 CFR Part 55 Content: 55.41.10 55.43.5 55.45.12 Comments:

Reference 0POP05-EO-FO03 Rev 6, Heat Sink Critical Safety Function Status Tree Reference 0POP05-EO-FO02 Rev 2, Core Cooling Critical Safety Function Status Tree Examination Outline Cross-

Reference:

Level RO

4. Emergency Procedures / Plan Tier # 3 2.4.17 Knowledge of EOP terms and Group # 4 definitions K/A # 2.4.17 Importance Rating 3.9 Question 75 Abnormal Operating Procedures (AOPs), and Emergency Operating Procedures (EOPs) will give direction to "GO TO" Directs the operator to a later step in the same procedure OR to a different procedure.

When directed to "GO TO" another procedure, which of the following describes this term?

A. Complete the controlling procedure and then enter the procedure that you were directed to "GO TO" B. Immediately exit the controlling procedure and then enter the procedure that you were directed to "GO TO" C. Continue with the controlling procedure and perform the procedure that you are directed to "GO TO" in conjunction with it D. Immediately enter and complete the procedure you were directed to "GO TO" then return and complete the controlling procedure

Proposed Answer: B Explanation (Optional):

A. Incorrect, GO TO - Directs the operator to a later step in the same procedure OR to a different procedure. WHEN this term is used, THEN the operator does NOT return unless directed to do so by a later step B. Correct C. Incorrect, See A. above D. Incorrect, See A. above Technical Reference(s): 0POP01-ZA-0018, Emergency Operating Procedure User's Guide, Rev 21 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: New 2419 X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Difficulty 2 10 CFR Part 55 Content: 55.41.10 55.45.13 Comments:

Reference 0POP01-ZA-0018, Emergency Operating Procedure User's Guide, Rev 21 2.18 GO TO - Directs the operator to a later step in the same procedure OR to a different procedure. WHEN this term is used, THEN the operator does NOT return unless directed to do so by a later step.

Examination Outline Cross-

Reference:

Level SRO EPE 009 Small Break LOCA - Ability to Tier # 1 determine or interpret the following as they Group # 1 apply to a small break LOCA: K/A # EPE009/EA2.34 Importance Rating 4.2 EA2.34 Conditions for throttling or stopping HPI Question 76 The following conditions exist in Unit 2:

• Mode 3 following a manual reactor trip and SI

• RCS pressure 1450 psig and STABLE

• Core Exit TCs are 555°F

• Pressurizer level is 48% and INCREASING

• Containment pressure is 6.5 psig

• SG NR levels are: 35%, 35%, 33%, 36%

• Total AFW flow is 400 gpm The Unit Supervisor is at Step 15 of 0POP05-EO-EO10, Loss of Reactor or Secondary Coolant, and is evaluating whether Safety Injection should be terminated.

Which of the following operator actions is procedurally required for the above conditions?

A. SI termination criteria is met only if AFW flow is adjusted to > 576 gpm. Do NOT transition to 0POP05-EO-ES11, SI Termination, until AFW flow is adjusted.

B. SI termination criteria is met only if the level in C Steam Generator is increased to 34%.

Do NOT transition to 0POP05-EO-ES11, SI Termination, C Steam Generator level is increased to 34%.

C. SI termination criteria is NOT met since pressurizer level is still low and further actions in 0POP05-EO-EO10 need to be performed.

D. SI termination criteria is NOT met since RCS subcooling is less than the required value and further actions in 0POP05-EO-EO10 need to be performed.

Proposed Answer: D Explanation (Optional):

Since containment pressure is greater than 5 psig, adverse containment conditions exist.

A. INCORRECT. AFW flow greater than 576 gpm is a condition to satisfy secondary heat sink, however, since at least 1 steam generator narrow range level is greater than 34%,

secondary heat sink is satisfied.

B. INCORRECT. In order for the secondary heat sink to be satisfied, at least 1 steam generator narrow range level must be greater than 34%, not in all steam generators, thus the secondary heat sink is satisfied.

C. INCORRECT. Pressurizer level must be greater than 44% to satisfy the SI termination criteria, which is satisfied.

D. CORRECT. Saturation temperature at 1450 psig (1465 psia) is approximately 593°F, resulting in RCS subcooling of approximately 38°F. In order to satisfy SI termination criteria in adverse containment, RCS subcooling must be greater than 45°F Technical Reference(s): 0POP05-EO-EO10, Loss of Reactor or Secondary Coolant, Rec 22 Proposed references to be provided to applicants during examination: Steam Tables Learning Objective: (As available)

Question Source: Bank # 230 Question History: Last NRC Exam 2001 Question Cognitive Level: Comprehension or Analysis Question Level of Difficulty: 4 10 CFR Part 55 Content: 55.43.5 55.45.13 Comments:

Examination Outline Cross-

Reference:

Level SRO 000011 Large Break LOCA / 3 Tier # 1 EA2 Ability to determine or interpret the Group # 1 following as they apply to a Large Break K/A # 011, EA2.01 LOCA:

Importance Rating 4.7 EA2.01 Actions to be taken, based on RCS temperature and pressure - saturated and superheated Question 77 You are the Control Room Supervisor on Unit 2

• A LOCA has occurred on Unit 2

• RCS pressured has lowered to 1420 psig and stabilized

• HHSI pumps are operating

• RCPs have not yet been secured.

• You have reached step 10 of 0POP05-E0-E000, Rx Trip or Safety Injection.

What is the basis for verifying that at least 1 HHSI pump is running and RCS pressure below 1430 psig?

A. Securing the RCPs without at least 1 HHSI pump operating can cause an ICC condition, below 1430 PSIG RCP operation is not required for cool down in any circumstance.

B. At least one HHSI pump is required when RCPs are secured to minimize potential thermal shock if HHSI pumps are started after RCPs are secured, 1430 psig is based on SBLOCA where a loss of the RCP later would result in core uncovery C. Securing the RCPs without at least 1 HHSI pump operating can cause an ICC condition, 1430 psig is based on a SBLOCA where a loss of the RCP later would result in core uncovery D. At least one HHSI pump is required when RCPs are secured to minimize potential thermal shock if HHSI pumps are started after RCPs are secured, below 1430 PSIG RCP operation is not required for cool down in any circumstance.

Proposed Answer: C Explanation (Optional):

RCP trip criteria is based on certain small break LOCA conditions where excessive depletion of RCS water inventory (due to running RCPs) through the break might lead to severe core uncovery if the RCPs were tripped for some reason later in the accident.

The setpoint is selected such that RCPs will be tripped for those conditions which require tripping RCPs, but will not be tripped for those conditions which would have enhanced recoveries if the RCPs remained running, i.e., SGTRs including the DBA tube rupture.

The requirement for having at least one HHSI pump running (and capable of delivering flow to the RCS) prior to tripping the RCPs is based on:

Analysis has shown that if the SI system is not in operation, operation of the RCPs will allow the RCS to be safely depressurized to the point where the accumulators and the LHSI pump can ensure core heat removal before symptoms of Inadequate Core Cooling (ICC) are exhibited. If the RCPs are then tripped during RCS depressurization because of a LOOP or other support condition, the depressurization rate can be increased to maximum to obtain the benefits of the accumulators and LHSI pumps sooner.

Technical Reference(s):STPEGS EOP Technical Guideline EOPT-03.01 Technical Guidelines for 0POP05-EO-EO00, Reactor Trip or Safety Injection WOG Emergency Response Guidelines - Low Pressure Volume 0POP05-EO-EO00, Reactor Trip or Safety Injection 0POP01-ZA-0018, Emergency Operating Procedure Users Guide (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Difficulty 3 10 CFR Part 55 Content: 55.43.5 55.45.13 Comments:

Examination Outline Cross-

Reference:

Level SRO EPE 029 Anticipated Transient Without Tier # 1 Scram (ATWS) Group # 1 K/A # EPE 029/G2.1.25 2.1.25 Ability to interpret reference Importance Rating 4.2 materials, such as graphs, curves, tables, etc.

Question 78 Given the following:

• Unit 1 is at 100% power with all systems in a normal lineup.

• Current RCS Boron concentration is 500 ppm.

• Cycle Burnup is 12000 MWD/MTU.

• It has just been determined that 2 Shutdown Bank Control Rods are mechanically bound (untrippable).

Prior to beginning a Unit Shutdown the following occurs:

• Main Feedwater Reg Valve on SG 2B fails closed.

• Both Reactor Trip switches fail to make control rods drop into the core.

• The Control Room crew places Feeder Breaker Handswitches for 480V Load Centers 2K1 and 2L1 to OPEN and then back to NORMAL.

• Feeder Breaker indication for 480V Load Center 2K1 shows a red light lit.

• Feeder Breaker indication for 480V Load Center 2L1 shows a green light lit.

Based on the given conditions, in which of the following procedures would the Unit Supervisor FIRST direct that an Emergency Boration be performed AND what would be the correct amount of Emergency Boration to perform?

FIRST procedure the US should direct an Emergency Borate the RCS Emergency Boration be performed.

A. 0POP05-EO-ES01, Reactor Trip Response. with 1880 gallons of boric acid B. 0POP05-EO-FRS1, Response to Nuclear to 1398 ppm Power Generation - ATWS.

C. 0POP05-EO-ES01, Reactor Trip Response. with 7200 gallons of boric acid D. 0POP05-EO-FRS1, Response to Nuclear to 836 ppm Power Generation - ATWS.

Proposed Answer: B Explanation (Optional):

A. INCORRECT. 0POP05-EO-ES01 is not the correct procedure to use because the given conditions indicate an ATWS event. The amount of borated water to add is also incorrect for the associated procedure because it's based on 2 stuck rods below a position of 18 steps. The question stem indicates there are 2 stuck rods at a position greater than 18 steps.

B. CORRECT. The conditions given are indicative of the reactor failing to trip from the Control Room which requires entry into 0POP05-EO-FRS1. The procedure will have the US emergency borate to Plant Curve Book Figure 5.5, 68 degree F. curve which requires a boron concentration of 1398 ppm for the given cycle burnup.

C. INCORRECT. 0POP05-EO-ES01 is not the correct procedure to use because the given conditions indicate an ATWS event has occurred. The amount of borated water to add would be correct for the associated procedure for 2 stuck rods at a position greater than 18 steps as provided in the question stem.

D. INCORRECT: This ppm is from the Plant Curve Book Figure 5.5, 567 degree F. curve, not the 68 degree F curve as required.

Technical Reference(s): 0POP05-EO-FRS1 Response to Nuclear Power Generation - ATWS, Rev 17 0POP05-EO-ES01, Reactor Trip Response, Rev. 26 Plant Curve Book Figure 5.5, Shutdown Margin Limit Curve, Unit 1, Cycle 19, Rev 28 Proposed references to be provided to applicants during examination: Plant Curve Book Figure 5.5, Shutdown Margin Limit Curve, Unit 1, Cycle 19 Learning Objective: (As available)

Question Source: Bank # 2140 Question History: Last NRC Exam 2011 Question Cognitive Level: Comprehension or Analysis Question Level of Difficulty: 3 10 CFR Part 55 Content: 55.41.10 55.43.5 55.45.12 Comments:

PARENT QUESTION:

Given the following:

• Unit 2 is at 100% power with all systems in a normal lineup.

• Current RCS Boron concentration is 550 ppm.

• Cycle Burnup is 14000 MWD/MTU.

• It has just been determined that 2 Shutdown Bank Control Rods are mechanically bound (untrippable).

Prior to beginning a Unit Shutdown the following occurs:

• Main Feedwater Reg Valve on SG 2B fails closed.

• Both Reactor Trip switches fail to make control rods drop into the core.

• The Control Room crew places Feeder Breaker Handswitches for 480V Load Centers 2K1 and 2L1 to OPEN and then back to NORMAL.

• Feeder Breaker indication for 480V Load Center 2K1 shows a red light lit.

• Feeder Breaker indication for 480V Load Center 2L1 shows a green light lit.

Based on the given conditions, in which of the following procedures would the Unit Supervisor FIRST direct that an Emergency Boration be performed AND what would be the correct amount of Emergency Boration to perform?

FIRST procedure the US should Borate RCS direct an Emergency Boration be performed.

A 0POP05-EO-FRS1, Response to to 701 ppm Nuclear Power Generation - ATWS.

B 0POP05-EO-ES01, Reactor Trip with 1880 gallons of boric acid Response.

C 0POP05-EO-FRS1, Response to to 1303 ppm Nuclear Power Generation - ATWS.

D 0POP05-EO-ES01, Reactor Trip with 7200 gallons of boric acid Response.

Examination Outline Cross-

Reference:

Level SRO 000038 Steam Gen. Tube Rupture / 3 Tier # 1 Group # 1 APE: 038 Steam Generator Tube Rupture K/A # 038 G2.2.25 G2.2.25 Knowledge of the basis in the Importance Rating 4.2 Technical Specifications for limiting conditions and safety limits Question 79 Which of the following is the basis for the requirement to cooldown the plant to below 500 F within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> when the specific activity of the reactor coolant exceeds Technical Specification limits?

A. Maintains the fission product gases in solution in the reactor coolant.

B. Increases reliability of the data collected for iodine release.

C. Prevents the release of activity in the event of a steam generator tube rupture.

D. Ensures on-site exposures will not exceed 10CFR20 limits.

Proposed Answer: C Explanation (Optional):

A. Incorrect, while having a colder water does allow gasses to remain in solution this is not the concern for this basis B. Incorrect, This has no effect on the reliability of data collected C. Correct D. Incorrect, this does not ensure that on-site exposure limits are met, exposure on site will go up due to the RCS activity, the cooldown is to ensure that the safety relief valves do not open in the event that a SGTR occurs Technical Reference(s): T.S. 3.4.8 bases (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination:

Learning Objective: (As available)

Question Source: Bank # NRC 185 Question History: Last NRC Exam Not on Previous NRC exam part of the NRC exam bank from STP Question Cognitive Level: Memory or Fundamental Knowledge Difficulty 3 10 CFR Part 55 Content: 55.43.2 Comments:

Tech Spec Basis 3.4.8 Reducing Tavg to less than 500°F prevents the release of activity should a steam generator tube rupture since the saturation pressure of the reactor coolant is below the lift pressure of the atmospheric steam relief valves. The Surveillance Requirements provide adequate assurance that excessive specific activity levels in the reactor coolant will be detected in sufficient time to take corrective action. A reduction in frequency of isotopic analyses following power changes may be permissible if justified by the data obtained.

Parent:

Which ONE of the following is the basis for the requirement to cooldown the plant to below 500 F within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> wh Specification limits?

A. Maintains the fission product gases in solution in the reactor coolant.

B. Increases reliability of the data collected for iodine spiking studies.

C. Prevents the release of activity in the event of a steam generator tube rupture.

D. Ensures on-site exposures will not exceed 10CFR20 limits.

Correct Answer: C

Examination Outline Cross-

Reference:

Level SRO APE: 062 Loss of Nuclear Service Water Tier # 1 Group # 1 2.4.18 Knowledge of the specific bases for K/A # APE: 062, 2.4.18 EOPs.

Importance Rating 4.0 Question 80 While performing the steps in procedure 0POP04-EW-0001, Loss of Essential Cooling Water, a subsequent loss of Component Cooling Water limits the operation time of the Centrifugal Charging Pumps.

Which of the following is the most limiting event?

A. Supplementary Cooler resulting in exceeding the maximum bearing temperature limits causing pump bearing damage.

B. Supplementary Cooler resulting in a breakdown of electrical insulation and shorting of the motor.

C. Lube Oil Cooler resulting in exceeding the maximum bearing temperature limits causing pump bearing damage.

D. Lube Oil Cooler resulting in a breakdown of electrical insulation and shorting of the motor.

Proposed Answer: B Explanation (Optional):

A. INCORRECT. Plausible because the student has to know that the Supplementary Cooler only cools the air in the CCP room and not directly the bearings.

B. CORRECT. Loss of cooling to the Supplementary Cooler allows extreme heat to build up in the relatively small room causing electrical insulation damage that will short the motor in as little as 4 minutes making it the most limiting event.

C. INCORRECT. Plausible because this is a limiting event but it takes about 8 minutes so it is not the most limiting event.

D. INCORRECT. Plausible because some large motors do have lube oil coolers that directly cool the motor oil such as the RCPs which in turn could cause electrical insulation damage if cooling was lost. But the CCP lube oil coolers only directly cool the pump bearing oil.

Technical Reference(s): 0POP04-EW-0001, Loss of Essential Cooling Water, Rev. 1 Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: New Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Question Level of Difficulty: 4 10 CFR Part 55 Content: 55.41.10 55.43.1 55.45.13 Comments:

Examination Outline Cross-

Reference:

Level SRO 000065 Loss of Instrument Air / 8 Tier # 1 AA2. Ability to determine and interpret the Group # 1 following as they apply to the Loss of K/A # 0065 AA2.05 Instrument Air:

Importance Rating 4.1 AA2.05 When to commence plant shutdown if instrument air pressure is decreasing Question 81 Unit is at 100% power when the IA header starts to lower, attempts to identify the reason for the IA leak have been unsuccessful up to this point.

IA header pressure is 60 psig and lowering slowly In accordance with 0POP04-IA-0001 Loss of Instrument Air as the CRS you should direct the board operators to _________

A. Trip the Rx, enter 0POP05-EO-EO00, Reactor Trip or Safety Injection and continue with the actions of 0POP04-IA-0001, Loss of Instrument Air, as resources permit.

B. Identify portion of IA system with high flow and isolate IAW 0POP04-IA-0001, Loss of Instrument Air C. Open Service Air Isolation Valve IA-PV-9785 to allow service air to supply IA, IAW with step 6 0f 0POP0-IA-001 Loss of Instrument Air D. Commence a power reduction in preparation to take the unit off line as the Loss of Instrument Air has started to affect the Feedwater control valves.

Proposed Answer: A Explanation (Optional):

A. Correct, 0POP04-IA-0001 step 1 has you monitor if 60 psig and then trip the reactor B. Incorrect, this would be correct action if IA pressure was greater than 60 pisg C. Incorrect, this valve would be closed per the procedure D. Incorrect, the feedwater control valves are affected at 67 psig but attempting to adjust them is non-conservative Technical Reference(s): 0POP04-IA-0001 Rev 16 Loss of Insturment (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: New 2421 X Question History: Last NRC Exam N/A Question Cognitive Level: Comprehension or Analysis Difficulty 3 10 CFR Part 55 Content: 55.43.5 Comments:

Reference 0POP4-IA-0001

Examination Outline Cross-

Reference:

Level SRO 074 Inadequate Core Cooling Tier # 1 Group # 2 2.4.21 Knowledge of the parameters and K/A # 074, 2.4.21 logic used to assess the status of safety Importance Rating 4.6 functions, such as reactivity control, core cooling and heat removal, reactor coolant system integrity, containment conditions, radioactivity release control, etc.

Question 82 While reviewing the core cooling critical safety function, the following conditions exist:

• The plant is in mode 3

• A small break LOCA has occurred

• Applicable actions of 0POP05-EO-EO00 have been completed

• Core exit thermocouple temperatures are 1150 °F

• RCS core exit thermocouples indicate 5°F superheat

• Reactor vessel water plenum level is 15%

Which of the following should be directed by the unit supervisor:

Enter Procedure A. 0POP05EOEO10, Loss OF Reactor OR Secondary Coolant B. 0POP05-EO-FRC1, Response to Inadequate Core Cooling C. 0POP05-EO-FRC2, Response to Degraded Core Cooling D. 0POP05-EO-FRC3, Response to Saturated Core Cooling

Proposed Answer: C Explanation (Optional):

A. INCORRECT. The critical safety function is not satisfied.

B. INCORRECT. Entry into 0POP05-EO-FRC1 occurs at core exit thermocouple temperatures > 1200 °F.

C. CORRECT. Entry into 0POP05-EO-FRC2 occurs if RCS subcooling is < 35 °F AND plenum level is < 20% AND core exit thermocouple temperatures > 708 °F.

D. INCORRECT. Entry into 0POP05-EO-FRC3 occurs if RCS subcooling is < 35 °F AND either plenum level is > 20% OR core exit thermocouple temperatures < 708 °F.

The path through the status tree leads to 0POP05-EO-FRC2 Technical Reference(s): 0POP05-EO-FO02 Core Cooling Critical Safety Function Status Tree, Rev 2 Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: New 2389 Question History: Last NRC Exam N/A Question Cognitive Level: Comprehension or Analysis Question Level of Difficulty: 3 10 CFR Part 55 Content: 55.41.7 55.43.5 55.45.12 Comments:

Examination Outline Cross-

Reference:

Level SRO 000032 Loss of Source Range NI / 7 Tier # 1 APE: 032 Loss of Source Range Nuclear Group # 2 Instrumentation K/A # 032 G2.1.20 2.1.20 Ability to interpret and execute Importance Rating 3.4 procedure steps Question 83 Given the following Unit 1 conditions:

• Refueling operations in progress.

• Extended Range Flux Monitor N46 is OPERABLE.

• Extended Range Flux Monitor N45 is INOPERABLE.

• Source Range N31 is in service with its associated audible indication in the Control Room and Containment OPERABLE.

• Maintenance and just been completed on Source Range N32 and the detector is Bypassed waiting on post maintenance testing to be completed.

• Source Range N31 has just failed LOW.

The Control Room Supervisor entered 0POP04-NI-001 Nuclear Instrument Malfunction and proceeds to addendum 1, Source Range Nuclear Instrument Malfunction.

Step 6 of addendum 1 states:

CHECK Nuclear Instrumentation Channels (N31, N32, N45 or N46) Meets the Following:

• One Source Range With Audible Indication

• Additional Source Range or Extended Range Operable Based on Step 6 of addendum 1 above the Control Room supervisor should ensure which of the following actions:

A. Immediately Restore Extended Range Flux Monitor N45 to operable status to meet the requirements of step 6.

B. Initiate visual monitoring of the Extended Range Nuclear Instruments to meet the requirements of ITS 3.9.2.

C. Direct Chemistry to sample the RCS boron concentration at least once every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> per ITS 3.9.2 D. Immediately suspend core alterations.

Proposed Answer: D Explanation (Optional):

A. Incorrect, Check does not mean to restore in the world of procedure use. restoring Extended Range Nuclear Instrument N45 to Operable would not satisfy the requirements of step 6 and is not directed in 0POP04-NI-001.

B. Incorrect, While visual monitoring of the extended range Nuclear Instrument is required it is done per the RNO in step 5 of 0POP04-NI-001 and visual monitoring of extended range NIs do not meet the requirements of ITS 3.9.2 C. Incorrect, RCS sampling a minimum of once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is also a required action of 0POP04-NI-001; it is directed per the RNO of step 5. RCS sampling is required per the technical specification when both of the required source range Nuclear instruments are not operable. In this case NI45 is Operable and therefore the requirement to sample within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is not required D. Correct the RNO of step 6 requires suspending core alterations.

Technical Reference(s): TS 3.9.2, 0POP04-NI-001 Nuclear Instrument Malfunction, addendum 1, Source Range Nuclear Instrument Malfunction (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: New X Question History: Last NRC Exam None Question Cognitive Level: Memory or Fundamental Knowledge Difficulty 3 10 CFR Part 55 Content: 55.43.7 Comments:

Reference TS 3.9.2 3.9.2 As a minimum, two Source Range Neutron Flux Monitors' shall be OPERABLE, each with continuous visual indication in the control room and one with audible indication in the containment and control room.

APPLICABILITY: MODE 6.

ACTION:

a. With one of the above required monitors inoperable or not operating, immediately suspend all operations involving CORE ALTERATIONS or operations that would cause

introduction into the RCS of coolant with boron concentration less than required to meet the boron concentration of LCO 3.9.1.

b. With both of the above required monitors inoperable or not operating, determine the boron concentration of the Reactor Coolant System at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

Reference 0POP04-NI-0001

Examination Outline Cross-

Reference:

Level SRO W/E 16 High Containment Radiation - Ability to Tier # 1 determine and interpret the following as they Group # 2 apply to the High Containment Radiation K/A # W/E16, EA2.1 EA2.1 Facility conditions and selection of Importance Rating 3.3 appropriate procedures during abnormal and emergency operations.

Question 84 Plant conditions are as follows:

• RWST Level = 175,000 gallons

• Containment pressure = 9.2 psig

• Containment water level = 62 inches

• Containment radiation level = 2500 R/hr 0POP05-EO-EO00, Reactor Trip or Safety Injection, and 0POP05-EO-E10, Loss of Reactor or Secondary Coolant, have already been entered. According to procedure 0POP05-EO-FO05, Containment Critical Safety Function Status Tree, which procedure is appropriate for the above conditions?

A. 0POP05-EO-ES13, Transfer to Cold Leg Recirculation B. 0POP05-EO-FRZ1, Response to High Containment Pressure C. 0POP05-EO-FRZ2, Response to Containment Flooding D. 0POP05-EO-FRZ3, Response to High Containment Radiation Level

Proposed Answer: D Explanation (Optional):

A. INCORRECT. Entry into 0POP05-EO-ES13 occurs at RWST level < 75,000 gallons B. INCORRECT. Entry into 0POP05-EO-FRZ1 occurs at containment pressure > 9.5 psi C. INCORRECT. Entry into 0POP05-EO-FRZ2 occurs at containment water level > 69 in D. CORRECT. Entry into 0POP05-EO-FRZ3 occurs at containment radiation level > 2000 R/h Technical Reference(s): 0POP05-EO-F005, Containment Critical Safety Function Status Tree, Rev 2; 0POP05-EO-E010, Loss of Reactor or Secondary Coolant, Rev 22 Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: Bank # 1099 Question History: Last NRC Exam 2003 Question Level of Difficulty: 3 Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 55.43.5 55.45.13 Comments:

Examination Outline Cross-

Reference:

Level SRO 000076 High Reactor Coolant Activity / 9 - Tier # 1 Ability to determine and interpret the Group # 2 following as they apply to the High K/A # 076 AA2.02 Reactor Coolant Activity:

Importance Rating 3.4 AA2.02 Corrective actions required for high fission product activity in RCS.

Question 85 Unit 1 is operating at 100%

RT-8039 (RM-11 CRT) CVCS Letdown Failed Fuel Monitor indications are rising rapidly and are in alarm.

A source Check of the RT indicates that it is operating properly All CVCS demineralizers are operating properly A RCS Sample has been taken and indications are that DEI is 10 micro curries per gram The specific activity limits in technical specification 3.4.8 are based on ensuring that the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> limit at the site boundary will not exceed a small fraction of the 10CFR100 limits concurrent with a ____(1)______ and for the given plant conditions above an emergency declaration ____(2)______ be entered.

(1) (2)

A. 150 gpd Steam generator tube rupture, Should NOT B. 25 gpm RCS leak outside containment, Should NOT C. 150 gpd Steam generator tube rupture, Should D. 25 gpm RCS leak outside containment, Should

Proposed Answer: C Explanation (Optional):

A. INCORRECT. correct basis for the specific activity limits but with indication of failed fuel an emergency classification is warrented B. INCORRECT. 25 gpm RCS leak is an emergency classification threshold but is not the basis for the specific activity limit C. CORRECT.

D. INCORRECT. 25 gpm RCS leakage is an emergency classification threshold but it not the basis for the specific activity limits Technical Reference(s): 0POP04-RC-0001,High Reactor Coolant System Activity (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: 0ERP01-ZV-IN01, Emergency Classification, Page 2 and 3 of the Emergency Classification Tables Learning Objective: (As available)

Question Source: New 2423 X Question History: Last NRC Exam N/A Question Cognitive Level: Comprehension or Analysis Difficulty 3 10 CFR Part 55 Content: 55.43.4 55.43.1 Comments:

ITS 3.4.8 3/4.4.8 SPECIFIC ACTIVITY The limitations on the specific activity of the reactor coolant ensure that the resulting 2-hour doses at the SITE BOUNDARY will not exceed an appropriately small fraction of 10 CFR Part 100 dose guideline values following a steam generator tube rupture accident in conjunction with an assumed steady-state reactor-to-secondary steam generator leakage rate of 150 gpd per steam generator. The values for the limits on specific activity represent limits based upon a parametric evaluation by the NRC of typical site locations. These values are conservative in that specific site parameters of the STPEGS site, such as SITE BOUNDARY location and meteorological conditions, were not considered in this evaluation.

Reference:

0POP04-RC-0001,High Reactor Coolant System Activity STEP: IMPLEMENT Actions Of 0ERP01-ZV-IN01, Emergency Classification.

PURPOSE: To refer the operator to the appropriate Emergency Plan Classification to ensure the applicable E-plan declaration is made, if required.

BASIS: The entry conditions for this procedure correspond to potential plant conditions that could represent criteria that may meet an Emergency Action Level (EAL) of emergency classification procedure, 0ERP01-ZV-IN01 Emergency Classification. Tables within the Emergency Classification procedure give guidance for determining Emergency Action Levels that may apply.

ACTIONS:

De te rm ine if RCS a ctivity lim its ha ve e xce e de d the Fue l Cla d De gra da tion EAL thre s hold values as read on RT-8039, or RCS sample results.

If thre s hold va lue s ha ve be e n e xce e de d, m a e k the appropriate E-plan declaration.

Examination Outline Cross-

Reference:

Level SRO 010 Pressurizer Pressure Control System Tier # 2 Group # 1 2.2.12 Knowledge of surveillance K/A # 010, 2.2.12 procedures.

Importance Rating 4.1 Question 86 Regarding the surveillance requirements for the Pressurizer PORVs, which of the following are requirements to satisfy Surveillance Requirement 4.0.5?

A. Full exercise in both directions including a stroke time measured in the open direction only.

B. Full exercise in both directions including a stroke time measured in the closed direction only.

C. Full exercise in the open Direction only including a stroke time measured in the open direction only.

D. Full exercise in the closed Direction only including a stroke time measured in the closed direction only.

Proposed Answer: A Explanation (Optional):

A. CORRECT. Full exercise in both directions including a stroke time measured in the Open direction only is required.

B. INCORRECT. Full exercise in both directions is required, but a stroke time measured in the Closed direction is not required.

C. INCORRECT. Full exercise is required in both directions.

D. INCORRECT. Full exercise is required in both directions.

Technical Reference(s): 0PSP03-RC-0010, Pressurizer Power Operated Relief Valve Operability Test, Rev 13 Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: New 2390 Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Question Level of Difficulty: 3 10 CFR Part 55 Content: 55.43.2 55.45.13 Comments:

Examination Outline Cross-

Reference:

Level SRO 013 Engineered Safety Features Tier # 2 Actuation Group # 1 APE: 013 Engineered Safety Features K/A # 00013 G 2.1.7 Actuation System (ESFAS)

Importance Rating 4.7 2.1.7 Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation.

Question 87 Unit 1 is operating at power when the following occur in sequence:

• A Loss of Coolant Accident (LOCA) occurs

• All ESF equipment is functioning as designed

• Containment Phase A Isolation is reset

• ESF Load Sequencers are reset in the Control Room only

• Containment Pressure increases to 9.8 psig.

• 0POP05-EO-EO00, Reactor Trip or Safety Injection, is complete including Addendum 5.

• A Reactor Operator reports that no Containment Spray Pumps are running, but their discharge valves are open.

Based on these conditions, what actions would the Unit Supervisor be required to implement?

A. In accordance with 0POP05-EO-FRZ1, Response to High Containment Pressure, direct the Reactor Operator to manually start the Containment Spray Pumps using their individual pump control switches.

B. In accordance with 0POP05-EO-FRZ1, Response to High Containment Pressure, direct the Reactor Operator to manually actuate Containment Spray by using the Containment Spray Manual Actuation Switches.

C. In accordance with 0POP05-EO-EO10, Loss of Reactor or Secondary Coolant, direct the Reactor Operator to manually actuate Containment Spray by using the Containment Spray Manual Actuation Switches.

D. In accordance with 0POP05-EO-EO10, Loss of Reactor or Secondary Coolant, direct the Reactor Operator to manually start the Containment Spray Pumps using their individual pump control switches.

Proposed Answer: A Explanation (Optional):

The student must recognize that the ESF Load Sequencers were reset before Containment pressure increased to above the Containment Spray actuation setpoint thereby disabling the actuation logic both automatically and manually. Thus the Containment Spray Pumps would have to be started manually with their control switches. Candidate must also know the applicable procedure that directs the actions to be taken.

Technical Reference(s): 0POP05-EO-FRZ1, Response to High Containment Pressure (Rev 9, Step 4)

(Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: Bank # Question 87 of 2005 Exam Question History: Last NRC Exam 2005 Question Cognitive Level: Comprehension or Analysis Difficulty 3 10 CFR Part 55 Content: 55.41.5 55.43.5 55.45.12 55.45.13 Comments:

Reference:

0POP05-EO-FRZ1, Response to High Containment Pressure

Examination Outline Cross-

Reference:

Level SRO 039 Main and Reheat Steam Tier # 2 Group # 1 2.2.38 Knowledge of conditions and K/A # 039, 2.2.38 limitations in the facility license.

Importance Rating 4.5 Question 88 The atmospheric steam relief valve automatic controls must be OPERABLE with a nominal setpoint of 1225 psig in Modes 1 and 2 because the safety analysis assumes automatic operation of the atmospheric steam relief valves with a nominal setpoint of 1225 psig with uncertainties for mitigation of which accident?

A. Small break LOCA.

B. Feedwater line break.

C. Loss of all main feedwater pumps.

D. Loss-of-offsite power.

Proposed Answer: A Explanation (Optional):

Distractors are all plausible because the atmospheric relief valves provide protection against all four of the listed accidents.

A. CORRECT. Small break LOCA.

B. INCORRECT. Auxiliary Feedwater line break.

C. INCORRECT. Loss of all main feedwater pumps.

D. INCORRECT Loss-of-offsite power.

Technical Reference(s): Tech Spec Bases Document (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: New 2391 Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Question Level of Difficulty: 4 10 CFR Part 55 Content: 55.41.7, 55.41.10 55.43.1 55.45.13 Comments:

Examination Outline Cross-

Reference:

Level SRO 061 Auxiliary/Emergency Feedwater Tier # 2 A2 Ability to (a) predict the impacts of the Group # 1 following malfunctions or operations on K/A # 061 A2.04 the AFW; and (b) based on those Importance Rating 3.8 predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

A2.04 pump failure or improper operation.

Question 89 Given the following:

• ESF DG #11 is out of service

• A loss of offsite power (LOOP) resulted in a reactor trip

• The crew has entered 0POP05-EO-ES01, Reactor Trip Response

• AFW Pump 14 trips on overspeed

• SG NR Levels are 16-21 %

Assuming no operator actions have been taken with AFW, which one of the below correctly describes the effect on the AFW system AND the actions to be taken by the Unit Supervisor?

The loss of AFW Pumps # 11 and 14...

A. could cause runout conditions on AFW Pumps 12 and 13. The Unit Supervisor should continue in 0POP05-EO-ES01 and throttle flow through AFW Pumps 12 and 13 to prevent runout.

B. will have no effect on the flow of AFW Pumps 12 and 13. The Unit Supervisor should continue in 0POP05-EO-ES01 and have the operator cross-tie AFW to feed SGs A and D.

C. could cause runout conditions on AFW Pumps 12 and 13. The Unit Supervisor MUST transition to 0POP05-EO-FRH1, Loss of Secondary Heat Sink, to re-establish heat sink requirements for AFW flow and SG levels.

D. will have no effect on the flow of AFW Pumps 12 and 13. The Unit Supervisor MUST transition to 0POP05-EO-FRH5, Response to Steam Generator Low Level, to re-establish AFW to SGs A and D.

Proposed Answer: B Explanation (Optional):

A. INCORRECT - The AFW Trains are independent following an actuation, thus the loss of two pumps will not affect the operation of the remaining two (which would not be true if the trains were cross-tied).

B. CORRECT - The AFW Trains are independent following an actuation. Based on these conditions, the US should continue in ES01.

C. INCORRECT - The AFW Trains are independent following an actuation, thus the loss of two pumps will not affect the operation of the remaining two (which would not be true if the trains were cross-tied). The given conditions do not warrant entry into FRH1.

D. CORRECT - The given conditions do not warrant entry into FRH5 (SG levels are greater than 14%).

Technical Reference(s): POP05-EO-FO03, R5; LOT202.28, AFW (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: 92206 LOT 504.33 Question Source: Bank # NRC 1727 Question History: Last NRC Exam 2007 Question Cognitive Level: Comprehension or Analysis Difficulty 3 10 CFR Part 55 Content: 55.41.5 55.43.5 55.43.3 55.45.13 Comments:

Examination Outline Cross-

Reference:

Level SRO 076 Service Water System - Ability to (a) Tier # 2 predict the impacts of the following Group # 1 malfunctions or operations on the SWS; K/A # 076, A2.01 and (b) based on those predictions, use Importance Rating 3.7 procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Question 90 Given the following:

The unit is at 100% power, the ECW PUMP 1B TRIP alarm is received.

The reactor operator reports that the previously running ECW Pump 1B is now stopped.

Which of the following describes the procedural action to be taken by the Unit Supervisor?

A. Declare ECW Pump 1B inoperable. B Train ESF Diesel Generator and Essential Chiller may remain operable provided the ESSENTIAL CHILLER X-CONN valves are opened per POP02-EW-0001, Essential Cooling Water Operations.

B. Declare B Train ECW Pump and ESF Diesel Generator inoperable. B Train Essential Chiller may remain operable provided the ESSENTIAL CHILLER X-CONN valves are opened per POP02-EW-0001, Essential Cooling Water Operations.

C. Declare B Train ESF Diesel Generator, Essential Chiller and ECW Pump inoperable.

Place the ECW Pump in PULL-TO-LOCK but leave the ESF Diesel Generator and Essential Chiller functional and available for emergency use per the annunciator response procedure.

D. Declare B Train ESF Diesel Generator, Essential Chiller and ECW Pump inoperable.

Place the ESF Diesel Generator in PULL-TO-STOP and the ECW Pump and Essential Chiller in PULL-TO-LOCK per the annunciator response procedure.

Proposed Answer: D Explanation (Optional):

A. INCORRECT. The DG and chiller are supported equipment and must be declared inoperable. Opening the x-tie valves would supply cooling to these components and POP02-EW-0001 allows the x-tie valves to be opened, but only for system fill.

B. INCORRECT. The chiller is supported equipment and must be declared inoperable.

Opening the x-tie valves would supply cooling to the chiller and POP02-EW-0001 allows the x-tie valves to be opened, but only for system fill.

C. INCORRECT. The annunciator directs placing all supported equipment in PTL to prevent operation without cooling.

D. CORRECT. ECW is a support system for both the ESF DG and Essential Chiller so they are inoperable when ECW is inoperable. The annunciator response procedure has the operators place the ECW pump and any supported equipment in PTL to prevent operation without cooling.

Technical Reference(s): POP09-AN-02M4, Annunciator Lampbox 2M04 Response Instructions, Rev. 29 Proposed references to be provided to applicants during examination:

Learning Objective: (As available)

Question Source: Bank # 1636 Question History: Last NRC Exam 2007 Question Cognitive Level: Comprehension or Analysis Question Level of Difficulty: 3 10 CFR Part 55 Content: 55.41.5 55.43.5 55.45.3, 45.13 Comments:

Examination Outline Cross-

Reference:

Level SRO 014 Rod Position Indication Tier # 2 A2 Ability to (a) predict the impacts of the Group # 2 following malfunctions or operations on K/A # 0014 A2.04 the RPIS; and (b) based on those on Importance Rating 3.9 those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Question 91 You are the Unit supervisor on Unit 2 and control rod testing is in progress when a safety rod is inserted 13 steps and then will not move. All attempts to move the control rod have been unsuccessful. The system engineer has reported that the control rod is mechanically bound.

In accordance with Technical Specification 3.4.1.3 Movable Control Assemblies, Continued Power operations _______________ continue and one of the required actions required to be taken is to_________________.

A. May, Declare the control Rod Inoperable within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and align all other rods in that bank within 12 steps of the misaligned control rod B. May, Align the remaining rods in that bank within 12 steps of the misaligned rod within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and A reevaluation of each accident analysis of Table 3.1-1 is performed within 7 days.

C. May Not, Verify SDM within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and be in Hot Standby within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> D. May Not, Verify SDM within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Mode 4 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />

Proposed Answer: C Explanation (Optional):

A. Incorrect, if control rod was considered trippable this would be a correct answer however with the control rod mechanically bound it is not considered trippable.

B. Incorrect, Control Rod is not considered trippable and this action does not apply.

C. Correct, a non trippable control rod falls into action A of ITS 3.1.3.1 D. Incorrect, there is no requirement to descend to mode 4 for this condition.

Technical Reference(s): Technical Specification 3/4.1.3 Movable Control Assemblies (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: New X Question History: Last NRC Exam Question Cognitive Level: Comprehension or Analysis Difficulty 3 10 CFR Part 55 Content: 55.43.2 55.43.5 55.45.3 55.45.13 Comments:

Reference from TS 3.1.3

Examination Outline Cross-

Reference:

Level SRO 016 Non-nuclear Instrumentation Tier # 2 Group # 2 2.2.37 Ability to determine operability K/A # 016, 2.2.37 and/or availability of safety related Importance Rating 4.6 equipment.

Question 92 The reactor is at full power.

A surveillance test was completed on PT-456, showing a safety injection action signal occurs at a setpoint of 1854 psig.

Pressurizer pressure instrument PT-455 just failed low and actions have been completed in response to the failure.

ESFAS pressurizer pressure Channel II is ___(1)___. Technical Specification 3.3.2 required action for the above condition is to ____(2)_____.

A. (1) Operable; (2) Bypass ESFAS pressurizer pressure Channel I within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> B. (1) Operable; (2) Place ESFAS pressurizer pressure Channel I in trip within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> C. (1) Inoperable; (2) Restore at least one of the inoperable ESFAS pressurizer pressure channels to OPERABLE within one hour D. (1) Inoperable; (2) Enter TS. 3.0.3

Proposed Answer: B Explanation: Per the TS surveillance requirement, 1854 psig is in the allowable range, although it is below the trip setpoint value of 1857 psig. This is allowed by action A. on the TS.

Therefore, Channel II is operable, leaving one channel inoperable. Per Action 20, the channel must be placed in trip within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. Answer A is plausible because the channel can be bypassed for up to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. Answer C is plausible because it is less than the trip setpoint, but is within the allowable range. Also, restoring a channel to operable is a TS action (Action 20A),

but it doesnt apply to this instrument. Answer D is plausible because it is less than the trip setpoint, and would be the correct answer if both channels were inoperable (there is no action statement for more than one inoperable channel for this instrument).

Technical Reference(s): TS 3.3.2 Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: New Question History: Last NRC Exam N/A Question Cognitive Level: Comprehension or Analysis Question Level of Difficulty: 3 10 CFR Part 55 Content: 55.41.

55.43.2 55.45.

Comments:

Examination Outline Cross-

Reference:

Level SRO A1 Ability to predict and/or monitor Tier # 2 changes in parameters (to prevent Group # 2 exceeding design limits) associated with K/A # 034 A1.01 operating the Fuel Handling System Importance Rating 3.2 controls including:

A1.01 Load limits Question 93 A fuel assembly SHALL not be disengaged from the refueling machine in the core until all of the following conditions are met in accordance with 0POP08-FH-0001, Refueling Machine Operating Instruction:

A. * "SLACK CABLE" light is ON

• Load cell indicates 435 lbs.

• Z-Axis tape measure shows the fuel assembly to be fully down

• Source range count rate is stable

• Directed by Core Loading Supervisor B. * "SLACK CABLE" light is OFF

• Load cell indicates 485 lbs.

• Z-Axis tape measure shows the fuel assembly to be fully down

• Source range count rate is stable

• Directed by Core Loading Supervisor C. * "SLACK CABLE" light is ON

• Load cell indicates 485 lbs.

• Z-Axis tape measure shows the fuel assembly to be fully down

• Source range count rate is stable

• Directed by Core Loading Supervisor D. * "SLACK CABLE" light is OFF

• Load cell indicates 435 lbs.

• Z-Axis tape measure shows the fuel assembly to be fully down

• Source range count rate is stable

• Directed by Core Loading Supervisor

Proposed Answer: A Explanation (Optional):

A. Correct, IAW 0POP08-FH-0001, Refueling Machine Operating Instruction B. Incorrect, Load Cell Less than 440 lbs C. Incorrect, Load Cell Less than 440 lbs D. Incorrect, load Cell Less than 440 lbs Slack cable should be on Technical Reference(s): 0POP08-FH-0001, Refueling Machine Operating Instruction (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: New 2425 X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Difficulty 2 10 CFR Part 55 Content: 55.43.7 Comments:

Reference:

0POP08-FH-0001, Refueling Machine Operating Instruction

Examination Outline Cross-

Reference:

Level SRO 2.1 Conduct of Operations Tier # 3 Group # 1 2.1.36 Knowledge of procedures and K/A # 2.1.36 limitations involved in core alterations.

Importance Rating 4.1 Question 94 Given the following:

Date / Time Activity 9/5/2015 0000 Plant Shutdown commenced.

9/5/2015 0630 Mode 3 Entry.

9/5/2015 1320 Mode 4 Entry.

9/5/2015 2210 Mode 5 Entry.

9/7/2015 2200 First Reactor Vessel Head Stud detensioned.

9/8/2015 0900 Reactor Vessel Head removed.

Which ONE of the following is the EARLIEST time that irradiated fuel movements may commence in accordance with 0POP08-FH-0009, Core Refueling?

A. 9/9/2015 at 1030.

B. 9/10/2015 at 0210.

C. 9/12/2015 at 0200.

D. 9/12/2015 at 1300.

Proposed Answer: A Explanation (Optional):

A. CORRECT. Reactor must be subcritical for 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> prior to irradiated fuel movement.

Mode 3 trip breakers open in the moment of subcriticality.

B. INCORRECT. Plausible because it is 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> from a milestone in the shutdown to Refueling. In this case, Mode 5 conditions. Applicant may consider 76 hours8.796296e-4 days <br />0.0211 hours <br />1.256614e-4 weeks <br />2.8918e-5 months <br /> from beginning of other plant conditions, such as Mode 5 (Water temperature is below boiling), Mode 6 (first head stud detensioned), and vessel head removed.

C. INCORRECT. Plausible because it is 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> from a milestone in the shutdown to Refueling. In this case, Mode 6 conditions. Applicant may consider 76 hours8.796296e-4 days <br />0.0211 hours <br />1.256614e-4 weeks <br />2.8918e-5 months <br /> from beginning of other plant conditions, such as Mode 5 (Water temperature is below boiling), Mode 6 (first head stud detensioned), and vessel head removed.

D. INCORRECT. Plausible because it is 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> from a milestone in the shutdown to Refueling. In this case, Mode 6 conditions in preparation for cavity fill. Applicant may consider 76 hours8.796296e-4 days <br />0.0211 hours <br />1.256614e-4 weeks <br />2.8918e-5 months <br /> from beginning of other plant conditions, such as Mode 5 (Water temperature is below boiling), Mode 6 (first head stud detensioned), and vessel head removed.

Technical Reference(s): 0POP08-FH-0009, Core Refueling, Rev. 41 Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: Bank # Wolf Creek Question History: Last NRC Exam 2009 Question Cognitive Level: Comprehension or Analysis Question Level of Difficulty: 3 10 CFR Part 55 Content: 55.41.10 55.43.6 55.45.7 Comments:

Examination Outline Cross-

Reference:

Level SRO 2.1.1 Knowledge of conduct of operations Tier # 3 requirements. Group # 1 (CFR: 41.10 / 45.13) K/A # G 2.1.1 Importance Rating 4.2 Question 95 Load Center 1J2 Breaker 11A Fuel Handling Heating Coil has tripped open. Mechanics are working in the Fuel Handing Building and have reported that it is cold and would like permission to reset the breaker.

IAW Conduct of Operations Chapter 2, as the Unit Supervisor you ________________.

A. Authorize the Mechanics to reset the breaker one time B. Tell Maintenance that they are not allowed to reset the breaker because there is no apparent cause for the breaker trip C. Determine/Understand the cause of the breaker tip and log, ensure that no additional conditions exist that would prevent closing breaker and then authorize operators to reset the breaker one time D. Tell the Mechanics that they are allowed to reset as the breaker as there is no apparent cause and the equipment is needed for plant operation per control room supervisor discretion

Proposed Answer: C Explanation (Optional):

A. Incorrect, IAW Conduct of Operations, If the affected equipment is needed to respond to a plant transient, then thermal overload devices or protective relay devices may be reset one time at the discretion of the Shift/Unit Supervisor. It is not needed for a plant transient.

B. Incorrect, IAW Conduct of Operations, If a MCC breaker trips and no apparent cause for the trip is identified, the Shift/Unit Supervisor can direct the breaker to be re-closed if the equipment is necessary for operations. If possible, cycle the load on the breaker and monitor for trips. The heating coil is not necessary for plant operations C. Correct, The cause of the trip should be understood and recorded, Annunciators and/or plant conditions associated with the initiating event should be recorded, Evaluation of expected/not expected indications should be performed, Equipment Condition is checked to verify that no abnormal conditions exist that would prevent reset.

D. Incorrect, See above you are allowed to authorize a reset if the conditions of Answer C are met.

Technical Reference(s): Conduct of Operations, Chapter 2 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: New X Question History: Last NRC Exam Question Cognitive Level: Comprehension or Analysis Difficulty 3 10 CFR Part 55 Content: 55.43.6 Comments:

Examination Outline Cross-

Reference:

Level SRO 2.2 Equipment Control Tier # 3 Group # 2 2.2.18 Knowledge of the process for K/A # 2.2.18 managing maintenance activities during Importance Rating 3.9 shutdown operations, such as risk assessments, work prioritization, etc.

Question 96 Which of the following should be designated as the Shutdown Risk Assessment Group (SRAG)

Leader, per 0PGP03-ZA-0101, Shutdown Risk Assessment?

A. Operations Manager or designee B. Supervisor Engineering Risk Management or designee C. Licensing Manager or designee D. Outage Manager or designee

Proposed Answer: A Explanation (Optional):

A. CORRECT. The operations manager or designee is the SRAG Leader.

B. INCORRECT. Plausible since an engineer from the Nuclear Fuel and Analysis PRA Risk Management group is a SRAG member.

C. INCORRECT. Plausible since a licensing representative is a SRAG member.

D. INCORRECT. Plausible since an outage representative is a SRAG member.

Technical Reference(s): 0PGP03-ZA-0101, Shutdown Risk Assessment, Rev. 28 Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: New Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Question Level of Difficulty: 2 10 CFR Part 55 Content: 55.43.5 Comments:

Examination Outline Cross-

Reference:

Level SRO 2.2.6 Knowledge of the process for Tier # 3 making changes to procedures. Group # 1 (CFR: 41.10 / 43.3 / 45.13) K/A # G 2.2.6 Importance Rating 3.6 Question 97 Your Crew has been assigned a schedule task to perform a valve lineup on cooling water system in preparation for startup after a refueling outage. A Modification has occurred during the outage that replaced a check valve on a heat exchanger outlet with an isolation valve.

While performing the valve line up the building operator that you assigned to the task asks you a question as to the position of the newly installed isolation valve.

NLO: The position for the new isolation valve is installed but it does not indicate an open or closed position. In the past there would have been a flow path through the system when the lineup is complete. There is a procedure change that is approved but not issued that has the valve open. The approved change is scheduled to be issued in 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. How do you want me to disposition this component?

In order to complete the valve lineup you A. Direct the building operator to sign off the procedure as complete as the valve is installed as the approved procedure states.

B. Have the building operator verify the valve is open, and then sign off the procedure because the valve is installed.

C. Have the building operator use the new procedure and complete the alignment.

D. Direct the building operator, with assistance from the Procedure Group, to perform a Field Change to the procedure and change the position to open then complete the procedure.

Proposed Answer: D Explanation (Optional):

A. INCORRECT. While the approved procedure would be complete it would not meet the intent of the valve alignment.

B. INCORRECT. While this would create the correct valve alignment it would not establish the correct status control for the system C. INCORRECT. The use of the new approved but not issued procedure would not be correct as the new procedure has not been issued.

D. CORRECT. Per 0PAP-01-ZA-0102 Plant Procedures allows the use of one time filed Change to correct a procedure to the same requirements of a new version of the procedure that is approved but not issued Technical Reference(s): 0PAP01-ZA-0102 Plant Procedures (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: New 2427 X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Difficulty 2 10 CFR Part 55 Content: 55.41.10 55.43.3 55.45.13 Comments:

Examination Outline Cross-

Reference:

Level SRO 2.3 Radiation Control Tier # 3 Group # 3 2.3.13 Knowledge of radiological safety K/A # 2.3.13 procedures pertaining to licensed operator Importance Rating 3.8 duties, such as response to radiation monitor alarms, containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc.

Question 98 Which of the following is correct regarding the initial containment inspection, per 0PSP03-XC-0002, Initial Containment Inspection to Establish Integrity?

A. The Test Coordinator need NOT be XC2 Initial Cleanup and Closure Certified.

B. Covered and non-covered personnel ARE ALLOWED to operate the Personnel Access Door.

C. Personnel initialing or signing surveillance Form 9 SHALL be XC2 Initial Cleanup and Closure Certified.

D. Plant Operators performing Lineups 1 and 2 ARE exempted fromXC2 Initial Cleanup and Closure XC2INCLSR certification.

Proposed Answer: D Explanation (Optional):

A. INCORRECT. The Test Coordinator will be a Senior Reactor Operator or someone designated by the Shift Manager; individual SHALL be XC2 Initial Cleanup and Closure Certified in Qual King. (Course code XC2INCLSR in LMS)

B. INCORRECT. Only covered personnel can operate PAL (non-covered personnel needing entry will require a covered worker to operate PAL. Consider a designated PAL operator depending on resources)

C. INCORRECT. Personnel initialing or signing ANY part of surveillance except Form 9 SHALL be XC2 Initial Cleanup and Closure Certified in Qual King.

D. CORRECT. Plant Operators performing Lineups 1 and 2 are exempted fromXC2 Initial Cleanup and Closure XC2INCLSR certification.

Technical Reference(s): 0PSP03-XC-0002, Initial Containment Inspection to Establish Integrity, Rev. 57 Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: New Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Question Level of Difficulty: 3 10 CFR Part 55 Content: 55.41.12 55.43.4 55.45.5, 45.11 Comments:

Examination Outline Cross-

Reference:

Level SRO 2.4.30 Knowledge of events related to Tier # 3 system operation/status that must be Group # 4 reported to internal organizations or K/A # G 2.4.30 external agencies, such as the State, the Importance Rating 4.1 NRC, or the transmission system operator.

Question 99 While performing outage preps a scaffold builder inadvertently trips open the normal supply breaker to the A ES 4160 volt bus, the Emergency Diesel Generator automatically starts and ties in to re-energize the bus.

All systems actuated and responded as designed.

In the control room all required support systems have been restored and the Unit is stable at 100% power.

Based on the above conditions about the event, what is the maximum time allowed to make a notification to the NRC?

(NOTE: Reports are written and the time requirements are in days. Notifications are by phone via the ENS and are in hours. Technically for this event it requires an 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> ENS notification and a 60 day written report [LER].)

A. 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> B. 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> C. 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> D. 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />

Proposed Answer: C Explanation (Optional):

A. Incorrect, but plausible, 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> is one of the required notification times in 10CFR72 and the event does require a notification.

B. Incorrect, but plausible, 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is one of the required notification times in 10CFR72 and the event does require a notification.

C. Correct, IAW 10CFR72

• (iv)(A) Any event or condition that results in valid actuation of any of the systems listed in paragraph (b)(3)(iv)(B) of this section, except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation,

• (8) Emergency ac electrical power systems, including: Emergency diesel generators (EDGs); hydroelectric facilities used in lieu of EDGs at the Oconee Station; and BWR dedicated Division 3 EDGs.

D. Incorrect, but plausible, 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is one of the required notification times in 10CFR72 and the event does require a notification.

Technical Reference(s): 10CFR72 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: Section IVa of STP Reporting Manual (60 page section)

Learning Objective: (As available)

Question Source: New 2428 X Question History: Last NRC Exam Question Cognitive Level: Comprehension or Analysis Difficulty 3 10 CFR Part 55 Content: 55.43.1 Comments:

Examination Outline Cross-

Reference:

Level SRO 2.4.40 Knowledge of SRO responsibilities Tier # 3 in emergency plan implementation. Group # 4 K/A # 2.4.40 Importance Rating 4.5 Question 100 You are the Shift Manager acting as the Emergency Director. Which of the following Emergency Director responsibilities and authorities may you delegate?

A. Approving required communications with the NRC.

B. Approving required notifications to the State and County.

C. Approving radiological exposures in excess of 10CFR20 limits.

D. Approving departure from license conditions per 10CFR50.54(x).

Proposed Answer: A Explanation (Optional):

A. CORRECT. The SM may delegate approval of communications with the NRC.

B. INCORRECT. Credible because this is an ED's responsibility (but not one that is proceduralized as delegable).

C. INCORRECT. Credible because this is an ED's responsibility (but not one that is proceduralized as delegable).

D. INCORRECT. Credible because this is an ED's responsibility (but not one that is proceduralized as delegable).

Technical Reference(s): 0ERP01-ZV-SH01, Shift Manager Rev. 30 Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: Bank # 413 Question History: Last NRC Exam 1997 Question Cognitive Level: Memory or Fundamental Knowledge Question Level of Difficulty: 3 10 CFR Part 55 Content: 55.41.10, 41.12 55.43.5 55.45.11 Comments: