EXAMINATION ANSWER KEY - NRC: Home Page · • 24 4KV bus trips on a ground fault • The plant trips on Loss of Load • RCS pressure peaks at 2410 PSIA During EOP-0, the following

1

EXAMINATION ANSWER KEY LOI 2015 NRG RO Exam (June 2016)

ID: Q106473 Points: 1.00

U-2 is at 100% when the following transient occurs:

• Multiple alarms are received • You are directed to evaluate if RPS is calling for a trip

Which ONE of the following RPS Channel A Matrix Relay conditions, if any, indicate that RPS is calling for a trip?

~' -. .

. ·.con,dition 4o· MATRIX RELAYS: ..

A. Conditions 1, 2, 3, and 4.

B. Conditions 2, 3, and 4 ONLY.

C. Conditions 3 and 4 ONLY.

D. None.

Answer: C

OPERATIONS Page: 1 of 235 13 April 2016

OPERATIONS

EXAMINATION ANSWER KEY LOI 2015 NRC RO Exam (June 2016)

Answer Explanation:

A. Incorrect-All conditions is plausible to the Operator since any one light should result in the opening of Trip Circuit Breakers (TCBs). There are four matrix relay lights for each of the six logics. Any one matrix light out on the left two lights (AB-1/AB-2) along with any one matrix light out on right two lights (AB-3/AB-4) will result in enough TCB breakers opening to cause a reactor trip.

B. Incorrect-Conditions 2/3/4 is plausible to the Operator since this will result in at least two matrix relay lights out for each of these conditions and RPS trip logic for the cabinets is 2 out of 4 channels. However, the 2 out of 4 does not apply to the matrix relay lights. There are four matrix relay lights for each of the six logics. Any one matrix light out on the left two lights (AB-1/AB-2) along with any one matrix light out on right two lights (AB-3/AB-4) will result in enough TCB breakers opening to cause a reactor trip.

C. Correct-There are four matrix relay lights for each of the six logics. Any one matrix light out on the left two lights (AB-1/AB-2) along with any one matrix light out on right two lights (AB-3/AB-4) will result in enough TCB breakers opening to cause a reactor trip.

D. Incorrect-None is plausible to the Operator since all matrix relay lights must be out to open all TCBs. There are four matrix relay lights for each of the six logics. Any one matrix light out on the left two lights (AB-1/AB-2) along with any one matrix light out on right two lights (AB-3/AB-4) will result in enough TCB breakers opening to cause a reactor trip.

Page: 2 of 235 13 April 2016


Question 1 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 0.00

System ID: 106473 User-Defined ID: 0106473 Cross Reference Number:

Topic: Evaluate RPS if callinQ for a trip RO Importance: 4.3 SRO Importance: 4.5 Cognitive level (High or

H Low):



NRC K/A Info, References: Tier/Group 1/1

K/A Info EPE 007 - Reactor Trip EA2 Ability to determine or interpret the following as they apply to a reactor trip: EA2.06 Occurrence of a reactor trip

RO 4.3 Importance

Technical FSAR Chapter 7 Figure 7-1. References

References None provided

Learning Recall the RPS response to a Trip Signal, Objective including Trip Units, Logic Matrices, Trip

Paths, TCBs

Question New Source

Question None History

Cognitive Comprehension or Analysis Level

10CFR 55.41(b)(7) Part 55 Content

Comments None

CCNPP Task(s) 201.010 Determine if the reactor should have tripped automatically but did not (A TWS)



ID: 0106474 Points: 1.00

U-2 is at 100% power when the following transient occurs:

• 24 4KV bus trips on a ground fault • The plant trips on Loss of Load • RCS pressure peaks at 2410 PSIA

During EOP-0, the following conditions are noted:

• RCS is 1800 PSIA and continues to lower • Quench Tank temperatures and pressure are rising rapidly • Containment temperature and pressure are stable • RCS temperature is 532°F and stable

Which ONE of the following conditions explain the response to the transient?

A. 1) Both PORVs should NOT have opened. 2) At least one PZR Safety Valve is relieving pressure to the Quench Tank.

OPERATIONS

B. 1) Both PORVs would have opened and relieved pressure to the Quench Tank. 2) At least one PORV is still relieving pressure to the Quench Tank.

C. 1) Only one PORV would have opened and relieved pressure to the Quench Tank. 2) The one PORV is still relieving pressure to the Quench Tank.

D. 1) Only one PORV would have opened but would NOT have relieved pressure to the Quench Tank. 2) At least one Safety Valve is relieving pressure to the Quench Tank.

Answer: C

Page: 5 of 235 13 April 2016

OPERATIONS


Answer Explanation:

A. Incorrect-Both PORVs not opening is plausible to the Operator if they confuse PORV lift setpoint with PZR safety setpoints (>2475 PSIA). RCS pressure exceeding 2400 PSIA will result in a High RCS Pressure trip and a signal for the ERVs to open. With the loss of 24 bus, MCC-204 is lost, and ERV-404 loses power. Without power, ERV-404 can not open. Only ERV-402, powered from MCC-214, would have lifted and passed flow through a deenergized, but open, PORV Block MOV-403

B. Incorrect-Both PORVs opening is plausible to the Operator if they believe 2B DG will reenergize the 24 4Kv bus. The DG breaker will not close in on a bus with a fault. RCS pressure exceeding 2400 PSIA will result in a High RCS Pressure trip and a signal for the ERVs to open. With the loss of 24 bus, MCC-204 is lost, and ERV-404 loses power. Without power, ERV-404 can not open. Only ERV-402, powered from MCC-214, would have lifted and passed flow through a deenergized, but open, PORV Block MOV-403.

C. Correct-RCS pressure exceeding 2400 PSIA will result in a High RCS Pressure trip and a signal for the ERVs to open. With the loss of 24 bus, MCC-204 is lost, and ERV-404 loses power. Without power, ERV-404 can not open. Only ERV-402, powered from MCC-214, would have lifted and passed flow through a deenergized, but open, PORV Block MOV-403.

D. Incorrect-A PORV opening but not passing flow is plausible to the Operator if loss of power to PORV Block MOV-403 also meant the valve was shut (similar to ERV affects on loss of power.) RCS pressure exceeding 2400 PSIA will result in a High RCS Pressure trip and a signal for the ERVs to open. With the loss of 24 bus, MCC-204 is lost, and ERV-404 loses power. Without power, ERV-404 can not open. Only ERV-402, powered from MCC-214, would have lifted and passed flow through a deenergized, but open, PORV Block MOV-403.

Page: 6 of 235 13 April 2016



System ID: 106474 User-Defined ID: Q106474 Cross Reference Number:

Topic: Determine expected PORV status on a Loss of Load trip with a 4Kv bus loss

RO Importance: 4.1 SRO Importance: 4.1 Cognitive level (High or

H Low): NRC KIA Info, References: Tier/Group 1/1

KIA Info APE 008 -Pressurizer (PZR) Vapor Space Accident (Relief Valve Stuck Open) 2.1 Conduct of Operations 2.1.28 Knowledge of the purpose and function of major system components and controls.

RO 4.1 Importance

Technical TS 3.4.10, AOP-71-2, OP-AA-108-114, 01-References 27D-2


Learning List the signals that will actuate a PORV Objective under all conditions/modes

Question Modified from Q28869 Source



10CFR 55.41 (b)(3) Part 55 Content

Comments None


3


1~IT_a_sk~~~J_2_02_.0_1_9_1s_o_la_te_P_O_R_V_s~~~~~~~'I

.. · Points: 1.00

A Small Break Loss of Coolant Accident has occured.

Which ONE of the following 1) are the operational implications if natural circulation is established instead of maintaining forced circulation and 2) what are the consequences if reflux boiling is established as the heat removal method?

A. 1) Main PZR spray will not be available 2) SGs are no longer effective at removing any heat

B. 1) Main PZR spray will not be available. 2) RCS inventory will be below the top of the hot leg.

C. 1) Allowable cooldown rate is lower, increasing time until SOC can be established. 2) SGs are no longer effective at removing any heat.

0. 1) Allowable cooldown rate is lower, increasing time until SOC can be established. 2) RCS inventory will be below the top of the hot leg.

Answer: B


OPERATIONS


Answer Explanation:

A Incorrect 1) True. Per EOP-5, with no RCPs in operation, Main PZR spray is not available. 2) False. S/Gs not available as a heat sink is plausible to the Operator if they believe boiling in the core is the only heat transfer mechanism occurring during reflux boiling. During reflux boiling, RCS level will be below the top of the hot leg to allow the reflux process to occur. Steam leaving the core is condensed in the SG U-Tubes and that condensate than travels back to the core via the hot leg to repeat the heat removal process.

B. Correct 1) True. Per EOP-5, with no RCPs in operation, Main PZR spray is not available. 2) True. During reflux boiling, RCS level will be below the top of the hot leg to allow the reflux process to occur. Steam leaving the core is condensed in the SG U-Tubes and that condensate than travels back to the core via the hot leg to repeat the heat removal process.

C.lncorrect 1) False. A lower cooldown rate is plausible to the Operator since EOP-6 cooldown rate limits are lower when on natural circulation (35F/hr when on natural circulation.) During a natural circulation cooldown in EOP-5, the allowable cooldown rate is the same for both natural and forced circulation (<100F/hr.) 2) False. S/Gs not available as a heat sink is plausible to the Operator if they believe boiling in the core is the only heat transfer mechanism occurring during reflux boiling. During reflux boiling, RCS level will be below the top of the hot leg to allow the reflux process to occur. Steam leaving the core is condensed in the SG U-Tubes and that condensate than travels back to the core via the hot leg to repeat the heat removal process.

D. Incorrect 1) False. A lower cooldown rate is plausible to the Operator since EOP-6 cooldown rate limits are lower when on natural circulation (35F/hr when on natural circulation.) During a natural circulation cooldown in EOP-5, the allowable cooldown rate is the same for both natural and forced circulation (<100F/hr.) 2) True. During reflux boiling, RCS level will be below the top of the hot leg to allow the reflux process to occur. Steam leaving the core is condensed in the SG U-Tubes and that condensate than travels back to the core via the hot leg to repeat the heat removal process.

Page: 9 of 235 13 April 2016




Topic: Natural Gire and Reflux implications during a SBLOCA RO Importance: 4.2 SRO Importance: 4.7 Cognitive level (High or

L Low): NRC KIA Info, References:

Tier/Group 1/1

KIA Info EPE 009-Small Break LOCA EK1 Knowledge of the operational implications of the following concepts as they apply to the small break LOCA EK1 .01 Natural circulation and cooling, including reflux boiling

RO 4.2 Importance

Technical EOP-5, EOP-5-TB References


Learning Identify the core heat removal mechanisms Objective for a large or small break LOCA

Question New Source


Cognitive Memory or Fundamental Knowledge Level


Comments None

OPERATIONS Page: 1 O of 235 13 April 2016


Task 201.062 Control Core and RCS voiding


4


ID: Q106476 Points: 1~00

Seal degradation has been diagnosed on only 128 RCP.

Which ONE of the following conditions require the tripping of 128 RCP?

A Lower seal temperature >280°F.

8. C80 temperature >200°F.

C. C80 flow reads 0 GPM.

D. RCP C80 PRESS HI alarm.

Answer: 8

Answer Explanation:

A Incorrect-A lower seal temperature of >280F for >8 hours is the criteria for rebuilding a seal prior to restarting the pump. Operator could choose this if confusing rebuild criteria with trip criteria.

8. Correct-AOP-7C and 1C06 ARM both list C80 temperature >200F as a critical parameter to stop the pump.

C. Incorrect-A C80 flow of 0 GPM is a potential indicator that either C80 flow has been isolated or the vapor seal has failed. Per Ol-1A, C80 flow may be only used to confirm seal failure as indicated by other parameters, such as seal d/p and C80 temperature. The indication, by itself, is not a reason to trip the pump. Operator could choose this if they believe C80 has been isolated to all RCPs (1C07 ARM guidance to trip RCPS if C80 isolated and all middle and upper stage RCP pressures alarm high.)

D. lncorrect-RCP C80 PRESS HI alarm indicates that normal flowpath for C80 has been isolated. The alarm does not indicate that C80 flow is completely isolated as 1-CVC-199-RV will lift at a slightly higher pressure and provide a C80 flowpath to the RCDT vice the VCT. The indication is expected during a SIAS, but is not a reason to secure the pump. Operator could choose this if they believe C80 has been isolated to all RCPs (1C07 ARM guidance to trip RCPS if C80 isolated and all middle and upper stage RCP pressures alarm high.)

OPERATIONS Page: 12 of 235 13April2016


Question 4 Info Question Type: Multiple Choice Status: Active Always select on test? Yes Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 0.00

Svstem ID: 106476 User-Defined ID: Q106476 Cross Reference Number:

Topic: RCP Trip Criteria on CBO conditions RO Importance: 2.9 SRO Importance: 2.9 Cognitive level (High or

L Low): NRG KIA Info, References:

Tier/Group 1/1

KIA Info APE 015/017-Reactor Coolant Pump (RCP) Malfunctions AK2. Knowledge of the interrelations between the Reactor Coolant Pump Malfunctions (Loss of RC Flow) and the following: AK2.07 RCP seals

RO 2.9 Importance

Technical AOP-7C, 1 C06 ALM, Ol-1A References


Learning Given a set of plant conditions determine Objective when conditions would require a manual

reactor coolant pump trip

Question New Source






Comments KIA is considered a match since seal degradation led to the loss of RC flow when the pump was secured.

Task 064.002 Stop a Reactor Coolant Pump


5


ID: Q106477 Points:+t.o:o.

Unit-2 is operating at 100% power when the following transient occurs:

• A suction header ruptures on the Charging Pumps • AOP-28 is implemented • All Charging Pumps are secured • Letdown is isolated

Which ONE of the following is a reason letdown is immediately isolated after securing charging?

A To limit VCT level and prevent overfill of reference leg.

B. To prevent VCT pressure from reaching relief valve setpoint.

C. To minimize thermal transients on the Regenerative Heat Exchanger.

D. To prevent low Component Cooling Water header pressure alarms as Letdown HX TIC responds.

Answer: C

Answer Explanation:

A Incorrect-Limiting VCT level is plausible to the Operator since VCT will rise when charging is secured. However, any high VCT level should be diverted to the Waste System well before the VCT level rises to the reference leg tap location.

B. Incorrect-Preventing VCT pressure is plausible to the Operator since a rising VCT level will increase pressure. However, any high VCT level should be diverted to the Waste System well before the VCT pressure reaches an alarm or relief setpoint.

C. Correct-When charging is secured, that flow is no longer supplying the heat sink for the RCS flowing from the letdown nozzle and through the RHX. This will result in the temperature of the RHX and the letdown piping downstream to begin to rise. Eventually the temperature downstream of the RHX will reach approximately 470F and letdown will automatically isolate. The manual isolation of letdown limits the temperature rise and therefore, the thermal transients on the letdown system.

D. Incorrect-Preventing low CCW pressure alarms is plausible to the Operator since a rising letdown temperature will require a higher CCW flow to lower the temperature. However, the CCW flowrate through the NRXH does not change appreciably before letdown automatically isolates, and CCW pressures are therefore not challenged.





Topic: Why is Letdown Isolated when securing all charging RO Importance: 3.2 SRO Importance: 3.4 Cognitive level (High or


Tier/Group 1/1

KIA Info APE 022-Loss of Reactor Coolant Makeup AK3. Knowledge of the reasons for the following responses as they apply to the Loss of Reactor Coolant Makeup: AK3.04 Isolating letdown

RO 3.2 Importance

Technical AOP-28 References


Learning Recall the impact on CVCS letdown flow Objective

Question New Source



10CFR 55.41(b)(5/10) Part 55 Content

Comments None

OPERATIONS Page 16 of 235 13 April 2016

6


I Task 1202.135 Respond to a loss of all charging Modes 1 & 2

ID: Q10~478

Unit-1 is in Mode 6 with the following initial conditions:

• Both the north and south Refueling Pool cavities are filled • RCS level is at 43.5' • The reactor head is installed and detensioned

A common mode loss of SOC occurs and the loss of SOC will be long-term.

Which ONE of the following choices is the 1) optimal short term heat removal strategy and 2) the optimal long term heat removal strategy?

A. 1) Lineup a High Pressure Safety Injection (HPSI) Pump and initiate OnceThrough-Cooling flow. 2) Shift HPSI pump suction to the Containment sump and continue OnceThrough-Cooling flow.

B. 1) Cool the RFP using a SFP Pump/Cooler. 2) Maintain cooling to the RFP using a SFP Pump/Cooler.

C. 1) Lineup a Containment Spray Pump and fill the RCS. 2) Allow RCS heat up and cool with the SGs.

0. 1) Gravity fill from RWT. 2) Gravity fill from opposite unit RWT.

Answer: A

I


OPERATIONS


Answer Explanation:

A Correct-1) True. With a common mode loss of SOC and the RFP not available as a heat sink (due to head in place but not removed) and the RCS not intact (head is detensioned and PZR manway would have been removed as well to support this function), AOP-38 directs that a HPSI, LPSI, or CS pump be utilized to fill the RCS as boil off occurs, Once-Through-Cooling. 2) True. Suction for these pumps would be from the RWT. As RWT lowers, and goes below 2.5', AOP-38 directs the suction of the pumps be aligned to the containment sump versus the RWT.

8. lncorrect-1) False. Cooling the RFP with the SFP system is plausible since this is a heat sink method when the RFP is available. Although the RFP cavities are both full, none of this water is able to communicate with the reactor vessel due to the head being installed. Even if the RFP were to be filled, the RFP inventory would not assist with the heat removal of the core. 2) False. Cooling the RFP with the SFP system is plausible since this is a heat sink method when the RFP is available. Although the RFP cavities are both full, none of this water is able to communicate with the reactor vessel due to the head being installed. Even if the RFP were to be filled, the RFP inventory would not assist with the heat removal of the core.

C. lncorrect-1) True. With a common mode loss of SOC and the RFP not available as a heat sink (due to head in place but not removed) and the RCS not intact (head is detensioned and PZR manway would have been removed as well to support this function), AOP-38 directs that a HPSI, LPSI, or CS pump be utilized to fill the RCS as boil off occurs, Once-Through-Cooling. Suction for these pumps would be from the RWT. 2) False. Use of the S/Gs as a heat sink is plausible to the Operator since the S/Gs are used when the RCS can be pressurized. However, even though the reactor head is installed, the RCS can not be pressurized since the head is detensioned. The SGs would not be available as a heat sink since the RCS is not intact (with the head detensioned).

0. lncorrect-1) False. Gravity fill from the RWT is plausible to the Operator since it is a heat removal strategy available in AOP-38. However, the strategy is utilized only when there is a loss of both vital 4kv busses (and their respective HPSI, LPSI, and CS pumps.). 2) False. Using the opposite unit RWT is plausible to the Operator since there are 01-24 alignments that allow transferring RWT water. However, AOP-38 does not provide guidance on using the opposite units RWT as an inventory source.

Page: 18 of 235 13 April 2016




Topic: SOC leak requiring Containment Sump suction RO Importance: 2.6 SRO Importance: 2.6 Cognitive level (High or

H Low): NRC KIA Info, References:

Tier/Group 1 /1

KIA Info APE: 025 Loss of Residual Heat Removal System (RHRS) AK2. Knowledge of the interrelations between the Loss of Residual Heat Removal System and the following: AK2.05 Reactor building sump

RO 2.6 Importance



Learning Given plant conditions resulting in a loss of Objective shutdown cooling, direct or implement the

actions to maintain plant parameters within desired limits IAW with AOP-38

Question New Source





7


Comments None

Task 202.022 Respond to a complete loss of SOC with RCS open and the refueling pool not available

Points: 1.00

Unit-1 is operating at 100% power with the following initial conditions:

11 Component Cooling Water (CCW) pump is in operation

The following transient occurs:

• A CCW leak develops • Component Cooling Head tank level is 5 inches and slowly lowering • Component Cooling Standby header pressure is 75 PSIG and slowly lowering • 1C10 alarm "CNTMT NORMAL SUMP LVL HI" is in • 1C07B alarm "12A RCP CCW FLOW LO" is in • 12A Thrust Bearing and CBO temperatures are rising

Which ONE of the following actions is required based on existing plant conditions?

A. Shut the Component Cooling CVs to Containment, trip the reactor, and secure the RCPs.

B. Secure the operating Component Cooling pump due to cavitation caused by low head tank level.

C. Restore Component Cooling Head tank level by opening the bypass to the makeup CV.

D. Trip the Reactor since no Component Cooling pumps are operating.

Answer: A


OPERATIONS


Answer Explanation:

A Correct-Based on the RCP information given, it can be deduced that a CCW leak exists on the 12A RCP. Isolating the leakage by shutting the containment isolation valves requires tripping the reactor then the RCPs once reactivity control safety function is verified. AOP-7C Step V.C.5 states identify leak location, training is that when identified the leak will be isolated.

B. Incorrect-Plausible since a low head tank could cause pump cavitation if level continued to lower after the head tank was empty. Based on the standby header pressure, there is adequate pump pressure. Standby and normal CC header pressures are the same when the CCW system is in a normal alignment and does not depend on 11 or 12 CCW pump in operation. With a pressure of 60 PSIG, there is a CCW pump operating.

C. Incorrect-Plausible since this is a potential action taken for a low head tank level. Since a CCW System leak exists, priority would be placed on isolating the leak. Once the leak is isolated, bypassing the make-up CV would only occur if the normal make-up CV was not open. No malfunction of the makeup CV was given in the stem of the question.

D. Incorrect-Plausible if Operator confuses 12A RCP CCW FLOW LO to indicate that all CCW pumps are secured. Based on the standby header pressure, there is adequate pump pressure. Standby and normal CCW header pressures are the same when the CCW system is in a normal alignment and does not depend on 11 or 12 CCW pump in operation. With a pressure of 75 PSIG, there is a CCW pump operating.

Based on the simulator, the CCW pumps will not show signs of cavitation when head tank level is visible. This assumes that the pump is not in a runout condition, and a standby pressure of -75 PSIG does not indicate runout conditions.

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Topic: Actions to be taken for a CCW water leak at power on Unit-1

RO Importance: 2.913.3 SRO Importance: 3.0 / 3.1 Cognitive level (High or

High Low): NRC KIA Info, References: Tier/Group 1/1

KIA Info APE026 Loss of Component Cooling Water (CCW) AA 1 Ability to operate and I or monitor the following as they apply to the Loss of Component Cooling Water: AA 1.07 Flow rates to the components and systems that are serviced by the CCWS; interactions among the components.

RO 2.9 Importance

Technical AOP-7C, Loss of Component Cooling Water References


Learning Given plant operating conditions and a loss Objective of Component Cooling Water, evaluate and

determine the required actions

Question Bank Source

Question 2014 NRC RO Exam History



8


10CFR 55.41 (b )(7) Part 55 Content

Comments None

Task 202.067 Respond to a loss of Component Cooling in Modes 1 and 2

Additional KIA Match Info 008 Component Cooling Water System A4 Ability to manually operate and/or monitor in the control room: A4.01 CCW indications and controls

Question 7 Table-Item Links

A I 0 Training Program

Licensed Operator Requa! Training (LOR)

System Designations

Abnormal Operating Procedures (AOPs)

:; , ... eoin~:· 1.00

Unit-1 is at 100% with the following initial conditions:

• Unit-1 is at 100% power • Pressurizer Pressure Controller PIC-100X is the selected Control Channel and is in AUTO • Pressurizer Spray Valve Controller HIC-100 is in Auto • RCS Pressure is 2250 PSIA

Which ONE of the following is the IMMEDIATE plant response if Pressurizer Pressure Controller PIC-1 OOX Controller setpoint fails to 1500 PSIA?

A. Pressurizer Spray Valve Controller HIC-100 output goes to maximum and proportional heaters output goes to maximum.

B. Pressurizer Spray Valve Controller HIC-100 output goes to minimum and proportional heaters output goes to minimum.

C. Pressurizer Spray Valve Controller HIC-100 output goes to minimum and proportional heaters output goes to maximum.

D. Pressurizer Spray Valve Controller HIC-100 output goes to maximum and proportional heaters output goes to minimum.

Answer: D


OPERATIONS


Answer Explanation:

A. Incorrect-When PIC-100X fails low, this will cause the Spray Valve controller to go to maximum output as the actual RCS pressure of 2250 will be above setpoint of 1500. The Proportional heater output would go to minimal as well due to the mismatch. Heater output going to maximum is plausible if the Operator believes the Proportional Heaters act like the Backup heaters and the Proportional Heaters will also energize on an insurge. The insurge would be created by maximum spray collapsing the Pressurizer bubble causing Pressurizer level to rise.

B. Incorrect-Spray Valve controller going to minimal output is plausible if the Operator believes the Spray Valve controller output will go in the same direction as the Proportional Heater output. When PIC-100X fails low, this will cause the Spray Valve controller to go to maximum output as the actual RCS pressure of 2250 will be above setpoint of 1500. The Proportional heater output would go to minimal as well due to the mismatch.

C. Incorrect-Spray Valve controller going to minimal output is plausible if the Operator believes the Spray Valve controller output will go in the same direction as the Proportional Heater output actually responds. When PIC-1 OOX fails low, this will cause the Spray Valve controller to go to maximum output as the actual RCS pressure of 2250 will be above setpoint of 1500. The Proportional heater output would go to minimal as well due to the mismatch. Heater output going to maximum is plausible if the Operator believes the Proportional Heaters act like the Backup heaters and the Proportional Heaters will also energize on an insurge. The insurge would be created by maximum spray collapsing the Pressurizer bubble causing Pressurizer level to rise.

D. Correct-When PIC-100X fails low, this will cause the Spray Valve controller to go to maximum output as the actual RCS pressure of 2250 will be above setpoint of 1500. The Proportional heater output would go to minimal as well due to the mismatch.

Page: 24 of 235 13 April 2016




.. ·

Topic: Selected PZR pressure controller setpoint fails to 1500 PSIA

RO Importance: 2.6/3.8 SRO Importance: 2.9/3.9 Cognitive level (High or


Tier/Group 1 /1

KIA Info APE 027-Pressurizer Pressure Control System (PZR PCS) Malfunction AK2. Knowledge of the interrelations between the Pressurizer Pressure Control Malfunctions and the following: AK2.03 Controllers and positioners

RO 2.6 Importance

Technical System Description 064D, RCS References Instrumentation

ALM-1 C06, RCS Control References None provided

Learning Given a failure of any RCS pressure, Objective temperature or level instrument predict the

response on the system (heaters, main spray, charging, and letdown) to that failure







Comments None

Task 064.038 Verify main pressurizer spray is operatinq

Additional K/A Match Info 010 Pressurizer Pressure Control System (PZR PCS) K3 Knowledge of the effect that a loss or malfunction of the PZR PCS will have on the following: K3.01 RCS


9


ID: Q106493 Points: 1.00

Which ONE of the following Nuclear Instrument (NI) indications would indicate that an ATWS condition exists?

A WRNI power is 9x10-2 % and SUR is 2 DPM

B. VOPT was reset at 60% and NI power is 70%.

C. RRS NI power is 106%

D. NI power is 103%

Answer: B

Answer Explanation:

A lncorrect-2 DPM is plausible to the Operator since the POWER LVL RATE HI CH PRETRIP alarm comes in a 1.34 DPM. However, the SUR trip setpoint is 2.43 (2.27-2.59) DPM.

B. Correct-When VOPT is reset by the operator, the trip setpoint will rise 8.4% above current power. When reset at 60%, VOPT will reset to 68.4%. Any power over 68.4% should trip the reactor if not reset by the operator beforehand.

C. lncorrect-106% is plausible to the Operator since this value is high enough to actuate RPS (power above trip value of 105.4%.) However, RRS NI output is not fed into RPS.

D. Incorrect-The HI POWER TRIP RESET DEMAND setpoint is 2.6% less than the Reactor trip setpoint. A power of 103% is plausible to the Operator if they add the 2.6% to 100% and determine that 103% has exceeded this calculated value (102.6%.) This power.is not high enough to actuate RPS (since needs power above trip value of 105.4%.)


EXAMINATION ANSWER KEY LOI 2015 NRG RO Exam (June 2016}



Topic: ATWS conditions based on NI readinqs RO Importance: 4.4 SRO Importance: 4.7 Cognitive level (High or H Low): NRG KIA Info, References:

Tier/Group 1/1

KIA Info EPE 029 Anticipated Transient Without Scram (ATWS) EA2 Ability to determine or interpret the following as they apply to a ATWS: EA2.01 Reactor nuclear instrumentation

RO 4.4 Importance

Technical TS 3.3.1 References 1 COS Alarm Manual D-05


Learning Recognize the conditions which would Objective constitute an Anticipated Transient Without

Scram (A TWS)

Question New Source




Comments None


OPERATIONS


Task 201.010 Determine if the reactor should have tri ed automaticall but did not ATWS)

Page: 29 of 235 13 April 2016

10


JD: Q106494 . Points: 1.00

Unit-2 is operating with the following initial conditions:

• EOP-6 is in progress • RCS pressure is being reduced to 900 PSIA • Cooldown in progress

Which ONE of the following is the HIGHEST RCS temperature that still ensures 25°F subcooling after reaching 900 PSIA?

A 557°F

B. 532°F

C. 507°F

D. 497°F

Answer: C

Answer Explanation:

A Incorrect-At 900 PSIA, saturation temperature is 532F. To achieve 25F SCM, temperature must be 532-25=507F. Any temperature higher than 507F will result in a lower SCM. 557F is plausible to the Operator if 25F was added to the saturation temperature rather than subtracted.

B. Incorrect-At 900 PSIA, saturation temperature is 532F. To achieve 25F SCM, temperature must be 532-25=507F. Any temperature higher than 507F will result in a lower SCM. 532F is plausible to the Operator if they forgot to subtract the desired SCM value from the saturation temperature.

C. Correct-At 900 PSIA, saturation temperature is 532F. To achieve 25F SCM, temperature must be 532-25=507F.

D. Incorrect-At 900 PSIA, saturation temperature is 532F. To achieve 25F SCM, temperature must be 532-25=507F. Any temperature lower than 507F would exceed 25F SCM. The question asks for the highest temperature can be, with the correct answer of 507F being higher than 497F. 497F is plausible to the Operator if the cooldown rate limit of 35 (F/hr) was subtracted from saturation temperature rather than just 25F.





Topic: Use of steam tables in EOP-6 RO Importance: 3.1 SRO Importance: 3.4 Cognitive level (High or


Tier/Group 1/1

KIA Info EPE 038 Steam Generator Tube Rupture (SGTR) EK1 Knowledge of the operational implications of the following concepts as they apply to the SGTR: EK1 .01 Use of steam tables

RO 3.1 Importance

Technical EOP-6 References Steam Tables


Learning Recall the conditions to and bases for Objective maintaining RCS subcooling as close to 25°F

as possible while on natural circulation and as close to the NPSH limits with RCP's running

Question New Source






Comments None

Task 201.133 Depressurize the RCS to less than 950 PSIA prior to the affected SIG level increasing above 63"


11


ID: Q106495 ··. Points: 1.00

Unit-2 is at 2% with the following initial conditions:

• Startup in progress from a Refueling Outage


• The running SGFP trips • AOP-3G is implemented • Neither SGFP can be reset

Which ONE of the following is the reason power is reduced to <1 %?

A. To lower steam flow which allows AFW flowrate to maintain SG levels.

B. To lower SG pressure which allows Condensate Booster Pump injection.

C. To raise SG pressure which ensures AFW steam driven turbines have adequate pressure.

D. To lower power below the Point of Adding Heat which limits power changes resulting from temperature changes.

Answer: A

Answer Explanation:

A. Correct-Per the AOP-3G basis, power is lowered to <1% on a loss of both SGFPs to ensure feedwater flow from the AFW system can match a lower steam flow and maintain SG levels.

B. Incorrect-Condensate Booster Pump (CBP) injection is plausible to the Operator since CBP injection is a method utilized only in EOP-3 or EOP-8. SG pressure could lower based on TBV output at beginning of event. However, even if temperature/SG pressure were to lower by lowering power, CBP injection is not an AOP-3G procedurally driven method of restoring feed.

D. Incorrect-Pressure increasing is plausible to the Operator since SG pressure usually increases as steam flow is lowered. However, the act of increasing SG pressure is not necessary to ensure the AFW turbines will operate correctly to inject to the SGs. Only 65 PSIA is required in the SGs to successfully operate.

D. Incorrect-Lowering power to prevent power changes caused by temperature changes (i.e. loss of feed causing power to rise since temperature will rise) is plausible to the Operator since power control with a +MTC is more challenging. However, per the AOP-3G basis, power is lowered to <1 % on a loss of both SGFPs to ensure feedwater flow from the AFW system can match a lower steam flow and maintain SG levels.





Topic: Why steam and feed flow are matched during LOFW at lower power



Tier/Group 1/1

KIA Info APE 054 Loss of Main Feedwater (MFW) AK3. Knowledge of the reasons for the following responses as they apply to the Loss of Main Feedwater (MFW): AK3.02 Matching of feedwater and steam flows

RO 3.4 Importance

Technical AOP-3G References


Learning Given plant conditions associated with Objective various Feedwater malfunctions, the license

operator candidate will be able to correctly recall and/or identify the proper AOP-3G response and bases to successfully mitigate the effects of Feedwater malfunctions

Question New Source






Comments None

Task 202.037 Respond to loss of Condensate or SGFP at < 5% oower


12


ID: Q106496 Points: 1.00

Unit-1 is at 100% power when the following transient occurs:

• Loss of Offsite Power • 1 A DG failed to start • 1 B DG was tagged out • OC DG failed to start • Neither 11 or 14 4Kv busses are faulted • EOP-7 is implemented for a Station Blackout (SBO) • Offsite power and SMECO are still unavailable

60 minutes into the SBO, the OC DG issue has been repaired.

Which ONE of the following EOP-7 actions would be required to restore power to a Unit-1 vital 4Kv bus?

A. Prelube not necessary; Emergency start OC DG from Control Room.

B. Prelube not necessary; Depress OC Manual Emergency Start pushbutton locally.

C. Prelube OC DG for 5 minutes first; then Emergency start OC DG from Control Room.

D. Prelube OC DG for 20 minutes first, then depress OC Manual Emergency Start pushbutton locally.

Answer: D


OPERATIONS


Answer Explanation:

A. Incorrect-Emergency start of the OC DG without a prelube is plausible to the Operator since these are the actions in EOP-0. However, when restoring power with the OC DG in EOP-7, an emergency start is directed per Ol-21C using the OC Manual Emergency Start pushbutton locally. With power gone to OC DG auxiliaries for more than 40 minutes, 01-21 C requires a prelube of at least 20 minutes before the DG start. Operator could choose this if they believe a prelube could be omitted during an emergency. Ol-21C only allows the prelube to be omitted in an emergency where power was lost to the AC prelube pumps for < 30 minutes.

8. Incorrect-Starting the DG using the OC Manual Emergency Start pushbutton locally is plausible to the Operator since this is a required action in Ol-21C. However, with power gone to OC DG auxiliaries for more than 40 minutes, Ol-21C requires a prelube of at least 20 minutes before the DG start. Operator could choose this if they believe a prelube could be omitted during an emergency. 01-21 Conly allows the prelube to be omitted in an emergency where power was lost to the AC prelube pumps for< 30 minutes.

C. Incorrect-Emergency start of the OC DG is plausible to the Operator since these are the actions in EOP-0. With power gone to OC DG auxiliaries for more than 40 minutes, 01-21 C requires a prelube of at least 20 minutes before the DG start. Operator could choose a 5-minute prelube since this is a prelube time in Ol-21C, but only if prelube has been lost for <40 minutes.

D. Correct-When restoring power with the OC DG in EOP-7, an emergency start is directed per Ol-21C using the OC Manual Emergency Start pushbutton locally. With power gone to OC DG auxiliaries for more than 40 minutes, Ol-21C requires a prelube of at least 20 minutes before the DG start.

Page 37 of 235 13April2016




Topic: OC DG actions to restore power durinq a SBO RO Importance: 3.9 SRO Importance: 4.7 Cognitive level (High or


Tier/Group 1/1

KIA Info EPE 055-Loss of Offsite and Onsite Power (Station Blackout) EA2 Ability to determine or interpret the following as they apply to a Station Blackout: EA2.03 Actions necessary to restore power

RO 3.9 Importance

Technical EOP-7 References Ol-21C


Learning Given a SBO, identify, in order of priority, the Objective power sources available to reenergize at

least one 4KV bus

Question New Source




Comments None



'Task 1201.075 Align electrical system for power restoration I


13


ID: Q106497 Points,::1.00

Unit-2 is operating at 100% power when 2Y10 is lost.

Which ONE of the following are the immediate actions expected per the AOP?

A. Open VCT outlet MOV-501; Shut RWT to Charging MOV-504; Adjust Turbine to maintain Tcold on program; Place 2 Charging Pumps in PTL

B. Fast borate to reduce Reactor power; Promptly reduce Turbine load

C. Shift PZR Level, Pressure, and RRS controls to Channel X

D. Shift PZR Level, Pressure, and RRS controls to Channel Y; Verify Letdown HIC-110 is in Manual.

Answer: A

Answer Explanation:

A. Correct-A loss of 2Y10 results in Charging Pump suction shifting to the RWT with resultant boration of the RCS. Stabilizing actions as directed by the "Immediate Actions from 100% Power" placard direct the securing of boration (opening MOV-501, shutting MOV-504), adjusting Turbine load to maintain TCOLD on program, and securing 2 Charging Pumps at first due to the loss of letdown to prevent exceeding PZR TS limit of 225"-

B. Incorrect-Plausible since these are the actions directed by the "Immediate Actions from 100% Power" placard for loss of 2Y09.

C. Incorrect-Plausible since these are the actions directed by the "Immediate Actions from 100% Power" placard for loss of 2Y01.

D. Incorrect-Plausible since these are the actions directed by the "Immediate Actions from 100% Power" placard for loss of 2Y02.



Topic: Immediate actions for a loss of 2Y10 RO Importance: 4.6 SRO Importance: 4.4 Cognitive level (High or




Tier/Group 1/1

K/A Info APE 057 Loss of Vital AC Electrical Instrument Bus 2.4 Emergency Procedures/Plan 2.4.49 Ability to perform without reference to procedures those actions that require immediate operation of system components and controls

RO 4.6 Importance

Technical AOP-71-2 References AOP-7J-2


Learning Given an electrical bus malfunction, diagnose Objective the event and take the appropriate actions

per AOP-71

Question Modified Source




Comments Modified from L92859. L92859 used in NRC 2008 RO Exam.

Task 202.095 Respond to the Loss of a 208/120 volt Instrument Bus


14


ID: Q106498 P<.>ints: 1.00

Unit-1 is at 100% when the following transient occurs:

• 21 DC bus is lost • AOP-7J is implemented

Which ONE of the following prompt actions are 1) required by the AOP and 2) what Reactor Coolant Pump (RCP) actions should be taken prior to exiting EOP-0?

A. 1) Trip Reactor from 1 C05 (and the Turbine automatically trips as expected). 2) Trip 2 RCPs from 1 C06 and trip 2 RCPs locally at RCP breakers.

B. 1) Trip Reactor from 1 C05 (and the Turbine automatically trips as expected). 2) Trip 4 RCPs from 1C19.

C. 1) Trip Reactor from 1 C05 and then manually trip Turbine from 1 C02. 2) Trip 2 RCPs from 1 C06 and trip 2 RCPs locally at RCP breakers.

D. 1) Trip Reactor from 1 C05 and then manually trip Turbine from 1 C02 2) Trip 4 RCPs from 1C19

Answer: D


OPERATIONS


Answer Explanation:

A. Incorrect 1) False. The Turbine automatically tripping is plausible to the Operator since this is the normal response on a reactor trip. However, the turbine will not automatically trip due to the loss of 1Y02 (downstream of 21 DC bus) and personnel are stationed at 1C05 and 1 C02 to coordinate the tripping of the reactor and then the turbine. 2) False. Tripping 2 RCPs locally and 2 from the Control Room is plausible to the Operator since these are actions for a loss of 21 DC bus on U-2. However, since 2 RCPs are untrippable from 1 C06, all 4 RCPs are secured by opening the RCP bus feeder breakers at 1C19.

B. Incorrect 1) False. The Turbine automatically tripping is plausible to the Operator since this is the normal response on a reactor trip. However, the turbine will not automatically trip due to the loss of 1Y02 (downstream of 21 DC bus) and personnel are stationed at 1C05 and 1 C02 to coordinate the tripping of the reactor and then the turbine. 2) True. Once reactivity is completed, AOP-7 J directs that the RCPs be secured. Since 2 RCPs are untrippable from 1C06, all 4 RCPs are secured by opening the RCP bus feeder breakers at 1C19.

C. Incorrect 1) True. On loss of 21 DC bus, personnel are stationed at 1 C05 and 1 C02 to coordinate the tripping of the reactor and then the turbine. 2) False. Tripping 2 RCPs locally and 2 from the Control Room is plausible to the Operator since these are actions for a loss of 21 DC bus on U-2. However, since 2 RCPs are untrippable from 1 C06, all 4 RCPs are secured by opening the RCP bus feeder breakers at 1C19.

D. Correct 1) True. On loss of 21 DC bus, personnel are stationed at 1 C05 and 1 C02 to coordinate the tripping of the reactor and then the turbine. 2) True. Once reactivity is completed, AOP-7 J directs that the RCPs be secured. Since 2 RCPs are untrippable from 1C06, all 4 RCPs are secured by opening the RCP bus feeder breakers at 1C19.

Page 43 of 235 13 April 2016




Topic: Prompt actions for loss of 21 DC bus on U-1 RO Importance: 4.6 SRO Importance: 4.4 Cognitive level (High or


Tier/Group 1/1

KIA Info APE 058 Loss of DC Power 2.4 Emergency Procedures/Plan 2.4.49 Ability to perform without reference to procedures those actions that require immediate operation of system components and controls

RO 4.6 Importance

Technical AOP-7J References


Learning Given a loss of any 125 Volt DC bus, Objective implement the required actions to mitigate

the transient and restore power to affected equipment per AOP-7J

Question New Source






Comments None

Task 202.101 Respond to the loss of a 125 volt DC bus


15


ID:Q106499 Points: 1.00

Unit-1 is operating at 100% power with the following initial conditions:

• 11 SFP HX in service and cooling the SFP • 1A DG running at 4.0 MWe


• 11 Service Water Pump trips

Which ONE of the following Control Room indications are expected first?

A. 1A DG alarm on 1C18A.

B. Main Generator H2 temperatures rising on 1 C02

C. SFP Cooler Discharge Temperature High alarm on 1C13.

D. Main Turbine Lube Oil temperatures rising on the Process Plant Computer.

Answer: D

Answer Explanation:

A. lncorrect-1A DG alarm is plausible since SRW supplies cooling to the Fairbanks DGs. However, 11 SRW header does not supply SRW to the 1A DG.

B. lncorrect-H2 temperatures rising are plausible to the Operator since SRW cools the Generator Cooling system. However, a trip of 11 SRW pump and loss of 11 SRW header will affect cooling to EHC, IA, PA, and Turbine Lube Oil. The Generator Hydrogen coolers are supplied by 12 SRW header and would be unaffected by a loss of 11 SRW.

C. lncorrect-SFP temperatures rising is plausible since 11 SFP HX is supplied by SRW. However, the 12 SRW header supplies the 11 SFP HX, not 11 SRW header. 11 SFP HX would be unaffected by the loss of 11 SRW header.

D. Correct-For a trip of 11 SRW pump and loss of 11 SRW header, EHC, IA, PA, and Turbine Lube Oil will lose cooling. As Turbine Lube Oil warms, temperature indications on the Plant Computer will begin to rise until 11 SRW header is restored.





.

Topic: Indications in Control Room of a 11 SRW Header loss RO Importance: 3.2 SRO Importance: 3.3 Cognitive level (High or


Tier/Group 1/1

KIA Info APE 062 Loss of Nuclear Service Water AA 1 Ability to operate and I or monitor the following as they apply to the Loss of Nuclear Service Water (SWS): AA 1.02 Loads on the SWS in the control room

RO 3.2 Importance



Learning Differentiate between a loss of Service water Objective flow and a Service Water leak (rupture)

Question New Source




Comments None


OPERATIONS


!Task 1

202.065 Respond to a loss of Service Water I

in modes 1 or 2

Page: 48 of 235 13 April 2016

16


Points: 1.00

Unit-1 was operating at 100% power when a reduction in instrument air header pressure occurred. Given the following events and conditions:

• Instrument Air header pressure lowered to 87 PSIG • Plant Air header pressure lowered to 84 PSIG

Which ONE of the following conditions should have occurred?

1. The standby instrument air compressor started 2. CNTMT IA SUPPLY CV, 1-IA-2085-CV, shuts 3. Plant air to instrument air cross connect valve (PA-2061-CV) opened 4. Plant air header automatic isolation valve (PA-2059-CV) closed

A. Actions 1 and 4 ONLY.

B. Actions 2 and 3 ONLY.

C. Actions 1, 3, and 4 ONLY.

D. Actions 2, 3 and 4 ONLY.

Answer: C

Answer Explanation:

A. Incorrect-Action 1 occurred at 93 PSIG IA pressure and Action 4 occurred at 88 PSIG IA pressure, however, Action 3 also occurs. Plausible if Operator believes Action 3 occurs at lower pressure of 75 PSIG, when IA-2085 shuts.

B. Incorrect-Action 3 has occurred but Action 2 has not. Action 2 occurs at 75 PSIG IA pressure, and Action 3 occurred at 85 PSIG PA pressure. Plausible if Operator believes Action 2 occurs at higher pressure of 85 PSIG, when PA-201-CV opens.

C. Correct-Action 1 occurred at 93 PSIG IA pressure, Action 3 occurred at 85 PSIG PA pressure, and Action 4 occurred at 88 PSIG IA pressure.

D. Incorrect-Action 2 has not occurred based on IA pressure value. Actions 1, 3, and 4 have occurred based on IA and PA pressure. Plausible if Operator believes Action 1 has not occured due to IA lowering and believes Action 2 occurs at higher pressure of 85 PSIG, when PA-201-CV opens.


EXAMINATl,ON ANSWER KEY LOI 2015 NRC RO Exam (June 2016)



Topic: Lowering IA pressure effects on system operation RO Importance: 3.5 SRO Importance: 3.4 Cognitive level (High or

Low Low): Comments:

Tier/Group 1/1

KIA Info APE 065 Loss of Instrument Air AA 1 Ability to operate and I or monitor the following as they apply to the Loss of Instrument Air: AA 1. 04 Emergency Air Compressor

RO 3.5 Importance

Technical AOP-7D-1, Loss of Instrument Air References


Learning Recall the automatic functions of the Objective following components, and their functional

relationship to a loss of instrument air: • Plant Air to IA CV • PA to PA CV • CV-2085 • Standby Instrument Air Compressor


Question NRC 2012 RO Exam History





Comments Considered a KIA match since the Plant Air Compressors are powered from an Emergency DG powered Vital 4Kv bus and are a backup to the Instrument Air Compressors

Task 202.069 Respond to a Loss of Instrument Air in Modes 1 & 2



License Operator Initial Training (LOIT)


17


lD: Q106500 Points: 1.00

Unit-2 was operating at 100% with the following initial conditions:

• Main Generator has a reactive load of 200 MVARs in the LAG direction • Voltage Regulator is in MANUAL

The SO-TSO requests that you raise switchyard 500 Kv line voltage by 10 Kv

Which ONE of the following is a concern when raising the switchyard voltage by 10 Kv?

A. There will be less margin to the Underexcitation Reactive Amp Limit AND Unit-2 13 Kv bus voltages will rise.

B. There will be less margin to the Reactive Capability Curve AND Unit-2 120V Vital AC bus voltages will rise. ·

C. There will be an increase in Generator heating AND Unit-2 4Kv bus voltages will rise.

D. There will be an increase in grid frequency AND Unit-2 480V bus voltages will rise.

Answer: C

Answer Explanation:

A. Incorrect-With MVARs in the LAG direction, the generator is overexcited. The URAL applies to underexcited conditions only. Less margin is plausible if the Operator believes LAG is synonymous with underexcited.

B. Incorrect-Margin to the Reactive Capability Curve will decrease as MVARs are increased to raise switchyard voltage. 120V Vital AC voltage rising is plausible if Operator believes all downstream busses will rise when switchyard voltage is increased. However, 120V Vital AC bus voltage will be unaffected since Vital AC is derived from inverters off the 125VDC batteries. This system would not be affected by an increased switchyard voltage.

C. Correct-Raising MVARs to raise switchyard voltage will increase the apparent power of the Generator (higher reactive and the same true power.) This will result in increased generator heating. As 500 Kv bus voltage is raised, all downstream busses will increase (13Kv, 4Kv, and 480V) on both units. Per Ol-43A, the downstream bus voltage must be checked on both units to ensure all buses remain in spec and therefore operable.

D. Incorrect-Increase in grid frequency is plausible if operator confuses real and reactive loads. Raising switchyard voltage will not change grid frequency as frequency is controlled by true loads on the grid. Raising voltage will just change the reactive load on the Generator.



Question 17 lnfp Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 0.00


Topic: Consequences of an overexcited Main Generator RO Importance: 3.3 SRO Importance: 3.4 Cognitive level (High or


Tier/Group 1/1

KIA Info APE 077 Generator Voltage and Electric Grid Disturbances AK1. Knowledge of the operational implications of the following concepts as they apply to Generator Voltage and Electric Grid Disturbances: AK1 .02 Over-excitation

RO 3.3 Importance

Technical AOP-7M References Ol-43A


Learning Given an anticipated or actual grid Objective disturbance, determine and implement the

correct actions to mitigate the event in accordance with plant operating procedures

Question New Source






Comments None

Task 202.132 Respond to major grid disturbances



1& ID: Q106501 Points: 1.00

Unit-1 was operating at 100% when the following transient occurs:

• Reactor trips due to RPS testing • EOP-1 is entered • Tcold is 532°F and steady • Turbine Bypass Valve (TBV) controller output is in AUTO with an output of -2% and steady

The following then occurs:

• A SG safety suddenly lifts • SGBD is secured • The Main Steam Upstream Drain Isolation Valves are shut • Tcold has lowered to 516°F and is continuing to lower.

Which ONE of the following actions are required prior to exiting EOP-1?

Action 1-Shift TBV controller to Manual and shut the TBVs Action 2-Shut the Main Steam Isolation Valves Action 3-lnitiate Auxiliary Feedwater

A Action 1 ONLY.

B. Action 2 ONLY.

C. Actions 1 and 2 ONLY.

D. Actions 2 and 3 ONLY.

Answer: D


OPERATIONS


Answer Explanation:

A. Incorrect-Including Action 1 is plausible to the Operator since shutting TBVs would lower impact of an excess steam event. However, the TBVs will not need to be shut. A temperature well below 532F will result in the TBV controller providing a shut signal to the TBVs. Plausible since EOP-1 directs operator to verify that TBVs operate to maintain temperature 525-535, but manual actions will not be required. Actions 2 and 3 are required since EOP-1 directs that below 518F, the MSIVs are to be closed and AFW established using the motor driven pump.

B. Incorrect-Actions 2 and 3 are required since EOP-1 directs that below 518F, the MSIVs are to be closed and AFW established using the motor driven pump. Not including Action 3 is plausible to the Operator since this action could cause SG temperatures to lower further and further lower RCS temperature. However, EOP-1 assumes a heat sink is established, which includes AFW. Actions to control SG levels and feed would be addressed in EOP-4.

C. Incorrect-Including Action 1 is plausible to the Operator since shutting TBVs would lower impact of an excess steam event. However. the TBVs will not need to be shut. A temperature well below 532F will result in the TBV controller providing a shut signal to the TBVs. Plausible since EOP-1 directs operator to verify that TBVs operate to maintain temperature 525-535, but manual actions will not be required. Actions 2 and 3 are required since EOP-1 directs that below 518F, the MSIVs are to be closed and AFW established using the motor driven pump. Not including Action 3 is plausible to the Operator since this action could cause SG temperatures to lower further and further lower RCS temperature. However, EOP-1 assumes a heat sink is established, which includes AFW. Actions to control SG levels and feed would be addressed in EOP-4.

D. Correct-Actions 2 and 3 are required since EOP-1 directs that below 518F, the MSIVs are to be closed and AFW established using the motor driven pump.

Page: 56 of 235 13 April 2016




Topic: EOP-1 actions for a stuck open SG Safety RO Importance: 3.8 SRO Importance: 4.0 Cognitive level (High or


Tier/Group 1/1

KIA Info EPE E05 Excess Steam Demand EK3 Knowledge of the reasons for the following responses as they apply to the (Excess Steam Demand) EK3.3 Manipulation of controls required to obtain desired operating results during abnormal, and emergency situations

RO 3.8 Importance

Technical EOP-1 References


Learning Recall the alternate actions taken for an Objective unexpected plant cooldown.

Question New Source



10CFR Part 55.41(b)(10) 55 Content



Comments Considered a KIA match since must be aware of the reason (i.e. lowering temperature) and the setpoint that controls are manipulated during an excess steam demand

Task 201.024 Maintain RCS Heat Sink


19


ID: Q106502 . Points: 1.00


• A Regulating CEA partially drops into the core • AOP-1 Bis implemented • CEA recovery is underway per the AOP

Which ONE of the following could cause the CEA to fully drop as the CEA is recovered?

A. Loss of 2Y09.

B. CEDS NO CONTR VOLT alarm is received with CEDS Control Panel in OFF.

C. Not holding the depressed Motion Inhibit Bypass pushbutton for at least 5 seconds after CEA motion.

D. Having more than one Group Inhibit Bypass depressed when attempting to move the CEA.

Answer: C

Answer Explanation:

A. Incorrect-A loss of 2Y09 removes power from the CEA reed switch indications. The CEAs being monitored on CEAPDS will indicate full in upon a loss of 2Y09. Plausible since this would look like CEA fully inserted but in reality is only due to indication. The CEA would remain in its current withdrawn position.

B. Incorrect-The CEDS is upstream of the CEA Coil Power Programmers. Loss of voltage will not result in loss of power to the CEA itself, just the CEDS Control Panel and the CEDS Logic Cabinet. Plausible if operator believes loss of power to the CEDS Control Panel will result in loss of power to the selected CEA and the CEA will drop.

C. lncorrect-CMI can be bypassed to multiple groups. Plausible if student believes that AOP direction to select group with misaligned CEA means only one group can be bypassed when CMI is active.

D. Correct-01-42 contains a precaution that if the CEDS is turned off too early following rod motion, the CEDM cycling logic and gripper operations may not be complete and the CEA could drop. Too early is defined as 5 seconds. AOP-18 also includes the 5-second requirement where it requires the Motion Inhibit Bypass pushbutton be depressed and held a minimum of 5 seconds before moving CEAs and 5 seconds after moving CEAs.





Topic: Reasons a sinqle U-2 CEA could drop RO Importance: 2.5 SRO Importance: 2.8 Cognitive level (High or


Tier/Group 1/2

KIA Info APE 003-Dropped Control Rod AK2. Knowledge of the interrelations between the Dropped Control Rod and the following: AK2.05 Control rod drive power supplies and logic circuits

RO 2.5 Importance

Technical 01-42, 1C05 ARM, AOP-18 References


Learning Determine the actions necessary for the Objective following CEA misalignments:

• Regulating CEA greater than 7.5 inches but less than 15 inches from its group.

• Shutdown Bank CEA less than 129 inches withdrawn but less than 7.5 inches misaligned.

• Any regulating or shutdown CEA misaligned by greater than 15 inches

Question New Source






Comments None

Task 202.010 Withdraw mispositioned CEAs usinq the pull and wait method



20 ID: Q106503 · Points: 1.00

Unit-2 is at 100% power with the following initial conditions:

• PZR Level Control selected to Channel X. • Both L T-11 OX and L T-11 OY read 216" and steady

A break occurs on the reference line to L T-110X.

Which ONE of the following choices is the expected response to the reference line leak?

A. Indicated level on LT-110X rises AND indicated level on LT-110Y rises.

B. Indicated level on L T-11 OX rises AND indicated level on L T-11 OY lowers.

C. Indicated level on L T-110X lowers AND indicated level on L T-110Y rises.

D. Indicated level on LT-110X lowers AND indicated level on LT-110Y lowers.

Answer: B


OPERATIONS


Answer Explanation:

A. lncorrect-L T-11 OX indicated level will rise. PZR level indication reference lines are filled to a constant level due to condensing PZR vapor space steam that keeps the reference line full. With the reference line as the higher pressure, any abnormal condition that decreases D/P will cause indicated level to be higher than actual level.

LT-110Y indicated level will lower. Since Level control is selected to Channel X, the control system will respond to not only the leak but also to the indicated>actual condition on LT-110X. The level control system will increase letdown since indicated level will also be > setpoint, lowering actual PZR level. Indicated level on L T-11 OY will be the same as actual PZR level, which is lowering. L T-11 OY rising is plausible if the logic surrounding letdown response is reversed, and the Operator determines letdown flow will lower rather than rise.

B. Correct-L T-110X indicated level will rise. PZR level indication reference lines are filled to a constant level due to condensing PZR vapor space steam that keeps the reference line full. With the reference line as the higher pressure, any abnormal condition that decreases D/P will cause indicated level to be higher than actual level.

LT-110Y indicated level will lower. Since Level control is selected to Channel X, the control system will respond to not only the leak but also to the indicated>actual condition on L T-11 OX. The level control system will increase letdown since indicated level will also be > setpoint, lowering actual PZR level. Indicated level on L T-11 OY will be the same as actual PZR level, which is lowering.

C. lncorrect-L T-11 OX indicated level will rise. PZR level indication reference lines are filled to a constant level due to condensing PZR vapor space steam that keeps the reference line full. With the reference line as the higher pressure, any abnormal condition that decreases D/P will cause indicated level to be higher than actual level. LT-110X indicated level lowering is plausible to the Operator if the DIP logic across the LT is reversed.

L T-110Y indicated level will lower. Since Level control is selected to Channel X, the control system will respond to not only the leak but also to the indicated>actual condition on L T-110X. The level control system will increase letdown since indicated level will also be > setpoint, lowering actual PZR level. Indicated level on L T-11 OY will be the same as actual PZR level, which is lowering. L T-11 OY rising is plausible if the logic surrounding letdown response is reversed, and the Operator determines letdown flow will lower rather than rise.

D. lncorrect-L T-110X indicated level will rise. PZR level indication reference lines are filled to a constant level due to condensing PZR vapor space steam that keeps the reference line full. With the reference line as the higher pressure, any abnormal condition that decreases D/P will cause indicated level to be higher than actual level. LT-110X indicated level lowering is plausible to the Operator if the D/P logic across the LT is reversed.

L T-11 OY indicated level will lower. Since Level control is selected to Channel X, the control system will respond to not only the leak but also to the indicated>actual condition on L T-11 OX. The level control system will increase letdown since indicated level will also be> setpoint, lowering actual PZR level. Indicated level on L T-110Y will be the same as actual PZR level, which is lowering.

Page: 63 of 235 13 April 2016




Topic: PZR reference line leak impact on L T-11 OY level indication RO Importance: 2.8 SRO Importance: 3.1 Cognitive level (High or


Tier/Group 1/2

KIA Info APE 028 Pressurizer (PZR) Level Control Malfunction AK1. Knowledge of the operational implications of the following concepts as they apply to Pressurizer Level Control Malfunctions: AK1 .01 PZR reference leak abnormalities

RO 2.8 Importance

Technical System Description 064 (Reactor Coolant References System Instrumentation)

OM 60729SH0001 References None provided

Learning Predict the effect of the changes in the Objective following parameters on pressure and

differential pressure detectors: • Temperature • Pressure • Radiation

Question New Source






Comments None

Task 202.012 Respond to RCS leakage exceeding capacity of one charqinq pump, modes 1 & 2

OPERATIONS Page 65 of 235 13April2016

21 .


ID: Q106504 Points: 1.00

Both units are at 100% power with the following initial conditions:

• U-1 Quench Tank pressure is 8 PSIG with preparations being made to vent the tank. WGS-2180-CV, WGS CONTMT ISOL, has been opened.

• U-2 CVCS IX transfer to the SRMT in progress • WGST is 3 PSIG and slowly lowering • 11 WG Compressor is running • 12 WGDT is in service at 60 PSIG and slowly rising • SFP fuel moves are in progress


• Gaseous Waste Discharge RMS, O-Rl-2191, comes into alarm • Both the WRNGM and Main Vent Gaseous RMS are rising • You are contacting Health Physics to help identify the abnormal conditions

Which ONE of the following systems/locations is the most likely source of the Gaseous Release?

A SFP Area

B. SRMT vent

C. 12 WGDT RV

D. WG lines to the WGST

Answer: B

Answer Explanation:

A Incorrect-A SFP gaseous release from a fuel failure will cause both Main Vent Gaseous and WRNGM indications to rise, but it will not cause the Gaseous Rl-2191 to come into alarm. The release would be contained to the Aux Building, the SFP ventilation, and Main Vent ventilation systems, not the Waste Gas system.

B. Correct-With Gaseous Rl-2191 in alarm, an uncontrolled gaseous release is occurring. Sources of gaseous waste that are downstream of the WG Discharge CVs (2191/2192) are always possible sources of a release. This includes the plant vent header. Transferring eves IX is the most likely source since the SRMT is vented to the vent header at all times.

C. Incorrect-A lifting WGDT RV will not discharge to the Waste Gas header, but rather, back to the WGST. If WGST is lowering, discharge is not occurring.

D. Incorrect-When venting the QT to the WG system, 3 valves need to be opened to conduct the vent (WGS-2180/2181/RC-400-CV). Only one has been opened. If operator believed vent was in progress, WGST and WGDT parameters are still normal for the venting process. Any leaks on the vent path would again enter the Aux Building ventilation system and not register on Gaseous Rl-2191.





Topic: Likely source of uncontrolled Gaseous release RO Importance: 3.1 SRO Importance: 4.0 Cognitive level (High or

H Low): NRG KIA Info, References:

Tier/Group 1/2

KIA Info APE 060 Accidental Gaseous Radwaste Release AA2 Ability to determine and interpret the following as they apply to the Accidental Gaseous Radwaste: AA2.02 The possible location of a radioactive-gas leak, with the assistance of PEO, health physics and chemistry personnel

RO 3.1 Importance

Technical AOP-6C, 60735SH0001 References


Learning Given AOP-6C, respond to the following: Objective • Waste Gas system rupture or leakage

• An uncontrolled Waste Gas release • High Airborne in the Auxiliary Building

Question New Source






Comments None

Task 202.049 Respond to uncontrolled Waste Gas release


22


ID: 0106505 · Points: 1 :oo Unit-2 is operating at 100% power with the following initial conditions:

• Positive pressure vent of the Containment in progress per 01-41 B

A Large Break LOCA occurs and the following RMS indications are noted:

• CNTMT RAD LVL HI alarm is received on 1C10 • Rl-5317A reads 22 R/Hr • Rl-53178 reads 23 R/Hr

Which ONE of the following automatic actions are initiated by the Containment High Range Radiation Monitors (Rl-5317 A and Rl-53178)?

A. The 21, 22, and 23 Iodine Removal Units STARTED.

B. The 21 and 22 Penetration Room Exhaust Fans STARTED.

C. The Containment Purge CVs, 2-CPA-1410 and 2-CPA-1412, SHUT.

D. The Hydrogen Purge INBD and OUTBD ISOL MOVs, 2-HP-6900 and 2-HP-6901, SHUT.

Answer: D

Answer Explanation:

A. Incorrect-The IRUs only start because of a SIAS. Plausible since degrading containment conditions start the IRUs.

B. Incorrect-The Pene Room fans only start because of a CIS. Plausible since degrading containment conditions start the Pene Room fans.

C. Incorrect-The Containment purge CVs are normally shut at 100% power. They are not the method of performing a containment vent at power, but rather only in lower modes. Plausible if RMS 5316A-D and CRS, which will shut the purge CVs, are confused with Containment Hi-Range RMS 5317NB.

D. Correct-A positive pressure vent of the Containment will open HP-6900/6901. Per 1C10 ARM J-04, receiving the COTMT RAD LVL HI alarm will shut HP-6900/6901. SIAS also closes the HP MOVs.





Topic: Automatic actions performed by Containment High Range Radiation Monitors (RE-5317-A&B)



Tier/Group 1/2

KIA Info APE 061 Area Radiation Monitoring (ARM) System Alarms AA 1. Ability to operate and I or monitor the following as they apply to the Area Radiation Monitoring (ARM)System Alarms: AA 1 . 01 Automatic actuation

RO 3.6 Importance

Technical 1C10 ARM References 01-41 B


Learning Recall the purpose of the following Objective systems:

• Radiation Monitoring Systems (Area & Process)

• Containment Hi-Range Radiation Monitor

• Main Steam Line Radiation Monitor (MSLRM)

• Wide Range Noble Gas Monitor (WRNGM)

Question Modified from Q20587 Source





10CFR Part 55.41 (b)(11) 55 Content

Comments 020587 used in 2002 NRC RO Exam

Task 079.006 Operate the Containment High Range Radiation Monitor

OPERATIONS Page: 71 of 235 13Apri12016

23


ID: Q74599 Pqints:' 1'.QO

The following transient has occured:

• A fire has forced the evacuation of the Unit-1 Control Room • AOP-9A, Control Room Evacuation and Shutdown due to a Severe Control Room Fire, is

implemented.

Which ONE of the following describes 1) the AFW source that will be used first to maintain SG level in AOP-9A and 2) the maximum amount of time that may elapse per procedure assumptions prior to establishing AFW?

A. 1) 11or12 AFW pumps 2) 30 minutes

B. 1) 11or12 AFW pumps 2) 60 minutes

C. 1) 13 AFW pump 2) 60 minutes

D. 1) 13 AFW pump 2) 30 minutes

Answer: A

Answer Explanation:

A. Correct-Per AOP-9A, the steam driven AFW pumps are utilized first to feed the SGs. The AOP-9A assumes that AFW flow will be established within 30 minutes.

B. Incorrect-Per AOP-9A, the steam driven AFW pumps are utilized first to feed the SGs. The AOP-9A assumes that AFW flow will be established within 30 minutes. Plausible since 60 minute time is associated with reestablishing charging to the RCS in the AOP.

C. Incorrect-Per AOP-9A, the steam driven AFW pumps are utilized first to feed the SGs. The AOP-9A assumes that AFW flow will be established within 30 minutes. Plausible since the 13 AFW is started in the procedure, but not until later when SG fill and chills are being conducted. Also, 60 minute time is associated with reestablishing charging to the RCS in the AOP.

D. Incorrect-Per AOP-9A, the steam driven AFW pumps are utilized first to feed the SGs. The AOP-9A assumes that AFW flow will be established within 30 minutes. Plausible since the 13 AFW is started in the procedure, but not until later when SG fill and chills are being conducted.





Topic: AOP-9A AFW pumps and time limits to restore AFW RO Importance: 4.6/4.1 SRO Importance: 4.6/4.3 Cognitive level (High or

L Low): NRC K/A Info, References:

Tier/Group 1/2

K/A Info APE 068 Control Room Evacuation 2.1 Conduct of Operations 2.1.20 Ability to interpret and execute procedure steps

RO 4.6 Importance

Technical AOP-9A References


Learning Operate the AFW system components as Objective required to maintain RCS temperatures and

SG levels to support AOP-9A plant shutdown.


Question No use on NRC exam History



Comments None



I Task I 202.113 Initiate AFW to both S/Gs I Additional KIA Match Info APE 061-Area Radiation Monitoring (ARM) System Alarms AA 1. Ability to operate and I or monitor the following as they apply to the Control Room Evacuation: AA 1. 03 SIG level


24


10':: Q28439: > Points: 1.00

A caution within EOP-3, Loss of All Feedwater, states that Once Through Core Cooling (OTCC), must be initiated before CETs reach or exceed 560°F.

Which ONE of the following is the basis for this temperature limit?

A. Ensures the RCS is maintained subcooled throughout OTCC.

B. Ensures the inventory in the core will not be displaced into the Pressurizer.

C. Ensure RCS core cooling flow is sufficient to lower core temperature.

D. Ensures RCS pressure remains high enough to prevent HPSI Pump damage.

Answer: C

Answer Explanation:

A. Incorrect-Plausible since subcooled conditions are desired at all times. However, the RCS will be in a saturated condition due to the PORVs being opened.

B. Incorrect-Plausible since inventory in the core is desired at all times. The RCS will be in a saturated condition due to the PORVs being opened. Water will be displaced into the low pressure area (the Pressurizer).

C. Correct-Per the EOP-3 Basis Doc, If OTCC initiated above this value the HPSI pump flow may be insufficient for core cooling flow.

D. Incorrect-Plausible since, with saturated conditions, a higher temperature is associated with a higher pressure on the RCS Pressure/Temperature curve. However, runout of the HPSI pumps is not probable (OBA). Would also be prevented by complying with procedure direction to verify HPSI flow PER EOP ATTACHMENT (10), HIGH PRESSURE SAFETY INJECTION FLOW.




System ID: 28439 User-Defined ID: Q28439 Cross Reference Number: SR0-201-3-1-14

,

Topic: Basis for initiatinq OTCC prior to 560°F RO Importance: 3.9/3.2 SRO Importance: 4.1/3.7 Cognitive level (High or


Tier/Group 1/2

KIA Info EPE 074-lnadequate Core Cooling EK2 Knowledge of the interrelations between the and the following Inadequate Core Cooling: EK2.04 HPI pumps

RO 3.9 Importance

Technical EOP-3, Loss of All Feedwater References


Learning Given a loss of all feedwater the trainee will Objective be able to identify, understand the basis and

take appropriate actions per plant operating procedures to mitigate the event







Comments None

Task 201.034 Start once through core cooling

Additional KIA Match Info CE/E06 Loss of Feedwater EK3 Knowledge of the reasons for the following responses as they apply to the (Loss of Feedwater) EK3.2 Normal, abnormal and emergency operating procedures associated with (Loss of Feedwater)


25


ID::ld8935 Points: 1.00

During an excess steam demand event, the unaffected SIG is maintained within 25°F of CET temperature using its ADV.

Which ONE of the following is the reason for controlling SIG temperature in this band?

A. Minimizes the potential for pressurized thermal shock if a heatup of the RCS occurs following an excessive cooldown of the RCS.

B. Minimizes the formation of tube voids, in the affected SIG, after blowdown is complete.

C. Minimizes the RCS cooldown that takes place during blowdown of the affected SIG.

D. Minimizes the differential pressure between the SIGs, thereby allowing reset of the AFAS Block signal.

Answer: A

Answer Explanation:

A. Correct-See EOP-4 Technical Basis Document, Step IV.H.2 (page 27). The 25 °F limit is an operational limit associated with PTS mitigation during a cooldown event. Its basis supports the same basis as the broader concept of cooldown limits as referenced in the KA, under which this limit lies. This action sets up the operational controls to support PTS prevention when the uncontrolled cooldown has been completed.

B. Incorrect- Plausible since SIG pressure will lower on the secondary side of the SIG, and the Operator could interpret an elevated temperature with a lower pressure on the secondary side as leading to the formation of tube voids. However, SIG tube voiding is determined by RCS pressure being less than saturation pressure for that SIG, and once BID is complete SIG pressure will be zero.

C. lncorrect-Plausibe if Opertor confuses intent of step which is to minimize the temperture difference between the RCS and the unaffected SIG. If the Operator interprets this step to mean the unaffected SIG will be cooler than the RCS, than by holding temperature within 25F of CETs, then this would minimize the RCS cooldown. However, the unaffected SIG will be higher in temperature and will act as a heat source to the RCS. The RCS cooldown during the blowdown phase is determined by the size of the leak.

D. Incorrect-Plausible since limiting chance for SIG DIP maximizes AFW flow. However, AFW should be blocked to the affected SIG. AFAS Block will occur, isolating Auxiliary Feedwater flow to the SIG with the lower pressure which is the affected SIG.




System ID: 18935 User-Defined ID: L18935 Cross Reference Number:

...

Topic: ESDE: Basis for maintaining unaffected SIG within 25° F of CETs


Low Low): NRG KJA Info, References:

Tier/Group 1/2

KJA Info CE/A 11 RCS Overcooling AK3. Knowledge of the reasons for the following responses as they apply to the (RCS Overcooling) AK3.2 Normal, abnormal and emergency operating procedures associated with (RCS Overcooling).

RO 2.9 Importance

Technical EOP-4, Excess Steam Demand Event References


Learning Given conditions and/or parameters Objective associated with an ESDE, determine the

appropriate operator actions


Question 2010 NRG RO Exam History




10CFR SS.41(b)(S/10) Part SS Content

Comments None

Task 201.024 Maintain RCS Heat Sink Additional KIA Match Info 039 Main and Reheat Steam System (MRSS) KS Knowledge of the operational implications of the following concepts as the apply to the MRSS: KS.OS Bases for RCS cooldown limits


26


10: L14420 Points: 1.00


• The reactor is manually tripped due to low vacuum • All Reactor Coolant Pumps were tripped due to an inadvertent CIS that actuated in EOP-0

The following conditions now exist:

• Thot is 547°F and lowering • Tcold is 534°F and lowering • RCS pressure is 2050 PSIA and slowly rising • Condenser vacuum is 23 In-Hg • No manual operator action has been taken with the Atmospheric Dump Valves (ADVs) or

Turbine Bypass Valves (TBVs)

Which ONE of the following ADV and TBV responses would be expected?

A. ADVs - full shut; TBVs - modulating open.

B. ADVs - full open; TBVs - full open.

C. ADVs - modulating open; TBVs - full shut.

D. ADVs - modulating open; TBVs - modulating open.

Answer: D

Answer Explanation:

A. lncorrect-ADVs full shut is plausible since ADVs are full shut when Tavg is <534F. However, Tavg is >535F and the ADVs will modulate. The TBVs should be modulating to maintain SIG pressure per the setpoint of the TBV Controller.

B. lncorrect-ADVs full open since ADVs go full open on a trip. However, the ADVs will modulate once Tavg is <557F. The TBVs full open is plausible since the TBVs go full open initially on a trip from 100%. However, the TBVs will modulate once Tavg is <557F The TBVs should be modulating to maintain SIG pressure per the setpoint of the TBV Controller.

C. Incorrect-With Tave >535F, the ADVs will modulate. The TBVs full shut is plausible if the Operator confuses the AOP-7G trip criteria parameter of 23.5 In-HG with the 20 In-Hg value at which the TBVs will be locked out due to low condenser vacuum and shut. However, the TBVs will modulate once Tavg is <557F. The TBVs should be modulating to maintain SIG pressure per the setpoint of the TBV Controller.

D. Correct-With Tave >535F, the ADVs will modulate. The TBVs should be modulating to maintain SIG pressure per the setpoint of the TBV Controller.





/

Topic: L 14420 Given post-trip Natural Gire conditions, identify ADV & TBV status



Tier/Group 1/2

KIA Info CE/A 13-Natural Circulation Operations AA 1. Ability to operate and I or monitor the following as they apply to the (Natural Circulation Operations) AA 1.1 Components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

RO 3.3 Importance



Learning Given a Loss of Forced Circulation/Loss of Objective Offsite Power the trainee will be able to

identify, understand the basis and take appropriate actions per EOP-2, AOP-3E, or AOP-3F to mitigate the event


Question NRG 2008 RO Exam History





Comments None


Additional KIA Match Info CE A 13-Natural Circulation Operations AK2. Knowledge of the interrelations between the (Natural Circulation Operations) and the following: AK2.1 Components and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.



ID: Q106513 Points: 1.00

Unit-1 is performing a startup following a trip. An RCS soak is being performed at 1900 PSIA with the following conditions:

• RV-200 leakage to the Quench Tank - 0.7 GPM • Charging Pump packing leakage - 0.8 GPM • Total RCS leakage - 1.6 GPM • Sump frequency - 6 hours

RCS pressure is raised to 2250 PSIA and a leak on 12 SG L T-1123C root valve is identified. The following conditions now exist:

• RV-200 leakage to the Quench Tank - 0.6 GPM • PORV-404 leakage to the Quench Tank - 0.6 GPM • Charging Pump packing leakage - 0.8 GPM • Total RCS leakage - 2.2 GPM • Sump frequency - 30 mins

Which ONE of the following Technical Specification leakage limits, if any, are exceeded?

A. Identified leakage.

B. Unidentified leakage.

C. Primary to Secondary leakage.

D. None have been exceeded.

Answer: D


OPERATIONS


Answer Explanation:

A. Incorrect-With pressure at 1900 PSIA, the leakage values are: ldentified=1.5 GPM (0. 7 GPM RV-200 + 0.8 GPM Charging.) Unidentified=0.1 GPM (1.6 GPM Total RCS - 1.5 GPM Identified.) A sump frequency of 6 hours is equal to 0.14 GPM of water entering the sump (49 Gallons/360 Mins.). When pressure is raised to 2250 PSIA, the leakage values are: ldentified=2.0 GPM (0.6 GPM RV-200 + 0.6 GPM PORV-404 + 0.8 GPM Charging.) Unidentified=0.2 GPM (2.2 GPM Total RCS - 2.0 GPM Identified.) A sump frequency of 30 mins is equal to 1.6 GPM of water entering the sump (49 Gallons/30 Mins.) None of the leakage limits in TS 3.4.13 have been exceeded. Plausible if operator believes that increase in sump frequency is due to RCS leakage.

B. Incorrect-With pressure at 1900 PSIA, the leakage values are: ldentified=1.5 GPM (0. 7 GPM RV-200 + 0.8 GPM Charging.) Unidentified=0.1 GPM (1.6 GPM Total RCS - 1.5 GPM Identified.) A sump frequency of 6 hours is equal to 0.14 GPM of water entering the sump (49 Gallons/360 Mins.). When pressure is raised to 2250 PSIA, the leakage values are: ldentified=2.0 GPM (0.6 GPM RV-200 + 0.6 GPM PORV-404 + 0.8 GPM Charging.) Unidentified=0.2 GPM (2.2 GPM Total RCS - 2.0 GPM Identified.) A sump frequency of 30 mins is equal to 1.6 GPM of water entering the sump (49 Gallons/30 Mins.) None of the leakage limits in TS 3.4.13 have been exceeded. Plausible if operator believes that increase in sump frequency is due to RCS leakage and makes a math error while converting sump frequency to actual RCS leak (49/3=16.3 GPM versus correct conversion of 49/30/1.6 GPM.)

C. Incorrect-With pressure at 1900 PSIA, the leakage values are: ldentified=1.5 GPM (0. 7 GPM RV-200 + 0.8 GPM Charging.) Unidentified=0.1 GPM (1.6 GPM Total RCS - 1.5 GPM Identified.) A sump frequency of 6 hours is equal to 0.14 GPM of water entering the sump (49 Gallons/360 Mins.). When pressure is raised to 2250 PSIA, the leakage values are: ldentified=2.0 GPM (0.6 GPM RV-200 + 0.6 GPM PORV-404 + 0.8 GPM Charging.) Unidentified=0.2 GPM (2.2 GPM Total RCS - 2.0 GPM Identified.) A sump frequency of 30 mins is equal to 1.6 GPM of water entering the sump (49 Gallons/30 Mins.) None of the leakage limits in TS 3.4.13 have been exceeded. Plausible if operator believes that a leak from a SG LT is associated with the RCS primary side and the increase in the unidentified RCS leak (-equal to the increase in the sump frequency) is due to primary to secondary leakage.

D. Correct-With pressure at 1900 PSIA, the leakage values are: ldentified=1.5 GPM (0. 7 GPM RV-200 + 0.8 GPM Charging.) Unidentified=0.1 GPM (1.6 GPM Total RCS - 1.5 GPM Identified.) A sump frequency of 6 hours is equal to 0.14 GPM of water entering the sump (49 Gallons/360 Mins.). When pressure is raised to 2250 PSIA, the leakage values are: ldentified=2.0 GPM (0.6 GPM RV-200 + 0.6 GPM PORV-404 + 0.8 GPM Charging.) Unidentified=0.2 GPM (2.2 GPM Total RCS - 2.0 GPM Identified.) A sump frequency of 30 mins is equal to 1.6 GPM of water entering the sump (49 Gallons/30 Mins.) None of the leakage limits in TS 3.4.13 have been exceeded.

Page: 85 of 235 13 April 2016




Topic: TS RCS LeakaQe Evaluation RO Importance: 2.9 SRO Importance: 3.7 Cognitive level (High or


Tier/Group 1/2

KIA Info CE/A16 Excess RCS Leakage AA2 Ability to determine and interpret the following as they apply to the (Excess RCS Leakage) AA2.2 Adherence to appropriate procedures and operation within the limitations in the facility*s license and amendments.

RO 2.9 Importance

Technical TS 3.4.13 References STP 0-27


Learning Given the Technical Specifications and Objective various plant conditions, determine the

correct actions per the LCO

Question New Source






Comments None

Task 064.016 Evaluate RCS leakage



ID: Q106514 Points: 1.00

Reactor Coolant Pump (RCP) power supplies have been aligned as follows:

• Unit-2 RCPs (21A-21 B-22A-228) are powered from their normal source • Unit-1 RCPs (11A-11 B-12A-128) have been shifted to their alternate power source

Which ONE of the following MUST be lost to trip all Unit-1 and Unit-2 RCPs?

OPERATIONS

A. ONLY 12 13Kv Bus.

B. ONLY 22 13Kv Bus.

C. BOTH 11 and 22 13Kv Susses.

D. BOTH 12 AND 21 13Kv Susses.

Answer: B

Answer Explanation:

A. Incorrect-The normal power supply to U-2 RCPs is 22 13Kv Bus. When U-1 RCPs are shifted to their alternate power supply, they will be aligned also to 22 13Kv bus. The loss of just 22 13Kv bus in this situation will result in the loss of both units RCPs. Plausible if operator confuses normal power supplies and believes all will be powered from 12 13Kv bus.

B. Correct-The normal power supply to U-2 RCPs is 22 13Kv Bus. When U-1 RCPs are shifted to their alternate power supply, they will be aligned also to 22 13Kv bus. The loss of just 22 13Kv bus in this situation will result in the loss of both units RCPs.

C. Incorrect-The normal power supply to U-2 RCPs is 22 13Kv Bus. When U-1 RCPs are shifted to their alternate power supply, they will be aligned also to 22 13Kv bus. The loss of just 22 13Kv bus in this situation will result in the loss of both units RCPs. 11 13Kv bus does not have to be lost to trip any of the RCPs. Plausible if operator believes alternate U-1 power supply is 11 13Kv bus.

D. Incorrect-The normal power supply to U-2 RCPs is 22 13Kv Bus. When U-1 RCPs are shifted to their alternate power supply, they will be aligned also to 22 13Kv bus. The loss of just 22 13Kv bus in this situation will result in the loss of both units RCPs. Plausible if operator confuses normal power supplies and believes one unit will be powered from 12 13Kv bus and one unit from 21 13Kv bus.

Page: 88 of 235 13 April 2016




Topic: RCP Impacts on loss of 13 KV power RO Importance: 3.1 SRO Importance: 3.1 Cognitive level (High or


Tier/Group 2/1

KIA Info 003 Reactor Coolant Pump System (RCPS) K2 Knowledge of bus power supplies to the following: K2.01 RCPS

RO 3.1 Importance

Technical Ol-1A, 01-278 References


Learning State the power supply for the RCPs Objective Question New Source




Comments None

Task 064.011 Shift power sources to RCP busses


29


10: Q106637 Points: 1.00

Unit-1 is operating with the following conditions:

• SOC has been secured to support Plant Heatup • RCS Temperature is 205°F and slowly rising • Neither 118 or 12A RCP are available for start

Which ONE of the following conditions is the minimum required for STP 0-948-1, RCS/SOC Loop Operability Verification (Modes 4,5 and 6)?

A. 11 Loop-11A RCP is Off with 11 S/G level >-40", aligned to a Unit-1 power supply, and meets the starting criteria of Ol-1A 12 Loop-128 RCP is Off with 12 SIG level >-40", aligned to a Unit-1 power supply, and meets the starting criteria of Ol-1A

8. 11 Loop-11A RCP is Operating with 11 SIG level >-170" 12 Loop-128 RCP is Off with 12 S/G level >-170", aligned to a Unit-2 power supply, and meets the starting criteria of Ol-1A

C. 11 Loop-11A RCP is Off with 11 S/G level >-40" aligned to a Unit-1 power supply, and meets the starting criteria of Ol-1A 12 Loop-128 RCP is Operating with 12 SIG level >-40"

0. 11 Loop-11 A RCP is Operating with 11 S/G level >-170" 12 Loop-128 RCP is Operating with 12 SIG level >-170"

Answer: C


OPERATIONS


Answer Explanation:

A. lncorrect-STP 0-948-1 is the surveillance that supports TS 3.4.6. Given the RCS temperature of 205F, the plant is in Mode 4 and TS 3.4.6 requires 2 loops to be operable and one of those loops to be in operation. Per STP 0-0948, a RCP is operable if its associated SIG level is >-40", the pump is running, or if not running, the pump is electrically aligned and the starting requirements of Ol-1A are met. Two pumps off are plausible to the Operator since a note in the TS allows forced circulation to be secured for a period of time when transitioning between RCPs and SOC.

8. lncorrect-STP 0-948-1 is the surveillance that supports TS 3.4.6. Given the RCS temperature of 205F, the plant is in Mode 4 and TS 3.4.6 requires 2 loops to be operable and one of those loops to be in operation. Per STP 0-0948, a RCP is operable if its associated SIG level is >-40", the pump is running, or if not running, the pump is electrically aligned and the starting requirements of Ol-1A are met. A SIG level of -170" is plausible to the Operator since this is the EOP-0 value associated with a SIG meeting Core and RCS Heat Removal Safety Function Status Check.

C. Correct-STP 0-948-1 is the surveillance that supports TS 3.4.6. Given the RCS temperature of 205F, the plant is in Mode 4 and TS 3.4.6 requires 2 loops to be operable and one of those loops to be in operation. Per STP 0-0948, a RCP is operable if its associated SIG level is >-40", the pump is running, or if not running, the pump is electrically aligned and the starting requirements of Ol-1A are met.

D. lncorrect-STP 0-948-1 is the surveillance that supports TS 3.4.6. Given the RCS temperature of 205F, the plant is in Mode 4 and TS 3.4.6 requires 2 loops to be operable and one of those loops to be in operation. Per STP 0-0948, a RCP is operable if its associated SIG level is >-40", the pump is running, or if not running, the pump is electrically aligned and the starting requirements of 01-1 A are met. A SIG level of -170" is plausible to the Operator since this is the EOP-0 value associated with a SIG meeting Core and RCS Heat Removal Safety Function Status Check.

Page: 91 of 235 13 April 2016




Topic: RCP Requirements to Satisfy STP 0-948 RO Importance: 3.7 SRO Importance: 4.1 Cognitive level (High or High Low): NRC KIA Info, References:

Tier/Group 2/1

KIA Info 003 Reactor Coolant Pump 2.2 Equipment Control 2.2.12 Knowledge of surveillance procedures

RO 3.7 Importance

Technical STP 0-948-1 (RCS/SOC Loop Operability References Verification), TS 3.4.6, OP-1 (Plant Startup

from Cold Shutdown), EOP-0 (Post Trip Immediate Actions)


Learning Specify the minimum flow required by Objective Technical Specifications for the RCS for g_[l_

Modes

Question New Source


Cognitive Comprehension or Analysis Memory or Level Fundamental Knowledge


Comments None



I~' T_as_k -~' 0_64_.o_o_3 c_o_m_m_en_ce_R_C_S_h_ea_tu_p __ ~11


30


ID: Q106515 Points~ 1.00

Unit-1 is operating at 100% with the following initial conditions:

• Power just returned to 100% following Turbine Valve testing • PZR boron equilization in progress • Two Charging pumps in operation • RRS Channel X is selected at 1 C06 • 1-HS-111 below 1 C06 selected to "Loop 11 A to RRS", which inputs 1-TE-111 Y to RRS and 1-

TE-115 to MPT


• RTD 1-TE-111Y (11A Loop Tcold} fails low

Which ONE of the following, if any, is the immediate impact on the RCS and why?

A PZR level lowers since one Charging Pump immediately stops and Letdown rises.

B. No change to actual RCS parameters since RRS Channel "X" is selected.

C. RCS Pressure lowers since all Backup Heaters turn off.

D. RCS Temperature lowers since ADV begins to open.

Answer: A


OPERATIONS


Answer Explanation:

A. Correct-When the Tcold input from TE-111Y fails low, this will cause Tavg calculated by RRS to lower. The PZR level control system will see a lower setpoint as the lower Tavg is normally associated with a lower power and hence, a lower PZR level setpoint. With actual PZR level greater than setpoint, CVCS will respond by attempting to lower level (by securing backup charging pumps and increasing letdown.) PZR level will begin to lower immediately.

B. Incorrect-When the Tcold input from TE-111Y fails low, this will cause Tavg calculated by RRS to lower. The PZR level control system will see a lower setpoint as the lower Tavg is normally associated with a lower power and hence, a lower PZR level setpoint. With actual PZR level greater than setpoint, eves will respond by attempting to lower level (by securing backup charging pumps and increasing letdown.) PZR level will begin to lower immediately. Plausible if Operator believes that since RRS-X is selected that TE-111Y will have no impact on RRS output when in fact it will.

C. Incorrect-When the Tcold input from TE-111Y fails low, this will cause Tavg calculated by RRS to lower. The PZR level control system will see a lower setpoint as the lower Tavg is normally associated with a lower power and hence, a lower PZR level setpoint. With actual PZR level greater than setpoint, CVCS will respond by attempting to lower level (by securing backup charging pumps and increasing letdown.) PZR level will begin to lower immediately. Backup heaters will remain energized as PZR level system will react as if a PZR insurge has occurred (lndicated>Setpoint). Plausible since RCS pressure could lower since PZR level is lowering.

D. Incorrect-When the Tcold input from TE-111Y fails low, this will cause Tavg calculated by RRS to lower. The PZR level control system will see a lower setpoint as the lower Tavg is normally associated with a lower power and hence, a lower PZR level setpoint. With actual PZR level greater than setpoint, CVCS will respond by attempting to lower level (by securing backup charging pumps and increasing letdown.) PZR level will begin to lower immediately. Plausible since Tavg will lower to a point where ADVs should modulate, but since a Turbine Trip is not active, the ADVs will remain shut.

Page 95 of 235 13 April 2016



System ID: 106515 User-Defined ID: Q106515 Cross Reference Number: 0

Topic: CVCS response for TE-111X failinq low RO Importance: 3.2 SRO Importance: 3.4 Cognitive level (High or


Tier/Group 2/1

KIA Info 004 Chemical and Volume Control System K5 Knowledge of the operational implications of the following concepts as they apply to the eves: K5.16 Source of T-ave. and T-ref. signals to control and RPS

RO 3.2 Importance

Technical OP-3 References 01-7


Learning Given conditions and/or parameters Objective determine where pressurizer level, setpoint,

and controller output signals for the following modes of pressurizer level control are being generated: • Remote auto • Local auto • Manual

Question New Source






Comments None

Task 056.003 Respond to failures of Temperature lnstrument(s) input to RRS/MPT


31


ID: Q106516 Points:A,oo

Unit-2 is operating in Mode 4 with the following initial conditions:

• Shutdown to Mode 5 in progress • Shutdown Cooling (SOC) has been initiated • RCS pressure is 240 PSIA and steady • RCS temperature is 250°F and slowly lowering • 21 LPSI Pump is tagged out


• 22 LPSI Pump trips • AOP-38 is implemented

Which ONE of the following actions are required to restore SOC using the Containment Spray system?

A RCS pressure can remain at 240 PSIA; RCS temperature can remain at 250°F; and a Containment Spray pump can be immediately started to restore SOC flow.

B. RCS pressure can remain at 240 PSIA; RCS temperature can remain at 250°F; and a Containment Spray pump must be mechanically aligned to SOC before restoring SOC flow.

C. RCS pressure must be lowered to <170 PSIA; RCS temperature can remain at 250°F; and a Containment Spray pump must be mechanically aligned to SOC before restoring SOC flow.

0. RCS pressure must be lowered to <170 PSIA; RCS temperature must be lowered to 200°F; and a Containment Spray pump must be mechanically aligned to SOC to restore SOC flow.

Answer: C


OPERATIONS


Answer Explanation:

A. Incorrect-With the loss of the single available LPSI pump, a SOC flowpath is still available since this is not a common mode failure. A CS pump is able to restore SOC flow provided RCS pressure is <170 PSIA (to protect CS piping) and the CS pump is mechanically aligned to SOC (since its discharge valve will normally be shut on SOC and the LPSI normal suction will need to be opened if on SOC to connect the CS to the SOC suction piping.) Plausible since AOP-38 allows the starting of a LPSI pump (if one were available) to restore SOC flow and Operator believes same action applies to CS pump.

B. Incorrect-With the loss of the single available LPSI pump, a SOC flowpath is still available since this is not a common mode failure. A CS pump is able to restore SOC flow provided RCS pressure is <170 PSIA (to protect CS piping) and the CS pump is mechanically aligned to SOC (since its discharge valve will normally be shut on SOC and the LPSI normal suction will need to be opened if on SOC to connect the CS to the SOC suction piping.)

C. Correct-With the loss of the single available LPSI pump, a SOC flowpath is still available since this is not a common mode failure. A CS pump is able to restore SOC flow provided RCS pressure is <170 PSIA (to protect CS piping) and the CS pump is mechanically aligned to SOC (since its discharge valve will normally be shut on SOC and the LPSI normal suction will need to be opened if on SOC to connect the CS to the SOC suction piping.)

0. Incorrect-With the loss of the single available LPSI pump, a SOC flowpath is still available since this is not a common mode failure. A CS pump is able to restore SOC flow provided RCS pressure is <170 PSIA (to protect CS piping) and the CS pump is mechanically aligned to SOC (since its discharge valve will normally be shut on SOC and the LPSI normal suction will need to be opened if on SOC to connect the CS to the SOC suction piping.) Plausible if Operator believes TS requires CS remain in service until Mode 5 is reached (remain operable in Modes 1-4) like Containment TS requires.

Page: 99 of 235 13 April 2016




Topic: Loss of LPSI and use of CS pumps to restore SOC RO Importance: 3.1 SRO Importance: 3.2 Cognitive level (High or


Tier/Group 2/1

KIA Info 005 Residual Heat Removal System (RHRS) K3 Knowledge of the effect that a loss or malfunction of the RHRS will have on the following: K3.06 CSS

RO 3.1 Importance



Learning Recall the steps per AOP-38 for restoring Objective SOC flow when neither LPSI pump is

available.

Question New Source




Comments None


32


I Task 1

202.025 Align Containment Spray pumps for I

soc

ID: LS'.21399····· .·· ·'·Points: 1.00

Unit-1 was operating at 100% power when a LOCA occurred. The following conditions exist:

• 12 HPSI Pump was OOS prior to the event • RCS pressure is 1400 PSIA and slowly lowering • Containment Pressure is 1.8 PSIG and slowly rising • EAST ECCS PP RM LVL HI alarm has annunciated on 1C10. • The ABO reports water level in the East ECCS Pp Room is approximately 10 inches and

rising and the source appears to be in the area of the LPSI pump. • 11 RWT LVL I TEMP Alarm has annunciated on 1 C09

Which ONE of the following (1) actions must be taken to address these conditions and (2) what impact will these actions have on the performance of the Emergency Core Cooling System?

A. (1) Place both LPSI pump, both HPSI pump and both Containment Spray Pump handswitches in PTL and shut both RWT Outlets; (2) ECCS flow is lost, SIG heat removal remains sufficient.

B. ( 1) Place 11 LPSI pump, 11 HPSI pump, and 11 Containment Spray Pump handswitches in PTL and shut the associated RWT Outlet; (2) ECCS flow is reduced to approximately one-half, heat removal capability is inadequate.

C. (1) Place both LPSI pump, both HPSI pump and both Containment Spray Pump handswitches in PTL and shut both RWT Outlets; (2) ECCS flow is lost, SIG heat removal capability is inadequate.

D. (1) Place 11 LPSI pump, 11 HPSI pump, 11 Containment Spray Pump handswitches in PTL and shut the associated RWT Outlet; (2) ECCS flow is reduced to approximately one-half, heat removal capability remains sufficient.

Answer: D


OPERATIONS


Answer Explanation:

A. Incorrect 1) False. Placing both trains of SI in PTL is plausible if Operator only assesses condition as lowering RWT level and does not determine that leak is in East ECCS Pump Room. Level provided in stem for ECCS Pump room indicates RWT still has sufficient level to support unaffected train. Securing ALL ECCS Pumps would be a wrong choice. 2) False. No ECCS flow is plausible if Operator places all SI in PTL, which would be an incorrect action.

B. Incorrect 1) True. Placing one train of SI in PTL to isolate the leak in the East ECCS Pump room is correct action. 2) False. Inadequate heat removal is plausible since HPSI and LPSI graphs have 2 pumps on them. However, heat removal is adequate as capability of one SI train meets design criteria.

C. Incorrect 1) False-Placing both trains of SI in PTL is plausible if Operator only assesses condition as lowering RWT level and does not determine that leak is in East ECCS Pump Room. Level provided in stem for ECCS Pump room indicates RWT still has sufficient level to support unaffected train. Securing ALL ECCS Pumps would be a wrong choice. 2) False. No ECCS flow is plausible if Operator places all SI in PTL, which would be an incorrect action. Given conditions indicate that SIG heat removal is adequate.

D.Correct 1) True. Placing one train of SI in PTL to isolate the leak in the East ECCS Pump room is correct action. With given indications the leak is from the RWT (low level alarm with RCS pressure still above pump shutoff head and Containment pressure below CSAS actuation). Pumps taking suction from the affected RWT suction header need to be secured to prevent damage. RWT outlet needs to be shut, to isolate the leak. 2) True. Heat removal is adequate as capability of one SI train meets design criteria.

Page: 102 of 235 13 April 2016




Topic: Loss of ECCS flowpath RO Importance: 2.8/3.9 SRO Importance: 3.1/4.3 Cognitive level (High or


Tier/Group 2/1

KIA Info 006 Emergency Core Cooling System (ECCS) K6 Knowledge of the effect of a loss or malfunction on the following will have on the ECCS: K6.13 Pumps

RO 2.8 Importance

Technical 1C09-ALM, 1C10-ALM, AOP-68 References


Learning Identify the core heat removal mechanisms Objective for a large or small break LOCA.





Comments None



Task 201.131 Recognizes/investigates degradation of ECCS and shuts RWT outlet (Sl-4143)

Additional KIA Match Info 006 Emergency Core Cooling System (ECCS) A2 Ability to (a) predict the impacts of the following malfunctions or operations on the ECCS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: A2.02 Loss of flow path



33 ID: Q106517 . Points:,1.00

Which ONE of the following could cause a "13 HPSI PP DISCH PRESS HI" alarm on 1C09?

OPERATIONS

A. A loss of MCC-104 occurs followed by a Large Break LOCA.

B. The HPSI Aux Header Isolation, Sl-646-MOV, is shut during a Large Break LOCA.

C. With RCS pressure at 1000 PSIA, the Mini-Flow MOV-660 is shut during a Small Break LOCA.

D. With RCS pressure at 1000 PSIA, both 12 HPSI and 13 HPSI are operating during a Small Break LOCA.

Answer: A

Answer Explanation:

A. Correct-A loss of MCC-104 will cause a loss of power to the Main HPSI Header MOVs. During a LBLOCA, SIAS will actuate and 13 HPSI will start. Regardless of RCS pressure, the 13 HPSI will not have a discharge flowpath other than the mini-flow valves, and this will result in 1 C09 alarm for 13 HPSI PP DISCH PRESS HI.

B. Incorrect-During a LBLOCA, SIAS will actuate and 13 HPSI will start. With a RCS pressure of 1000 PSIA, HPSI flow into the core will be greater than 350 GPM per operating pump per EOP Attachment 10. 13 HPSI PP DISCH PRESS HI is actuated if the discharge flow path is isolated or the pump is operating at minimum flow conditions. A flowrate of >350 GPM is well above minimum flow. Shutting MOV-656 will not isolate 13 HPSI flow, but 11 HPSI flow. Plausible if Operator considers the Aux Train to be the B ECCS Train.

C. Incorrect-During a SBLOCA, SIAS may have actuated or may have been blocked. Regardless, EOP-5 directs that 13 HPSI to be verified running. With a RCS pressure of 1000 PSIA, HPSI flow into the core will be greater than 350 GPM per operating pump per EOP Attachment 10. 13 HPSI PP DISCH PRESS HI is actuated if the discharge flow path is isolated or the pump is operating at minimum flow conditions. A flowrate of >350 GPM is well above minimum flow. Shutting MOV-660 will isolate mini-flow back to the RWT, but this will not lower flow appreciably since the mini-flow lines pass only 30 GPM per HPSI pump back to the RWT. Plausible if Operator believes HPSI flow at 1000 PSIA is negligible and shutting MOV-660 will place the pump in minimum flow conditions.

D. Incorrect-During a SBLOCA, SIAS may have actuated or may have been blocked. Regardless, EOP-5 directs that 13 HPSI to be verified running. With a RCS pressure of 1000 PSIA, HPSI flow into the core will be greater than 350 GPM per operating pump per EOP Attachment 10. 13 HPSI PP DISCH PRESS HI is actuated if the discharge flow path is isolated or the pump is operating at minimum flow conditions. A flowrate of >350 GPM is well above minimum flow. Starting 12 HPSI is directed in EOP-5 only if 11 HPSI is not operating, and when 12 HPSI is started, it will discharge to the AUX HPSI header, not the Main HPSI header that 13 HPSI is aligned to. Plausible if Operator believes 12 HPSI is normally aligned to Main HPSI header and increased flow resistance in Main HPSI Header from two pumps operating in one header will be large enough to cause minimum flow conditions.

Page: 105 of 235 13 April 2016




Topic: Cause of 13 HPSI DISCH PRESS HI Alarm RO Importance: 4.2 SRO Importance: 4.1 Cognitive level (High or


Tier/Group 2/1

KIA Info 006 Emergency Core Cooling System (ECCS) 2.4 Emergency Procedures/Plan 2.4.31 Knowledge of annunciator alarms, indications, or response procedures

RO 4.2 Importance

Technical 1C09 ALM, EOP-5, AOP-71 References


Learning Given RCS parameters identify appropriate Objective Safety Injection system response Question New Source




Comments None



Task 201.059 Monitor RCS Depressurization


34·


ID: Q106518 Points: 1.00

The following transient has occurred:

• A Large Break LOCA • Non-condensable gases have accumulated in the Reactor Vessel Head and the PZR • RCS pressure control has become erratic because of the voids

Which ONE of the following will occur while venting the RCS using 01-1 G?

A. When venting the PZR, PZR level will lower and Quench Tank pressure will rise.

B. When venting the PZR, PZR level will rise and Quench Tank pressure will rise.

C. When venting the RX Vessel Head, PZR level will lower and Reactor Coolant Drain Tank pressure will rise.

D. When venting the RX Vessel Head, PZR level will rise and Reactor Coolant Drain Tank pressure will rise.

Answer: B

Answer Explanation:

A. Incorrect-When voids are removed from the PZR steam space, PZR level will increase. The PZR vent valves discharge to the QT and pressure will rise while the PZR gases are vented to the QT. Plausible if operator confuses RVH void response with that of the PZR. When the RVH void is reduced, PZR level will lower.

B. Correct-When voids are removed from the PZR steam space, PZR level will increase. The PZR vent valves discharge to the QT and pressure will rise while the PZR gases are vented to the QT.

C. Incorrect-When voids are removed from the RVH, PZR level will lower. The RVH vent valves discharge to the QT and pressure will rise while the RVH gases are vented to the QT. Plausible if operator confuses QT and RCDT response during venting for one of two reasons. During normal operations, the QT can be manually drained to the RCDT. If Operator believes that manual drain on QT is opened to the RCDT while venting the RVH, then RCDT pressure will rise. Operator could also believe that RVH vent path discharges directly to RCDT as there is a sample line connection on RVH vent discharge piping RVH vent, although there is a checkvalve to prevent flow to the RCDT.

D. Incorrect-When voids are removed from the RVH, PZR level will lower. Plausible if operator confuses RVH void response with that of the PZR. When the PZR void is reduced, PZR level will rise. The RVH vent valves discharge to the QT and pressure will rise while the RVH gases are vented to the QT. Plausible if operator confuses QT and RCDT response during venting for one of two reasons. During normal operations, the QT can be manually drained to the RCDT. If Operator believes that manual drain on QT is opened to the RCDT while venting the RVH, then RCDT pressure will rise. Operator could also believe that RVH vent path discharges directly to RCDT as there is a sample line connection on RVH vent discharge piping RVH vent, although there is a checkvalve to prevent flow to the RCDT.




Svstem ID: 106518 User-Defined ID: Q106518 Cross Reference Number:

Topic: PZR Venting impact on QT RO Importance: 3.0 SRO Importance: 3.2 Cognitive level (High or


Tier/Group 2/1

KIA Info 007-Pressurizer Relief Tank/Quench Tank System (PRTS) K1 Knowledge of the physical connections and/or cause effect relationships between the PRTS and the following systems: K1.03 RCS

RO 3.0 Importance

Technical 01-1 G, 60734SH0001, 60729SH0001 References


Learning State the basis for the Pressurizer and Objective Reactor Vessel vent valves

Question New Source




Comments None



!Task I 064.004 Vent the RCS using various methods I

OPERATIONS Page: 11 O of 235 13 April 2016


ID: Q106519 Points: 1.00

Unit-2 is operating at 100% when:

• Loss of Off-Site Power occurs • Component Cooling Water leak develops in 21 CCHX

Which of the following are available as make-up sources to the Component Cooling Head Tank once the leak is isolated?

OPERATIONS

A. SRW, DI, Condensate, and Fire System.

B. DI, Condensate, and Fire System ONLY.

C. Condensate and Fire System ONLY.

D. Fire System ONLY.

Answer: D

Answer Explanation:

A. Incorrect-On a LOOP, the normal makeup for the CC Head Tank, DI, will be unavailable due to loss of power. The only source of pressurized water initially is from the Fire Main (via the diesel powered Fire Pump.) SRW is plausible if Operator believes SRW, which would be pressurized, can be used via common piping between the DI system and SRW/CCW to fill the CC Head Tank. A check valve prevents flow from the SRW system to the CC Head Tank. DI is plausible if Operator believes the DI Storage tank volume can be gravity fed to the CC Head Tanks, but due to elevation issues, this is not a possible source. Condensate is plausible since the condensate system piping is utilized to provide a flowpath from the fire hose station and Operator could believe that condensate is source of water.

B. Incorrect-On a LOOP, the normal makeup for the CC Head Tank, DI, will be unavailable due to loss of power. The only source of pressurized water initially is from the Fire Main (via the diesel powered Fire Pump.) DI is plausible if Operator believes the DI Storage tank volume can be gravity fed to the CC Head Tanks, but due to elevation issues, this is not a possible source. Condensate is plausible since the condensate system piping is utilized to provide a flowpath from the fire hose station and Operator could believe that condensate is source of water.

C. Incorrect-On a LOOP, the normal makeup for the CC Head Tank, DI, will be unavailable due to loss of power. The only source of pressurized water initially is from the Fire Main (via the diesel powered Fire Pump.) Plausible since the condensate system piping is utilized to provide a flowpath from the fire hose station and Operator could believe that condensate is source of water.

D. Correct-On a LOOP, the normal makeup for the CC Head Tank, DI, will be unavailable due to loss of power. The only source of pressurized water initially is from the Fire Main (via the diesel powered Fire Pump.)

Page: 111 of 235 13 April 2016




Topic: Makeup Sources to CC Head Tank RO Importance: 3.0 SRO Importance: 3.1 Cognitive level (High or


Tier/Group 2/1

KIA Info 008 Component Cooling Water System (CCWS) K1 Knowledge of the physical connections and/or cause-effect relationships between the CCWS and the following systems: K1 .05 Sources of makeup water

RO 3.0 Importance

Technical 01-16, 62710SH0001, AOP-7C References


Learning Given a Component Cooling malfunction, Objective diagnose the event and take the appropriate

action per AOP-7C.

Question New Source






Comments None

Task 202.068 Respond to loss of Component Cooling during Modes 3, 4, 5 or 6


36


ID: Q106520 Points: 1.00

Unit-2 is in Mode 4 with the following conditions:

• Shutdown Cooling (SDC) has just been initiated • RCS temperature is 245°F and slowly lowering • The Cooldown Rate (CDR) is 20°F/hr • Additional cooling is desired • Direction has been given to unlock and throttle open the 21/22 SDC HX CC INLET VLVs

Which ONE of the following choices both indicate the SDC HX CC INLET VLVs have been opened too far?

A. SDC HX(S) DIFF PRESS LO alarm is received and CDR is 105°F/hr.

8. SDC HX(S) DIFF PRESS LO alarm is received and CDR is 45°F/hr.

C. SDC HX(S) DIFF PRESS HI alarm is received and CDR is 105°F/hr.

D. SDC HX(S) DIFF PRESS HI alarm is received and CDR is 45°F/hr.

Answer: C

Answer Explanation:

A. Incorrect-When throttling open the CC inlet valve to the SDC HX, the higher flow rate will result in a higher D/P across the SDC HX. 01-38 directs that the D/P be maintained less than 8 PSID (the alarm limit.) Lower pressure is plausible if Operator confuses increased flowrate and believes a lowering CC header pressure will also result in a lowering D/P across the SDC HX, when in fact the D/P will increase. On U-2, the TS limit for CDR is <100°F/hr down to a temperature of 146°F. With the RCS at 245°F, any CDR above <100°F/hr means the CC valve has been opened too far.

8. Incorrect-When throttling open the CC inlet valve to the SDC HX, the higher flow rate will result in a higher D/P across the SDC HX. 01-38 directs that the D/P be maintained less than 8 PSID (the alarm limit.) Lower pressure is plausible if Operator confuses increased flowrate and believes a lowering CC header pressure will also result in a lowering D/P across the SDC HX, when in fact the D/P will increase. On U-2, the TS limit for CDR is <100°F/hr down to a temperature of 146°F. With the RCS at 245°F, any CDR above <100°F/hr means the CC valve has been opened too far. Plausible if Operator confused U-1 and U-2 CDR limits since U-1 limit is 40°F/hr at 245°F.

C. Correct-When throttling open the CC inlet valve to the SDC HX, the higher flow rate will result in a higher DIP across the SDC HX. 01-38 directs that the D/P be maintained less than 8 PSID (the alarm limit.) On U-2, the TS limit for CDR is <100°F/hr down to a temperature of 146°F. With the RCS at 245°F, any CDR above <100°F/hr means the CC valve has been opened too far.

D. Incorrect-When throttling open the CC inlet valve to the SDC HX, the higher flow rate will result in a higher D/P across the SDC HX. 01-38 directs that the D/P be maintained less than 8 PSID (the alarm limit.) On U-2, the TS limit for CDR is <100°F/hr down to a temperature of 146°F. With the RCS at 245°F, any CDR above <100°F/hr means the CC valve has been opened too far. Plausible if Operator confused U-1 and U-2 CDR limits since U-1 limit is 40°F/hr at 245°F.





Topic: Monitorinq the Throttlinq of CCW to SDC HXs RO Importance: 2.5 SRO Importance: 2.5 Cognitive level (High or


Tier/Group 2/1

KIA Info 008-Component Cooling Water System (CCWS) A4 Ability to manually operate and/or monitor in the control room: A4.06 Remote operation of hand-operated throttle valves to regulate CCW flow rate

RO 2.5 Importance

Technical 2C09 ALM, 01-38, TS 3.4.3, 62710SH0002 References


Learning Apply the Precautions, Initial Conditions, and Objective Cautions of 01-38

Question New Source






Comments None

Task 052/061.001 Initiate shutdown cooling


37


ID: Q106521 Points: 1.00

Unit-1 is operating at 100% when the following transient occurs:

• Loss of Offsite Power • 1A Diesel Generator fails to start • PZR level has recovered to >101" following the trip • No electrical buses have been tied yet in EOP-0

Which of the following is the correct PZR heater status when commencing RCS Pressure Control in the EOPs?

A 11 /12 Proportional and 11 /12/13/14 Backup Heaters are all available.

B. ONLY 12 Proportional and 13 Backup Heater are available.

C. ONLY 13 Backup Heater is available.

D. ONLY 12 Proportional is available.

Answer: B

Answer Explanation:

A Incorrect-On a LOOP concurrent with a loss of the 1A DG, only 14 4Kv bus will be reenergized. This will provide power to only 12 Prop and 13 BU heaters. PZR level will lower below 101" and then recover. Once PZR level is >101", the Prop heaters must be reset at 1 C06 before they will energize. 13 BU breaker will load shed on UV. The breaker must be manually charged and then closed locally to reenergize the BU heater. Plausible if Operator concentrates on fact that PZR level is above 101 ", clearing low level cutout to all heaters, but forgets LOOP effects.

B. Correct-On a LOOP concurrent with a loss of the 1A DG, only 14 4Kv bus will be reenergized. This will provide power to only 12 Prop and 13 BU heaters. PZR level will lower below 101" and then recover. Once PZR level is >101", the Prop heaters must be reset at 1 C06 before they will energize. 13 BU breaker will load shed on UV. The breaker must be manually charged and then closed locally to reenergize the BU heater.

C. Incorrect-On a LOOP concurrent with a loss of the 1A DG, only 14 4Kv bus will be reenergized. This will provide power to only 12 Prop and 13 BU heaters. PZR level will lower below 101" and then recover. Once PZR level is >101", the Prop heaters must be reset at 1 C06 before they will energize. 13 BU breaker will load shed on UV. The breaker must be manually charged and then closed locally to reenergize the BU heater. Plausible if Operator confuses impacts from a loss of 1Y10 ( 14 bus) with loss of 1 Y09 ( 11 bus.)

D. Incorrect-On a LOOP concurrent with a loss of the 1A DG, only 14 4Kv bus will be reenergized. This will provide power to only 12 Prop and 13 BU heaters. PZR level will lower below 101" and then recover. Once PZR level is >101", the Prop heaters must be reset at 1 C06 before they will energize. 13 BU breaker will load shed on UV. The breaker must be manually charged and then closed locally to reenergize the BU heater. Plausible if Operator believes 13 BU is powered from a 13 480V bus.





Topic: PZR Heater control following LOOP and 1A DG RO Importance: 3.0 SRO Importance: 3.4 Cognitive level (High or


Tier/Group 2/1

KIA Info 010 Pressurizer Pressure Control K2 Knowledge of bus power supplies to the following: K2.01 Pzr Heaters

RO 3.0 Importance

Technical AOP-71, EOP-0, EOP-2 References


Learning Given a Loss of Forced Circulation/Loss of Objective Offsite Power the trainee will be able to

identify, understand the basis and take appropriate actions per EOP-2 to mitigate the event






38


Comments Modified from 014487. 014487 used on 2012 NRC RO Exam.

Task 201.030 Verify RCS pressure and temperature within limits of EOP Att (1)

ID: Q106522 ·


• Plant trips due to a Loss of Load.

The following conditions exist in EOP-0:

• A SG Safety Valve on 22 SG is stuck open • 22 SG Safety Valve reseats • 21/22 SG pressures are both 660 PSIA when the cooldown terminates • RCS temperature is 500°F when the cooldown terminates

Which ONE of the following actions should be taken in EOP-0 Heat Removal and why?

A Maintain Tcold -500°F to prevent a rapid increase in RCS pressure.

B. Maintain Tcold -500°F to ensure the RCS temperature is within 25°F of the SGs.

C. Control Tcold -525°F to to prevent a rapid increase in RCS pressure.

D. Control Tcold -525°F to ensure the RCS temperature is within 25°F of the SGs.

Answer: A


OPERATIONS


Answer Explanation:

A. Correct-With a SG pressure of< 685 PSIA, two things will happen when performing HR in EOP-0. The MSIVs will be shut once < 800 PSIA and SGIS will be verified once < 685 PSIA. If SGIS has actuated and the cooldown terminates (which it has since the SG safety has reseated), HR directs that Tcold be stabilized. The EOP-0 bases states the reason for stabilizing temperature is pressure control. Initially, the insurge of water will result in a rapid increase in RCS pressure.

B. Incorrect-With a SG pressure of< 685 PSIA, two things will happen when performing HR in EOP-0. The MSIVs will be shut once < 800 PSIA and SGIS will be verified once < 685 PSIA. If SGIS has actuated and the cooldown terminates (which it has since the SG safety has reseated), HR directs that Tcold be stabilized. The EOP-0 bases states the reason for stabilizing temperature is pressure control. Initially, the insurge of water will result in a rapid increase in RCS pressure. Ensuring RCS temperature is maintained within 25F of the SG is plausible since these are parameters in EOP-4 that are monitored during the uncontrolled cooldown.

C. Incorrect-With a SG pressure of< 685 PSIA, two things will happen when performing HR in EOP-0. The MSIVs will be shut once < 800 PSIA and SGIS will be verified once < 685 PSIA. If SGIS has actuated and the cooldown terminates (which it has since the SG safety has reseated), HR directs that Tcold be stabilized. Controlling Tcold at 525F is plausible since the EOP-0 temperature band is normally 525-535F. The EOP-0 bases states the reason for stabilizing temperature is pressure control. Initially, the insurge of water will result in a rapid increase in RCS pressure.

D. Incorrect-With a SG pressure of< 685 PSIA, two things will happen when performing HR in EOP-0. The MSIVs will be shut once < 800 PSIA and SGIS will be verified once < 685 PSIA. If SGIS has actuated and the cooldown terminates (which it has since the SG safety has reseated), HR directs that Tcold be stabilized. Controlling Tcold at 525F is plausible since the EOP-0 temperature band is normally 525-535F. The EOP-0 bases states the reason for stabilizing temperature is pressure control. Initially, the insurge of water will result in a rapid increase in RCS pressure. Ensuring RCS temperature is maintained within 25F of the SG is plausible since these are parameters in EOP-4 that are monitored during the uncontrolled cooldown.

Page: 120 of 235 13 April 2016




Topic: Temperature/Pressure Control followinq ESDE in EOP-0 RO Importance: 3.3 SRO Importance: 4.0 Cognitive level (High or


Tier/Group 2/1

KIA Info 010-Pressurizer Pressure Control System (PZR PCS) 2.4 Emergency Procedures I Plan 2.4.18 Knowledge of the specific bases for EOPs

SRO 3.3 Importance

Technical EOP-0, EOP-0 Tech Bases References


Learning Recognize the basis for the following Objective parameters operating criteria established in

EOP-0: • Pressurizer Pressure • Pressurizer Level • RCS subcooling • RCS Temperature (Tcold} • SG pressures and level

Question New Source






Comments Considered KJA match since PCS will not prevent Pressurized Thermal Shock if a rapid temperature rise is allowed

Task 201.015 Verify the HR Safety Function is satisfied


39.


ID: Q106523 · Roints: 1.00.

Which ONE of the following is the 1) lowest Containment Pressure that will trip the Reactor and 2) the bases for the Containment Pressure trip?

A. 1) 2.0 PSIG. 2) Feedwater Line Break.

B. 1) 2.0 PSIG. 2) RCP Siezed Rotor.

C. 1) 2.4 PSIG. 2) Main Steam Line Break.

D. 1) 2.4 PSIG. 2) Steam Generator Tube Rupture.

Answer: C

Answer Explanation:

A. Incorrect 1) False. 2.0 PSIG is plausible since the CNTMT PRESS HIGH PRE-TRIP alarm comes in at 1.6 PSIG. The actual trip, however, is at 2.4 PSIG. 2) True. Per the TS bases, one of the bases for the Containment High Pressure trip is due to a Feedwater Line Break.

B. Incorrect 1) False. 2.0 PSIG is plausible since the CNTMT PRESS HIGH PRE-TRIP alarm comes in at 1.6 PSIG. The actual trip, however, is at 2.4 PSIG. 2) False. The RCP siezed rotor is plausible if the Operator determines a LOCA could will occur due to a siezed rotor as a result of aux impellar flow. However, the locked rotor is only listed as a bases for the the Low Flow trip.

C. Correct 1) True. Per the 1 C05 ALR, Containment pressure will trip the reactor at a value of 2.4 PSIG. 2) True. Per the TS bases, one of the bases for the Containment High Pressure trip is due to a Main Steam Line Break.

D. Incorrect 1) True. Per the 1 C05 ALR, Containment pressure will trip the reactor at a value of 2.4 PSIG. 2) False. A SGTR is plausible since its impact on the RCS is similar to a LOCA, and a LOCA is listed as a bases for the Containment Trip. However, a SGTR should not challenge the Containment safety function/parameters.





''

Topic: Match Containment RPS Trip with its basis RO Importance: 3.9 SRO Importance: 4.3 Cognitive level (High or


Tier/Group 2/1

KIA Info 012 Reactor Protection System K4 Knowledge of RPS design feature(s) and/or interlock(s) which provide for the following: K4.02 Automatic reactor trip when RPS setpoints are exceeded for each RPS function; basis for each

RO 3.9 Importance

Technical TS Bases, 1 C05 ALM References


Learning Recall the Automatic RPS Reactor Trips and Objective for Each Trip Function[SOER 83-8]:

• Basis for the Trip • Trip Setpoint as per Limiting Safety

System Settings • Whether Automatic or Manual Bypass

Exists (K6.01, K11.05,06)

Question New Source




40



Comments None

Task 201.010 Determine if the reactor should have tripped automatically but did not (ATWS)

ID: Q16704

The following conditions have existed for approximately one minute:

• Unit-1 Reactor has tripped • 1-Pl-1013A thru D indicate 890 PSIA (11 SIG pressure) • 1-Pl-1023A thru D indicate 750 PSIA (12 SIG pressure) • 1-Ll-1114C & 1-LR-11140 indicate -180 inches (11 SIG level) • 1-Ll-1124C & 1-LR-11240 indicate -210 inches (12 SIG level)

Which ONE of the following is the status of the Aux Feedwater system?

A. 11 SIG-AFAS actuated with normal AFW flow restoring SIG level; 12 SIG-AFAS actuated with AFAS Block signal isolating AFW flow.

B. 11 SIG-AFAS actuated with AFAS Block signal isolating AFW flow; 12 SIG-AFAS actuated with AFAS Block signal isolating AFW flow.

C. 11 SIG-AFAS actuated with AFAS Block signal isolating AFW flow; 12 SIG-AFAS actuated with normal AFW flow restoring SIG level.

D. 11 SIG-AFAS actuated with normal AFW flow restoring SIG level; 12 SIG-AF AS actuated with normal AFW flow restoring SIG level.

Answer: A


OPERATIONS

EXAMINATION ANSWER KEY LOI 2015 NRG RO Exam (June2016)

Answer Explanation:

A. Correct 11 SG-True. AFAS actuated to both SIGs. 12 SG-True. AFAS Block is present isolating AFW flow to 12 SIG which is at a lower pressure.

B. Incorrect 11 SG-False. Plausible if Operator determines AFAS Block applies to both SGs since AFAS applies to both SIGs. However, AFAS Block occurs at 115 PSID to SIG at lowest pressure and isolates AFW flow to 12 SIG, therefore, AFAS Block cannot be present to both SIGs simultaneously. 12 SG-True. AFAS Block is present isolating AFW flow to 12 SIG which is at a lower pressure.

C. Incorrect 11 SG-False. Plausible if Operator determines AFAS Block applies to both SGs since AFAS applies to both SIGs. However, AFAS Block occurs at 115 PSID to SIG at lowest pressure and isolates AFW flow to 12 SIG, therefore, AFAS Block cannot be present to both SIGs simultaneously. 12 SG-False. Plausible if Operator believes block occurs to SIG with higher pressure. However, AFAS Block is present isolating AFW flow to 12 SIG which is at a lower pressure.

D. Incorrect 11 SG-True. AFAS actuated to both SIGs. 12 SG-False. Plausible since AFAS has initiated for both SIGs. However, 12 SIG pressure is< 11 SIG pressure by 140 PSIA. AFAS Block occurs at 115 PSID to SIG at lowest pressure and isolates AFW flow to 12 SIG.

Page 126 of 235 13 April 2016



' System ID: 16704 User-Defined ID: Q16704 Cross Reference Number:

Topic: Post-Trip AFW status RO Importance: 4.3 SRO Importance: 4.5 Cognitive level (High or


KIA Info 013 Engineered Safety Features Actuation Signal (ESFAS) K4 Knowledge of ESFAS design feature(s) and/or interlock(s) which provide for the following: K4.04 Auxiliary feed actuation signal

RO 4.3 Importance

Technical System Description 36(AFW), 1 C03 ALM058 References


Learning Evaluate plant conditions and/or parameters Objective and determine the following:

• ESFAS/AFAS signal status • Conditions required to reset an

ESFAS/AFAS signal either remotely (control room) or locally (CSRs)

• ESFAS/AFAS component status • Operator action required





OPERATIONS



Comments None

Task 036.022/048.015 Test AFAS "A" and "B" logic




Licensed Operator Requa! Training (LOR)

System Designations

Auxiliary Feedwater

Emergency Operating Procedures (EOPs)

Engineered Safety Features Actuation (ESFAS)

Cognitive Level

COMPREHENSION

Page: 128 of 235 13 April 2016

41


ID: Q106533 Points: 1.00

Unit-2 was operating at 100% power with the following conditions:

• 22 Containment Spray (CS) Pump is OOS • 21, 22, and 23 Containment Air Coolers (CACs) are running in HIGH speed • 23 CAC trips on overload

Several minutes after the loss of 23 CAC, the following transient occurs:

• Loss of P-13000-2 • Small-break LOCA • Containment Pressure is 0.4 PSIG and slowly rising • Containment Temperature is 125°F and slowly rising • RCS Pressure is 1785 PSIA and slowly lowering

Which ONE of the following statements correctly describes (1) the impact on the Containment Cooling System when 23 CAC first tripped, and (2) the actions required for CAC alignment in EOP-0?

A. (1) Both Containment Cooling trains remained operable as the remaining CACs could still adequately cool the containment in a Design Basis Accident.

(2) Verify 21, 22, and 24 CAC have shifted to LOW speed.

B. (1) One Containment Cooling train was inoperable but the combination of CACs and CS could still adequately cool the containment in a Design Basis Accident.

(2) Verify 21, 22, and 24 CAC have shifted to LOW speed.

C. (1) Both Containment Cooling trains remained operable as the remaining CACs could still adequately cool the containment in a Design Basis Accident.

(2) Verify 21, 22, and 24 CAC are running in HIGH speed.

D. ( 1) One Containment Cooling train was inoperable but the combination of CACs and CS could still adequately cool the containment in a Design Basis Accident.

(2) Verify 21, 22, and 24 CAC are running in HIGH speed.

Answer: D


OPERATIONS


Answer Explanation:

A. Incorrect-When 23 CAC first trips, there is a loss of the 'B' train of containment cooling. Per TS 3.6.6, a containment cooling train consists of 2 CACs, and thus the 'B' train will be OOS. Per the TS bases, Containment cooling can be maintained during a DBA if at least 1 CS pump and 2 CACs are available or 3 CACs are available. The Operator determining that the Containment Cooling train remained operable is plausible if they only consider the combination of CS and CACs remaining and forgetting that a train is comprised of two CACs. After the transient occurred, the correct actions in EOP-0 would have been to start all available CACs in HIGH. EOP-0 step ensures that all CACs are operating if temperature is >120F. Since a SIAS setpoint has not reached, the CACs should be run in HIGH speed. The 24 CAC will have to be manually started since it initially lost power on the loss of P-13000-2, and the 2B/SDS will not start CACs. 21 and 22 CAC should remain running throughout the transient. Verifying the CACs in SLOW is plausible if the Operator confuses 1785 PSIA (the PZR Block Permitted alarm for blocking SIAS) with the actual SIAS setpoint of 17 40 PSIA.

B. Incorrect-When 23 CAC first trips, there is a loss of the 'B' train of containment cooling. Per TS 3.6.6, a containment cooling train consists of 2 CACs, and thus the 'B' train will be OOS. Per the TS bases, Containment cooling can be maintained during a DBA if at least 1 CS pump and 2 CACs are available or 3 CACs are available. After the transient occurred, the correct actions in EOP-0 would have been to start all available CACs in HIGH. EOP-0 step ensures that all CACs are operating if temperature is >120F. Since a SIAS setpoint has not reached, the CACs should be run in HIGH speed. The 24 CAC will have to be manually started since it initially lost power on the loss of P-13000-2, and the 2B/SDS will not start CACs. 21 and 22 CAC should remain running throughout the transient. Verifying the CACs in SLOW is plausible if the Operator confuses 1785 PSIA (the PZR Block Permitted alarm for blocking SIAS) with the actual SIAS setpoint of 1740 PSIA.

C. Incorrect-When 23 CAC first trips, there is a loss of the 'B' train of containment cooling. Per TS 3.6.6, a containment cooling train consists of 2 CACs, and thus the 'B' train will be OOS. Per the TS bases, Containment cooling can be maintained during a DBA if at least 1 CS pump and 2 CACs are available or 3 CACs are available. The Operator determining that the Containment Cooling train remained operable is plausible if they only consider the combination of CS and CACs remaining and forgetting that a train is comprised of two CACs. After the transient occurred, the correct actions in EOP-0 would have been to start all available CACs in HIGH. EOP-0 step ensures that all CACs are operating if temperature is >120F. Since a SIAS setpoint has not reached, the CACs should be run in HIGH speed. The 24 CAC will have to be manually started since it initially lost power on the loss of P-13000-2, and the 2B/SDS will not start CACs. 21and22 CAC should remain running throughout the transient.

D. Correct-When 23 CAC first trips, there is a loss of the 'B' train of containment cooling. Per TS 3.6.6, a containment cooling train consists of 2 CACs, and thus the 'B' train will be OOS. Per the TS bases, Containment cooling can be maintained during a DBA if at least 1 CS pump and 2 CACs are available or 3 CACs are available. After the transient occurred, the correct actions in EOP-0 would have been to start all available CACs in HIGH. EOP-0 step ensures that all CACs are operating if temperature is >120F. Since a SIAS setpoint has not reached, the CACs should be run in HIGH speed. The 24 CAC will have to be manually started since it initially lost power on the loss of P-13000-2, and the 2B/SDS will not start CACs. 21 and 22 CAC should remain running throughout the transient.

Page: 130 of 235 13 April 2016




Topic: Bases for shifting CACs to slow speed during a Containment pressurization event



Tier/Group 2/1

KIA Info 022 Containment Cooling System (CCS) A2 Ability to (a) predict the impacts of the following malfunctions or operations on the CCS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: A2.03 Fan motor thermal overload/high-speed operation

RO 2.6 Importance

Technical TS 3.6.6, TS 3.6.6 Bases, EOP-0 References


Learning Recall the bases for LCO 3.6.6, Containment Objective Spray and Cooling Systems

Question New Source




42



Comments None

Task 201.016 Verify the CE Safety Function is satisfied

Unit-1 has experienced a LOCA with the following conditions:

• Containment Pressure is 5 PSIG and slowly rising • RWT level is 37' and slowly lowering

Which ONE of the following conditions will result in a Containment Spray Pump shortly running with no flow?

A. SIAS B failure and 1-HS-2080A CIS OVERRIDE alarm on 1C09.

B. SIAS A failure and CNTMT IA ISOL 1-IA-2085-CV CLOSED alarm on 1C07.

C. CSAS A failure and both SI PP RECIRC LOCKOUT switches are taken to ON.

D. CSAS B failure and SI PPS RECIRC MOV 659 CLOSED RAS BLOCKED alarm on 1C09.

Answer: D


OPERATIONS


Answer Explanation:

A Incorrect-With the given initial conditions, SIAS and CSAS should have both actuated. Both CS pumps should start on SIAS and both CS pump discharge CV's open on a CSAS. CS pump cooling is supplied by minimum flowrates through the pump. A SIAS B failure means 11 CS Pump only is running, since 12 CS Pump will not get a start signal. 1-HS-2080A CIS OVERRIDE alarm is plausible if the Operator believies MOV-2080 will close when this HS is operated, and believes the loss of air to the CS discharge CV will not allow the valve to open. The 1-HS-2080A CIS OVERRIDE alarm means MOV-2080 is open and air is available to open the CS discharge CV, allowing normal flow through 11 CS pump.

B. Incorrect-With the given initial conditions, SIAS and CSAS should have both actuated. Both CS pumps should start on SIAS and both CS pump discharge CV's open on a CSAS. CS pump cooling is supplied by minimum flowrates through the pump. A SIAS B failure means 11 CS Pump only is running, since 12 CS Pump will not get a start signal. CNTMT IA ISOL 1-IA-2085-CV CLOSED alarm is plausible if the Operator believies the loss of air downstream of 2085 will not allow the discharge CV to open. The loss of air will actually fail the pump discharge CVs open, and 1-HS-2080A CIS OVERRIDE alarm means MOV-2080 is open and air is available to open the CS discharge CV, allowing normal flow through 12 CS pump.

C. Incorrect-With the given initial conditions, SIAS and CSAS should have both actuated. Both CS pumps should start on SIAS and both CS pump discharge CV's open on a CSAS. CS pump cooling is supplied by minimum flowrates through the pump. A CSAS A failure means 11 CS pump is running but the discharge CV is shut. Taking SI PP RECIRC LOCKOUT switches to ON is plausible if Operator believes that taking SI PP RECIRC LOCKOUT switches to ON will shut the mini-flow MOVs-659/660. Taking the SI PP RECIRC LOCKOUT switches to ON will enable the MOVs to shut when RAS occurs, but given the RWT at 37', which is well above RAS, the mini-flow MOVs will remain open. 11 CS pump will be running with the only flow to the RWT via the mini-flow line.

D. Correct-With the given initial conditions, SIAS and CSAS should have both actuated. Both CS pumps should start on SIAS and both CS pump discharge CV's open on a CSAS. CS pump cooling is supplied by minimum flowrates through the pump. A CSAS B failure means 12 CS pump is running but the discharge CV is shut. The SI PPS RECIRC MOV 659 CLOSED RAS BLOCKED alarm means MOV-659 is closed. This will eliminate the mini-flow path back to the RWT. Therefore, there will be no pump flow through 12 CS Pump.

Page: 133 of 235 13 April 2016




Topic: MonitorinQ CS Mini-Flow RO Importance: 2.7 SRO Importance: 3.0 Cognitive level (High or


Tier/Group 2/1

KIA Info 026 Containment Spray System (CSS) A 1 Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the CSS controls including: A1 .06 Containment spray pump cooling

RO 2.7 Importance

Technical 1 CO? ALM, 1 C09 ALM References


Learning Recall the required flow of a CS Pump upon Objective automatic initiation from ESFAS

Question New Source





EXAMl'NATION ANSWER KEY LOI 2015 NRC RO Exam (June 2016)

Comments Considered a KIA match since Containment Spray Pump cooling is supplied by flow

Task 048.028 Notes failure of Containment Spray and manually initiates



43 ID: Q106536 Points: 1.00

Which ONE of the following conditions will result in both MSIVs shutting on Unit-2?

OPERATIONS

A. SGIS-8 actuates on low SG Pressure.

8. SIAS has initiated due to high Containment Pressure.

C. N2 pressure on both MSIVs lowers below 1625 PSIG.

D. 21 SG PT-1013A fails low and 22 SG PT-10238 fails low.

Answer: A

Answer Explanation:

A. Correct-On a SGIS actuation, both MS IVs will shut, regardless of what logic channel actuates.

8. Incorrect-SIAS actuation is plausible if Operator believes high Containment Pressure to cause SIAS is same setpoint to initiate CSAS. SIAS and CIS will both actuate at same setpoint due to Containment Pressure, not CSAS.

C. Incorrect-A low N2 pressure is plausible since 1625 PSIG is the N2 pressure when the MSIV is declared inoperable. However, the MSIV would remain open.

D. Incorrect-A low SG pressure is plausible if the Operator believes that 214 conditions for low SG pressure have been met with 1013A and 10238. For SGIS to actuate, 214 PT on a SG must be lower than 703 PSIA. The conditions given are on separate SGs.

Question 43· f nfo Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 0 Difficulty: 0.00


Topic: Conditions that isolate MSIVs RO Importance: 3.7 SRO Importance: 3.7 Cognitive level (High or L Low): NRC KIA Info, References:

/Tier/Group / 2/1 I

Page: 136 of 235 13April2016


KIA Info 039 Main and Reheat Steam System (MRSS) K4 Knowledge of MRSS design feature(s) and/or interlock(s) which provide for the following: K4.05 Automatic isolation of steam line

RO 3.7 Importance

Technical 1C09 ALM, 61058ASH0001, Ol-8E References


Learning State the response of the MSIV upon receipt Objective of an SGIS/CSAS

Question New Source




Comments None

Task 048.031 Note failure of SGIS and manually initiates


44


ID: 0106537 Points~ 1.00


• 11 SGFP trips • Trip criteria have not yet been reached in AOP-3G

Which ONE of the following will 1) cause a trip of 11 SGFP and 2) immediately indicate that 11 SGFP has tripped?

A 1) Main Oil Pump A and Main Oil Pump B lose power 2) 11 SG FRV 1-FW-1111-CV is SHUT and 11 SG Bypass FRV Bypass 1-FW-1105-CV is 56% OPEN

B. 1) Suction Pressure on Ovation screen reads 215 PSIG for 30 secs 2) 11 SGFP Drain MOVs indications on 1 C03 are all OPEN

C. 1) Containment Pressure reaches 5 PSIG 2) 11 SGFP Mini-Flow 1-FW-4484-CV is SHUT

D. 1) Condenser Vacuum lowers to 23 In-Hg 2) 11 SGFPT TRIP alarm on 1C03

Answer: C

Answer Explanation:

A Incorrect 1) True. SGFP will trip if a both Main Oil Pumps lose power. 2) False. FRV going shut and the BFRV opening to 56% is plausible if Operator confuses feed response to a tripped SGFP with response to a Reactor trip.On a trip, the SGFP Mini-Flow CV will go shut and remain shut as long as the trip is active.

B. Incorrect 1) True. SGFPs will trip if SGFP suction pressure is <221 PSIG for >20 secs. 2) False. Drain MOVs opening is plausible if Operator confuses actions needed to take when shutting down a SGFP with those that automatically happen on a SGFP trip. The Drain MOVs would remain shut until Ol-12A was entered to reposition them for shutdown.

C. Correct 1) True. SGFPs will trip if a CSAS actuates. Since Containment Pressure is >4.25 PSIG, CSAS has actuated. 2) True. On a trip, the SGFP Mini-Flow CV will go shut and remain shut as long as the trip is active.

D. Incorrect 1) False. SGFPs will trip if vacuum is <20 In-Hg. 23" is plausible if Operator confuses manual reactor trip criteria of <23.5 In-Hg with SGFP automatic trip signal. 2) True. 11 SGFPT TRIP alarm would be received on the SGFP trip.





Topic: Monitoring SGFP trip indications RO Importance: 3.1 SRO Importance: 3.1 Cognitive level (High or

H Low): NRG KJA Info, References:

Tier/Group 2/1

K/A Info 059 Main Feedwater (MFW) System A4 Ability to manually operate and monitor in the control room: A4.01 MFW turbine trip indication

RO 3.1 Importance

Technical 1C03 ALM, AOP-3G, 61058ASH0001 References


Learning Identify the actions taken upon the failure of Objective any of the following pumps for ALL Reactor

Power Levels. • Steam Generator Feed Pump • Condensate Booster Pump • Condensate Pump • Heater Drain Pump

Question New Source






Comments Considered a High Cog question since Operator must evaluate conditions to determine a CSAS actuation has occurred, which will trip a SGFP.

Task 202.036 Respond to a failure of a pump with Reactor Power~ 5%

ID: Q10G5~9°:;' .·. Points: 1.00


• Loss of Offsite Power at 0200 • Loss of all AFW at 0210 • Once Through Core Cooling (OTCC) initiated at 0250 • Both SGs are isolated at 0255

Recovery is established with the following conditions:

• Power is restored to all 4Kv busses at 0310 • Main Feedwater (MFW) is available at 0325 • The Tech Support Center has directed that 21 SG remain isolated

Which ONE of the following are 1) the concerns with feeding 22 SG with MFW and 2) the EOP-3 actions required to minimize that impact?

OPERATIONS

A. 1) Waterhammer in the Main Feedwater System could occur. 2) Limit feed rate to 150 GPM for 5 minutes, then feed rate can be raised.

B. 1) Waterhammer in the Main Feedwater System could occur. 2) A 10-minute purge of the gooseneck must be performed first, then feed rate can be raised.

C. 1) Main Feedwater flowrates will be below SGFP mini-flow requirements. 2) Limit feed rate to 150 GPM for 5 minutes, then feed rate can be raised.

D. 1) Main Feedwater flowrates will be below SGFP mini-flow requirements. 2) A 10-minute purge of the gooseneck must be performed first, then feed rate can be raised.

Answer: B

Page: 140 of 235 13 April 2016

OPERATIONS


Answer Explanation:

A. Incorrect 1) True. Per EOP-3 bases, waterhammer is possible when restoring either AFW or MFW. Given the timeline of events, MFW was lost at 0200 due to loss of power. When MFW is restored at 0325, 85 minutes later, the feed ring has become partially drained. If feed flow is established with the feed ring empty, severe waterhammer is possible. 2) False. Plausible if Operator uses AFW restoration guidance of 5 minutes at 150 GPM instead of the MFW guidance to purge the gooseneck for 10 minutes. Per Ol-12A and EOP-3, if the RCS is >200F and there has been a loss of feed flow for >80 minutes, a purge of the gooseneck is required for at least 10 minutes before raising feed rate.

B. Correct 1) True. Per EOP-3 bases, waterhammer is possible when restoring either AFW or MFW. Given the timeline of events, MFW was lost at 0200 due to loss of power. When MFW is restored at 0325, 85 minutes later, the feed ring has become partially drained. If feed flow is established with the feed ring empty, severe waterhammer is possible. 2) True. Per Ol-12A and EOP-3, if the RCS is >200F and there has been a loss of feed flow for >80 minutes, a purge of the gooseneck is required for at least 10 minutes before raising feed rate.

C. Incorrect 1) False. Main Feedwater flowrates below SGFP mini-flow requirements is plausible since the initial feed flow to the SG will be minimal. However, when the MFW is restored, miniflow lines for the Condensate, CBP, and the SGFP will ensure mini-flow is maintained even before starting to feed the SGs. 2) False. Plausible if Operator uses AFW restoration guidance of 5 minutes at 150 GPM instead of the MFW guidance to purge the gooseneck for 10 minutes. Per Ol-12A and EOP-3, if the RCS is >200F and there has been a loss of feed flow for >80 minutes, a purge of the gooseneck is required for at least 10 minutes before raising feed rate.

D. Incorrect 1) False. Main Feedwater flowrates below SGFP mini-flow requirements is plausible since the initial feed flow to the SG will be minimal. However, when the MFW is restored, miniflow lines for the Condensate, CBP, and the SGFP will ensure mini-flow is maintained even before starting to feed the SGs. 2) True. Per Ol-12A and EOP-3, if the RCS is >200F and there has been a loss of feed flow for >80 minutes, a purge of the gooseneck is required for at least 1 O minutes before raising feed rate.

Page: 141 of 235 13 April 2016




Topic: Issues and Actions for Feedinq a Drv SG RO Importance: 2.9 SRO Importance: 3.4 Cognitive level (High or


Tier/Group 2/1

KIA Info 059 Main Feedwater (MFW) System A2 Ability to (a) predict the impacts of the following malfunctions or operations on the MFW; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: A2.04 Feeding a dry SIG

RO 2.9 Importance



Learning Recall operator actions to be taken when Objective transitioning from OTCC to heat removal via

steam generators

Question New Source






Comments None

Task 201.038 Evaluate feeding a Steam Generator


46


10:.Q106540 Points: 1.00


• 11 AFW Pump is in operation and feeding both SGs at 50 GPM per SG • Supply air to the 11 AFW Pump Speed Controller 1-HC-3987 positioner is lost • Supply air to the 11 S/G Flow Controller FIC-4511A positioner is lost

Which ONE of the following is the expected AFW response to the loss of air?

A 11 AFW Pump speed will lower and the AFW FCV to 11 S/G will fail open.

B. 11 AFW Pump speed will lower and the AFW FCV to 11 S/G will fail closed.

C. 11 AFW Pump speed will rise and the AFW FCV to 11 S/G will fail open.

D. 11 AFW Pump speed will rise and the AFW FCV to 11 S/G will fail closed.

Answer: C

Answer Explanation:

A Incorrect-Speed lowering is plausible if Operator believes control system is direct acting (i.e. air required to raise speed). On a loss of air, speed control will fail to maximum speed and the AFW pump speed will rise. The AFW FCVs fail open on a loss of air.

B. Incorrect-Speed lowering is plausible if Operator believes control system is direct acting (i.e. air required to raise speed). On a loss of air, speed control will fail to maximum speed and the AFW pump speed will rise. The FCVs failing closed is plausible if the Operator believes the FCVs are direct acting. However, the AFW FCVs fail open on a loss of air.

C. Correct-On a loss of air, speed control will fail to maximum speed and the AFW pump speed will rise. The AFW FCVs fail open on a loss of air.

D. Incorrect-On a loss of air, speed control will fail to maximum speed and the AFW pump speed will rise. The FCVs failing closed is plausible if the Operator believes the FCVs are direct acting. The AFW FCVs fail open on a loss of air.





Topic: AFW Impact on Loss of Controller Power RO Importance: 2.5 SRO Importance: 2.8 Cognitive level (High or


Tier/Group 2/1

K/A Info 061 Auxiliary I Emergency Feedwater (AFW) System K6 Knowledge of the effect of a loss or malfunction of the following will have on the AFW components: K6.01 Controllers and positioners

RO 2.5 Importance

Technical AOP-7J References


Learning Evaluate the effect on AFW system operation Objective of a loss of operating air to various AFW

system components

Question New Source






Comments None

Task 202.100 Respond to the loss of a 120 volt vital AC bus


47


ID: Q106542 Points: 1.00

A Loss of Offsite Power has occurred. SMECO is determined to be available.

Which ONE of the following combinations of 4Kv busses could be powered at the same time by SM ECO?

A. 11 and 13 4Kv busses

B. 12 and 21 4kV busses

C. 14 and 22 4kV busses

D. 21 and 24 4kV busses.

Answer: D

Answer Explanation:

A. Incorrect-Per Ol-27E, SMECO can supply 2 of the 4 SR 4Kv busses (11/14/21/24). The 01 does not provide guidance to use SMECO to energize a NSR 4Kv bus (12/13/22/23). Powering 13 4Kv bus is plausible if Operator believes SMECO aligned to 11/21 13Kv bus will allow alignment to a NSR 4Kv bus.

B. Incorrect-Per Ol-27E, SMECO can supply 2 of the 4 SR 4Kv busses (11/14/21/24). The 01 does not provide guidance to use SMECO to energize a NSR 4Kv bus (12/13/22/23). Powering 13 4Kv bus is plausible if Operator believes SM ECO aligned to 11 /21 13Kv bus will allow alignment to a NSR 4Kv bus.

C. Incorrect-Per Ol-27E, SMECO can supply 2 of the 4 SR 4Kv busses (11/14/21/24). The 01 does not provide guidance to use SMECO to energize a NSR 4Kv bus (12/13/22/23). Powering 13 4Kv bus is plausible if Operator believes SMECO aligned to 11/21 13Kv bus will allow alignment to a NSR 4Kv bus.

D. Correct-Per Ol-27E, SM ECO can supply 2 of the 4 SR 4Kv busses (11/14/21/24). The 01 does not provide guidance to use SMECO to energize a NSR 4Kv bus (12/13/22/23).





Topic: SMECO connections during LOOP RO Importance: 3.7 SRO Importance: 4.2 Cognitive level (High or


Tier/Group 2/1

KIA Info 062 AC. Electrical Distribution K1 Knowledge of the physical connections and/or cause-effect relationships between the ac distribution system and the following systems: K1 .04 Off-site power sources

RO 3.7 Importance

Technical Ol-27E References


Learning Describe the function and/or design Objective limitations of the following major systems and

controls: • 13KV transformers • SM ECO • Synchroscope • Voltage Regulating Transformers

Question New Source






Comments None

Task 201.075 Align electrical system for power restoration


48(


ID: Q106543


• A Station Blackout is in progress • EOP-7, Station Blackout, has been implemented • Both 11 and 12 Control Room HVAC are OOS

Which ONE of the following is the bases for ensuring the Plant Computer Inverter on Unit-2 is secured within 30 minutes into the blackout?

A Reduces load on 12 DC Bus which allows the DC bus to provide power for at least 4 hours.

B. Reduces load on 12 DC Bus which allows the DC bus to provide power for at least 7 hours.

C. Reduces load on 22 DC Bus which allows the DC bus to provide power for at least 4 hours.

D. Reduces load on 22 DC Bus which allows the DC bus to provide power for at least 7 hours.

Answer: C

Answer Explanation:

A Incorrect-Per EOP-7, removing the U-2 Plant Computer inverter from service will significantly reduce loading on the 22 DC bus. 12 DC bus is plausible if Operator confuses U-1 and U-2 Plant Computer inverter power supplies (U-1 is powered from 12 DC bus.) The Plant Computer inverter is expected to be downpowered within 30 minutes so the 22 battery will function for a minimum of 4 hours.

B. Incorrect-Per EOP-7, removing the U-2 Plant Computer inverter from service will significantly reduce loading on the 22 DC bus. 12 DC bus is plausible if Operator confuses U-1 and U-2 Plant Computer inverter power supplies (U-1 is powered from 12 DC bus.) The Plant Computer inverter is expected to be downpowered within 30 minutes so the 22 battery will function for a minimum of 4 hours. 7 hours is plausible if Operator confuses ELAP actions of FSG-4 which strip individual DC loads to extend battery life to 7 hours.

C. Correct-Per EOP-7, removing the U-2 Plant Computer inverter from service will significantly reduce loading on the 22 DC bus. The inverter is expected to be downpowered within 30 minutes so the 22 battery will function for a minimum of 4 hours.

D. Incorrect-Per EOP-7, removing the U-2 Plant Computer inverter from service will significantly reduce loading on the 22 DC bus. The Plant Computer inverter is expected to be downpowered within 30 minutes so the 22 battery will function for a minimum of 4 hours. 7 hours is plausible if Operator confuses ELAP actions of FSG-4 which strip individual DC loads to extend battery life to 7 hours.





..

Topic: Basis for shedding Computer Inverter load during a U-2 SBO



Tier/Group 2/1

KIA Info 063 DC Electrical Distribution A 1 Ability to predict and/or monitor changes in parameters associated with operating the DC electrical system controls including: A 1.01 Battery capacity as it is affected by discharge rate

RO 2.5 Importance

Technical System Description 058 (Reactor Protection References System)

References EOP-7, FSG-4 provided

Learning Given a Station Blackout, the license Objective candidate will be able to correctly recall and

identify the proper response and the basis to mitigate the event







Comments Modified from 095419. 095419 used in NRG 2012 RO Exam.

Task 201.075 Align electrical system for power restoration


49


ID: Q106544 Points:.~1.QO

The 2B Diesel Generator (DG) is being paralleled to the 24 4Kv bus with the following conditions:

• Incoming and Running voltages are matched • The Synchroscope is rotating slowly in the FAST direction • The 2B DG Output Breaker is closed when the Synchroscope is 5 degrees prior to the 12

o'clock position

Which ONE of the following is 1) the expected response when first closing the Output Breaker and 2) what action is required to increase Megawatt (MW) output to the desired level?

A. 1) Load should initially indicate 0.0 MW. 2) Raise load using Voltage Control.

B. 1) Load should initially indicate 0.0 MW. 2) Raise load using Speed Control.

C. 1) Load should initially indicate 0.5 MW. 2) Raise load using Voltage Control.

D. 1) Load should initially indicate 0.5 MW. 2) Raise load using Speed Control.

Answer: D

Answer Explanation:

A. lncorrect-1) False. 0.0 MW is plausible if Operator believes that having incoming and running voltages matched during the parallel equates to no load transfer initially. Per 01-21 B, when the DG output breaker is closed with the conditions provided, expected MW output is 0.5 MW as the DG picks up a small amount of load. 2) False. Voltage control is plausible if Operator confuses real load and apparent power, as increasing voltage will raise apparent power, but not MW loading. To further increase real load, speed control is raised.

B. lncorrect-1) False. 0.0 MW is plausible if Operator believes that having incoming and running voltages matched during the parallel equates to no load transfer initially. Per 01-21 B, when the DG output breaker is closed with the conditions provided, expected MW output is 0.5 MW as the DG picks up a small amount of load. 2) True. To further increase real load, speed control is raised.

C. lncorrect-1) True. Per 01-21 B, when the DG output breaker is closed with the conditions provided, expected MW output is 0.5 MW as the DG picks up a small amount of load. 2) False. Voltage control is plausible if Operator confuses real load and apparent power, as increasing voltage will raise apparent power, but not MW loading. To further increase real load, speed control is raised.

D. Correct-1) True. Per 01-218, when the DG output breaker is closed with the conditions provided, expected MW output is 0.5 MW as the DG picks up a small amount of load. 2) True. To further increase real load, speed control is raised.



Question 49 Info :

Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 0.00


Topic: Expected Actions During DG Parallel RO Importance: 2.8 SRO Importance: 2.8 Cognitive level (High or


Tier/Group 2/1

K/A Info 064 Emergency Diesel Generators (ED/G) A3 Ability to monitor automatic operation of the ED/G system, including: A3.1 O Function of ED/G megawatt load controller

RO 2.8 Importance

Technical 01-21 B References


Learning Apply the following theoretical concepts to Objective the operation of the DGs:

• Parallel operation of AC generators • Relationship between 4.16KV system

parameters and 13.8KV system parameters

• Relationship between DG parameters and 13.8KV system parameters when DG is paralleled to a 4.16KV bus

• Relationship between apparent power, reactive power, real power and generator output voltage for the DG

Question New Source






Comments None

Task 024.006 Parallel a Diesel Generator to a 4KV Bus


50


ID: Q106545 ·Points: 1.00

Unit-1 is at 100% power when the following occurs:

• A Hi-Line 5072 fault lowers Switchyard Voltage to 440 Kv • The fault clears after 10 seconds and Switchyard Voltage returns to Normal • All 4 Kv busses see voltage dip to 3.5 Kv for 10 seconds • The Unit is manually tripped • During EOP-0, while acknowledging DG panel alarms, the 1 B DG STOP pushbutton is also

depressed

Which ONE of the following describes 1) the impact on the 1 B DG and 2) what actions will be required to restart the 1 B DG?

A. 1) The 1 B DG will stop and all DG auxiliaries will lose power. 2) Verify 1 B DG automatically starts and repowers 14 4Kv bus after a oneminute time delay.

B. 1) The 1 B DG will stop and all DG auxiliaries will lose power. 2) Depress the 1 B DG ST ART Pushbutton and verify 1 B DG automatically starts and repowers 14 4Kv bus.

C. 1) The 1 B DG will stop and all DG auxiliaries will maintain power. 2) Verify 1 B DG automatically starts and repowers 14 4Kv bus after a oneminute time delay.

D. 1) The 1 B DG will stop and all DG auxiliaries will maintain power. 2) Depress the 1 B DG START Pushbutton and verify 1 B DG automatically starts and repowers 14 4Kv bus.

Answer: A

Answer Explanation:

A. Correct 1) True. The conditions provided result in the 1 B DG starting and providing the only power source to the 14 4Kv bus. The SWYD voltage transient will result in a UV condition, stripping both 11 and 14 4Kv busses (since voltage actuates the TUR below 3.630 V for 10 seconds.) The UV signal will start the 1 B DG and load it onto the 14 4Kv bus. If the 1 B DG STOP pushbutton is inadvertently depressed in this condition, the DG will stop and all power to 14 4Kv bus will be lost. The 1 B DG auxiliaries are powered from MCC-104, which will lose power when 14 4Kv bus loses power. 2) True. The 1 B DG STOP pushbutton will energize the Shutdown Relay for one minute and then clear. If there are no active trips, which there are none given, and an active start signal exists, the DG will restart and if a UV still exists, power up the 14 4Kv bus.

B. Incorrect 1) True. The conditions provided result in the 1 B DG starting and providing the only power source to the 14 4Kv bus. The SWYD voltage transient will result in a UV condition, stripping both 11 and 14 4Kv busses (since voltage actuates the TUR below 3.630 V for 10 seconds.) The UV signal will start the 1 B DG and load it onto the 14 4Kv bus. If the 1 B DG STOP pushbutton is inadvertently depressed in this condition, the DG will stop and all power to 14 4Kv bus will be lost. The 1 B DG auxiliaries are powered from MCC-104, which will lose power when 14 4Kv bus loses power.


OPERATIONS


2) False. Depressing the 1 B START pushbutton is plausible if Operator believes the DG will start immediately as it does when in a normal standby condition. The 1 B DG STOP pushbutton will energize the Shutdown Relay for one minute and then clear. If there are no active trips, which there are none given, and an active start signal exists, the DG will restart and if a UV still exists, power up the 14 4Kv bus.

C. Incorrect 1) False. Auxiliaries maintaining power is plausible if Operator believes auxiliaries are powered from additional off-site source like OC DG auxiliaries are. The conditions provided result in the 1 B DG starting and providing the only power source to the 14 4Kv bus. The SWYD voltage transient will result in a UV condition, stripping both 11 and 14 4Kv busses (since voltage actuates the TUR below 3.630 V for 10 seconds.) The UV signal will start the 1 B DG and load it onto the 14 4Kv bus. If the 1 B DG STOP pushbutton is inadvertently depressed in this condition, the DG will stop and all power to 14 4Kv bus will be lost. The 1B DG auxiliaries are powered from MCC-104, which will lose power when 14 4Kv bus loses power. 2) True. The 1 B DG STOP pushbutton will energize the Shutdown Relay for one minute and then clear. If there are no active trips, which there are none given, and an active start signal exists, the DG will restart and if a UV still exists, power up the 14 4Kv bus.

D. Incorrect 1) False. Auxiliaries maintaining power is plausible if Operator believes auxiliaries are powered from additional off-site source like OC DG auxiliaries are. The conditions provided result in the 1 B DG starting and providing the only power source to the 14 4Kv bus. The SWYD voltage transient will result in a UV condition, stripping both 11 and 14 4Kv busses (since voltage actuates the TUR below 3.630 V for 10 seconds.) The UV signal will start the 1 B DG and load it onto the 14 4Kv bus. If the 1 B DG STOP pushbutton is inadvertently depressed in this condition, the DG will stop and all power to 14 4Kv bus will be lost. The 1 B DG auxiliaries are powered from MCC-104, which will lose power when 14 4Kv bus loses power. 2) False. Depressing the 1 B START pushbutton is plausible if Operator believes the DG will start immediately as it does when in a normal standby condition. The 1 B DG STOP pushbutton will energize the Shutdown Relay for one minute and then clear. If there are no active trips, which there are none given, and an active start signal exists, the DG will restart and if a UV still exists, power up the 14 4Kv bus.

Page: 157 of 235 13 April 2016




...•

Topic: Impact of depressinq 1 B DG Stop pushbutton RO Importance: 2.7 SRO Importance: 2.9 Cognitive level (High or


Tier/Group 2/1

KIA Info 064 Emergency Diesel Generators (ED/G) A2 Ability to (a) predict the impacts of the following malfunctions or operations on the ED/G system; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: A2.14 Effects (verification) of stopping ED/G under load on isolated bus

RO 2.7 Importance

Technical 01-218, 1C18B ALM References

References Drawing 61086SH0013 (18 DG engine provided control)

Learning Given a copy of drawing 61086SH0013 (1 B Objective DG engine control), determine the sequence

of actions that occur when the DG receives a stop signal (control room pushbutton, local pushbutton, or engine fault with or without a SIAS present ). Include in your determination: • all relays that are energized and/or

deenergized • all contacts and/or switches that are

operated by the above relays • all valves and/or components that are

operated by the above contacts and/or switches



Question New Source




Comments None

Task 024.004 Start a Diesel Generator (DG)


51


>tQ: Q106,§53 Points: 1.00

Unit-1 is operating at 100% power with the following conditions:

• Radiation Safety is performing calibration checks of instrumentation • A calibration check source is completely removed from its portable storage container in the

immediate vicinity of the Service Water (SRW) Radiation Monitors

Which ONE of the following calibration check sources will result in a higher than normal reading on the Service Water Radiation Monitor?

A. Neutron Calibration Source

B. Gamma Calibration Source

C. Beta Calibration Source

D. Alpha Calibration Source

Answer: B

Answer Explanation:

A. Incorrect-Per SD-77, the SRW RMS is a scintillation detector that measure gamma radiation. Neutrons are plausible if Operator believes increase in these detectors during an RCS to SRW leak is due primarily to the increase in the neutron source.

B. Correct-Per SD-77, the SRW RMS is a scintillation detector that measure gamma radiation.

C. Incorrect-Per SD-77, the SRW RMS is a scintillation detector that measure gamma radiation. Beta is plausible if Operator believes increase in these detectors during an RCS to SRW leak is due primarily to the increase in the beta source.

D. Incorrect-Per SD-77, the SRW RMS is a scintillation detector that measure gamma radiation. Alpha is plausible if Operator believes increase in these detectors during an RCS to SRW leak is due primarily to the increase in the alpha source.





Topic: Radiation Type Monitored by RMS RO Importance: 2.5 SRO Importance: 3.0 Cognitive level (High or


Tier/Group 2/1

KIA Info 073 Process Radiation Monitoring (PRM) System K5 Knowledge of the operational implications as they apply to concepts as they apply to the PRM system: K5.01 Radiation theory, including sources, types, units, and effects

RO 2.5 Importance

Technical System Description 77 (RMS) References


Learning Recall the applications for the following types Objective of radiation detectors:

• Scintillation detector • Ion chamber detector • G-M detector • Dosimeter devices

Question New Source




52



Comments None

Task 079.008 Monitor the RMS

Unit-2 is operating at 100% power with the system conditions:

• 21, 22, and 24 Containment Air Coolers (CACs) running in High Speed • None of the CAC 8" Emergency Outlet CVs are open • SRW is in a normal alignment


• Feed line break in Containment • Unit-2 is manually tripped • Loss of Off-Site Power • 22 Service Water (SRW) Pump breaker trips • Containment Pressure is 6 PSIG and slowly rising

Which ONE of the following is the expected SRW flow through the CACs prior to performing Containment Environment in EOP-0?

A. 21 CAC-800 GPM 22 CAC-800 GPM 23 CAC-0 GPM 24 CAC-0 GPM

B. 21 CAC-0 GPM 22 CAC-0 GPM 23 CAC-2000 GPM 24 CAC-2000 GPM

C. 21 CAC-2000 GPM 22 CAC-2000 GPM 23 CAC-0 GPM 24 CAC-0 GPM

D. 21 CAC-2000 GPM 22 CAC-2000 GPM 23 CAC-2000 GPM 24 CAC-2000 GPM

Answer: D


OPERATIONS


Answer Explanation:

A. lncorrect-21/22 CAC flow of 800 GPM is plausible if Operator does not recognize that SIAS has occurred and is counting on manual actions in EOP-0 to raise flow. 0 GPM in 23/24 CAC is plausible if Operator believes 22 SRW not starting will result in loss of SRW header flow to 23/24 CACs. On a LOOP, the SRW pumps will start as soon as the DGs power up their respective busses. Since Containment Pressure is 6 PSIG, a SIAS has occurred. This will provide a start signal to 21/22 SRW Pump and if 22 SRW pump does not start, then 23 Pump will start 1-sec later. With SIAS present, this will initiate flow through all 8" SRW CVs on all 4 CACs. Since 21 SRW Pump will supply 21/22 CAC loads and 23 SRW Pump will supply 23/24 CACs, all 4 CACs should have flowrates of about 2000 GPM.

B. lncorrect-21/22 CAC flow of 0 GPM is plausible if Operator believes 22 SRW not starting will result in loss of SRW header flow to 21/22 CACs. On a LOOP, the SRW pumps will start as soon as the DGs power up their respective busses. Since Containment Pressure is 6 PSIG, a SIAS has occurred. This will provide a start signal to 21/22 SRW Pump and if 22 SRW pump does not start, then 23 Pump will start 1-sec later. With SIAS present, this will initiate flow through all 8" SRW CVs on all 4 CACs. Since 21 SRW Pump will supply 21/22 CAC loads and 23 SRW Pump will supply 23/24 CACs, all 4 CACs should have flowrates of about 2000 GPM.

C. lncorrect-23/24 CAC flow of 0 GPM is plausible if Operator believes 22 SRW not starting will result in loss of SRW header flow to 23/24 CACs. On a LOOP, the SRW pumps will start as soon as the DGs power up their respective busses. Since Containment Pressure is 6 PSIG, a SIAS has occurred. This will provide a start signal to 21/22 SRW Pump and if 22 SRW pump does not start, then 23 Pump will start 1-sec later. With SIAS present, this will initiate flow through all 8" SRW CVs on all 4 CACs. Since 21 SRW Pump will supply 21/22 CAC loads and 23 SRW Pump will supply 23/24 CACs, all 4 CACs should have flowrates of about 2000 GPM.

D. Correct-On a LOOP, the SRW pumps will start as soon as the DGs power up their respective busses. Since Containment Pressure is 6 PSIG, a SIAS has occurred. This will provide a start signal to 21/22 SRW Pump and if 22 SRW pump does not start, then 23 Pump will start 1-sec later. With SIAS present, this will initiate flow through all 8" SRW CVs on all 4 CACs. Since 21 SRW Pump will supply 21/22 CAC loads and 23 SRW Pump will supply 23/24 CACs, all 4 CACs should have flowrates of about 2000 GPM.

Page: 163 of 235 13 April 2016




....

Topic: SRW Pump Trip Impact on CACs RO Importance: 3.5 SRO Importance: 3.9 Cognitive level (High or


Tier/Group 2/1

KIA Info 076 Service Water System (SWS) K3 Knowledge of the effect that a loss or malfunction of the SWS will have on the following: K3.03 Reactor building closed cooling water

RO 3.5 Importance

Technical 63058ASHOOO, 2C13 ALM References


Learning Identify the ESFAS signal that initiates the Objective following:

• Starts Containment Spray Pps/CACs • Opens Containment Spray header CVs

and CAC outlet SRW CVs

Question New Source






Comments None

Task 048.033 Verifies shutdown sequencer loads are operating

Points: ~:00

Unit-1 is operating at 100% with the following conditions:

• 11 SFP Heat Exchanger in service and cooling the 11 Spent Fuel Pool • 12 SFP Heat Exchanger in service and cooling the 21 Spent Fuel Pool


• Manual trip due to the loss of Main Feed • EOP-3 is implemented since no AFW is available • PZR PRESS BLOCK A and PZR PRESS BLOCK B are initiated • Once-Through-Core-Cooling is initiated • Containment Pressure starts rising slowly and is now 3 PSIG and still rising slowly • 11/12 SRW HX SRW OUT TEMP HI alarm is received on 1C13

Which ONE of the following combinations could cause the 11/12 SRW HX SRW OUT TEMP HI alarm?

OPERATIONS

A. A Condensate Booster Pump is started to prepare for Condensate Booster Injection and an AFW Steam Driven Pump is started to initiate feed flow.

B. Heat load on 11 /12 Containment Air Coolers increases and 11 SIG Slowdown HX remains in service.

C. 1 B DG load increases and heat load on 13/14 Containment Air Coolers increases.

D. 1A DG load increases and heat load on 11 SFP HX increases.

Answer: C

Page: 165 of 235 13 April 2016

OPERATIONS


Answer Explanation:

A Incorrect-An increase on the heat load of the SRW side could yield a higher SRW inlet and outlet temperature across the SRW HX, resulting in the 11/12 SRW HX SRW OUT TEMP HI alarm. Both the 18 DG and the 13/14 CACs are loads on the 12 SRW subsystem. CBPs being started is plausible if the Operator is considering the heat load from the CBP lube oil cooler and the AFW pump start is plausible if the Operator is considering the increased temperature in the AFW room and the increased load on the AFW air conditioner. However, both of these loads would be isolated if a SIAS were to occur and the Turbine Building isolation CVs went shut. While SIAS was blocked for PZR pressure, Containment pressure was given as 3 PSIG, and a SIAS would have initiated based on high Containment Pressure. The actions concerning the CBP and AFW pumps would have no bearing on SRW header temperatures as only emergency loads would remain in service.

B. Incorrect-An increase on the heat load of the SRW side could yield a higher SRW inlet and outlet temperature across the SRW HX, resulting in the 11/12 SRW HX SRW OUT TEMP HI alarm. 11/12 CACs are loads on the 11 SRW subsystem. 11 S/G Slowdown HX is plausible if the Operator believes 11 HX is cooled by SRW. It is 12 Slowdown HX that is cooled by SRW. Additionally, OTCC is not initiated until both SIG levels are -350". This is well below AFAS setpoint of -170". If an AFAS were to occur, the Slowdown CVs would shut, securing all blowdown flow and heat input to any of the Slowdown HXs.

C. Correct-An increase on the heat load of the SRW side could yield a higher SRW inlet and outlet temperature across the SRW HX, resulting in the 11/12 SRW HX SRW OUT TEMP HI alarm. Both the 18 DG and the 13/14 CACs are loads on the 12 SRW subsystem.

D. Incorrect-An increase on the heat load of the SRW side could yield a higher SRW inlet and outlet temperature across the SRW HX, resulting in the 11/12 SRW HX SRW OUT TEMP HI alarm. Both the 1 B DG and the 13/14 CACs are loads on the 12 SRW subsystem. The 1A DG is plausible if the Operator believes cooling for the 1A is from 11 SRW header (like 2A and 21 header.) The1A DG is not cooled by the plant's SRW system, but operates with its own cooling subsystems. 11 SFP HX is a load on the 12 SRW subsystem.

Page: 166 of 235 13 April 2016



System ID: 106555 User-Defined ID: Q106555 Cross Reference Number: . Topic: SRW Emergency Heat Loads on OTCC RO Importance: 3.7 SRO Importance: 3.7 Cognitive level (High or

H Low): NRC K/A Info, References:

Tier/Group 2/1

K/A Info 076 Service Water System (SWS) A3 Ability to monitor automatic operation of the SWS, including: A3.02 Emergency heat loads

RO 3.7 Importance

Technical 1C13 ALM, AOP-7B, EOP-Att References


Learning Given any of the following alarms identify the Objective most likely cause and determine the

corrective action(s ) required to clear the alarms: • 11(21) or 12(22) SRW HX Outlet

Temperature High • 11 (21) or 12(22) SRW Header

Pressure Low • 13(23) SRW Pump Breaker Lineup

Improper • SRW Pump SIAS Blocked/Auto Start • Turbine Building SRW Header

Pressure Low 11(21) or 12(22) SRW Head Tank

Level

Question New Source






Comments None

Task 011.005 Monitor operation of the Service Water system

Points: 1.00

Both Units are operating at 100% power when the following occurs:

• Both Unit-2 Instrument Air Compressors are tripped and cannot be reset • At 0055, Instrument Air pressure was 85 PSIG and slowly lowering • 21 Plant Air Compressor is tripped and cannot be reset • At 0100, Plant Air pressure was 83 PSIG and lowering • No Operator actions have been taken yet per AOP-70

Which ONE of the following are the expected Unit-2 air pressures at 011 O?

OPERATIONS

A. 1) Unit-2 Instrument Air Header pressure will be >85 PSIG and slowly rising or steady. 2) Unit-2 Plant Air Header pressure will be >83 PSIG and slowly rising or steady.

B. 1) Unit-2 Instrument Air Header pressure will be >85 PSIG and slowly rising or steady. 2) Unit-2 Plant Air Header pressure will be <83 PSIG and continuing to lower.

C. 1) Unit-2 Instrument Air Header pressure will be <85 PSIG and continuing to lower. 2) Unit-2 Plant Air Header pressure will be >83 PSIG and slowly rising or steady.

D. 1) Unit-2 Instrument Air Header pressure will be <85 PSIG and continuing to lower. 2) Unit-2 Plant Air Header pressure will be <83 PSIG and continuing to lower.

Answer: B

Page: 168 of 235 13 April 2016

OPERATIONS


Answer Explanation:

A. Incorrect 1) True. As IA pressure lowers, the PA to IA CV (PA-2061-CV) will open. Since U-1 and U-2 PA systems are cross tied, the lowering PA pressure will cause the backup PAC to start. This will supply U-2 PA with U-1 PA and since U-2 PA and IA are now tied via PA-2061-CV, U-2 IA will be supplied by U-2 PA. 2) False. PA pressure starting to rise is plausible if the Operator believes that U-1 PA will maintain U-2 PA loads as forgets that the PA to PA CV shuts as PA pressure lowers. As PA pressure lowers, the PA to PA CV (PA-2059-CV) will shut and PA loads will cause PA pressure to continue to lower.

B. Correct 1) True. As IA pressure lowers, the PA to IA CV (PA-2061-CV) will open. Since U-1 and U-2 PA systems are cross tied, the lowering PA pressure will cause the backup PAC to start. This will supply U-2 PA with U-1 PA and since U-2 PA and IA are now tied via PA-2061-CV, U-2 IA will be supplied by U-1 PA. 2) True. As PA pressure lowers, the PA to PA CV (PA-2059-CV) will shut and PA loads will cause PA pressure to continue to lower.

C. Incorrect 1) False. Lowering IA pressure is plausible if Operator forgets cross-tie feature of PA and IA and evaluates only the status of the U-2 IAC. As IA pressure lowers, the PA to IA CV (PA-2061-CV) will open. Since U-1 and U-2 PA systems are cross tied, the lowering PA pressure will cause the backup PAC to start. This will supply U-2 PA with U-1 PA and since U-2 PA and IA are now tied via PA-2061-CV, U-2 IA will be supplied by U-1 PA. 2) False. PA pressure starting to rise is plausible if the Operator believes that U-1 PA will maintain U-2 PA loads as forgets that the PA to PA CV shuts as PA pressure lowers. As PA pressure lowers, the PA to PA CV (PA-2059-CV) will shut and PA loads will cause PA pressure to continue to lower.

D. Incorrect 1) False. Lowering IA pressure is plausible if Operator forgets cross-tie feature of PA and IA and evaluates only the status of the U-2 IAC. As IA pressure lowers, the PA to IA CV (PA-2061-CV) will open. Since U-1 and U-2 PA systems are cross tied, the lowering PA pressure will cause the backup PAC to start. This will supply U-2 PA with U-1 PA and since U-2 PA and IA are now tied via PA-2061-CV, U-2 IA will be supplied by U-2 PA. 2) True. As PA pressure lowers, the PA to PA CV (PA-2059-CV) will shut and PA loads will cause PA pressure to continue to lower.

Page: 169 of 235 13 April 2016




Topic: IA Impact on Cross-Tied Unit RO Importance: 3.0 SRO Importance: 3.4 Cognitive level (High or


Tier/Group 2/1

KIA Info 078 Instrument Air System (IAS) K3 Knowledge of the effect that a loss or malfunction of the IAS will have on the following: K3.03 Cross-tied units

RO 3.0 Importance

Technical AOP-7D, 60712SH0001 References


Learning Recall the automatic functions of the Objective following components, and their functional

relationship to a loss of instrument air: • Plant Air to IA CV • PA to PA CV • CV-2085 • Standby Instrument Air Compressor

Question New Source






Comments None

Task 202.069 Respond to a Loss of Instrument Air in Modes 1 & 2

JS5 Points~ 1coo: Unit-2 is operating at 100% when the following transient occurs:

• The reactor trips • SIAS has actuated

While implementing EOP-0, the following indications are noted:

• 2 stuck CEAs. • One charging pump is operating • 2-CVC-508-MOV, 22 BAST Gravity Feed is open • 2-CVC-501-MOV, VCT Outlet is closed • Pressurizer level is 120 inches and stable • Pressurizer pressure is 1925 PSIA and lowering • 21 and 22 SIG levels are -120 inches and lowering • 21 and 22 SIG pressures are 800 PSIA and lowering • Containment pressure is 1.0 PSIG and rising • Containment temperature is 165°F and rising • P-13000-2 is de-energized • No additional actions have been taken.

Which ONE of the following groups of safety functions must be reported as "Cannot be Met"?

A. Reactivity Control and RCS Pressure/Inventory Control.

B. Reactivity Control and Core/RCS Heat Removal.

C. Vital Auxiliaries and Containment Environment.

D. Core/RCS Heat Removal and Containment Environment.

Answer: D


OPERATIONS


Answer Explanation:

A Incorrect-Reactivity Not Met is plausible if the Operator does not believe Reactivity is met with only one charging pump operating to provide boration (since EOP-0 directs that all available Charging Pumps be started, and 12/13 would not be running due to loss of P-13000-2 until manually started.) PIG is Not Met due to negative trends in RCS pressure.

B. Incorrect- Reactivity Not Met is plausible if the Operator does not believe Reactivity is met with only one charging pump operating to provide boration (since EOP-0 directs that all available Charging Pumps be started, and 12/13 would not be running due to loss of P-13000-2 until manually started.) Core and RCS Heat Removal are Not Met since S/G pressure and level are not trending in a positive manner and SIG pressure will be below 800 PSIA shortly. Containment Environment is Not Met since Containment pressure is > 0.7 PSIG and temperature is >120F. ·

C. Incorrect-Vital Auxiliaries Not Met is plausible if the Operator believes that the loss of P-13000-2 will result in the loss of 14 4Kv bus and the loss of one 4Kv bus will require VA assessed as Not Met. Only one 4Kv bus is required for VA to be met in EOP-0. Containment Environment is Not Met since Containment pressure is > 0. 7 PSIG and temperature is >120F.

D. Correct-Core and RCS Heat Removal are Not Met since S/G pressure and level are not trending in a positive manner and SIG pressure will be below 800 PSIA shortly. Containment Environment is Not Met since Containment pressure is > 0. 7 PSIG and temperature is >120F.

Page: 172 of 235 13 April 2016




Topic: Given conditions, Assess Safety Function Status RO Importance: 3.7 SRO Importance: 4.1 Cognitive level (High or


KIA Info 103 Containment System A 1 Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the containment system controls including: A 1.01 Containment pressure, temperature, and humidity

RO 3.7 Importance

Technical EOP-0, Post Trip Immediate Actions References N0-1-201, Calvert Cliffs Operating Manual


Learning Given a plant condition requiring a reactor Objective trip the license candidate will demonstrate an

understanding of the strategy, basis and operator actions of EOP-0 to direct or implement the procedural steps, including warnings, notes, and cautions







Comments None

Task 201.016 Verify the CE Safety Function is satisfied



Licensed Operator Requal Training (LOR)

Operations Procedure References (from Nucleis)

EOP EOP-00-1 POST TRIP IMMEDIATE ACTIONS

System Designations

Emergency Operating Procedures (EOPs)


56


ID: 0106559 Points: 1.00

Unit-2 has tripped from 100%.

Which ONE of the following choices indicate the RCS parameter monitored on the Safety Parameter Display System is suspect and Control Panel indications should be used instead?

A Core Exit Temperature is Magenta and reads 536°F.

B. 11 Loop Subcooled Margin is Green and reads 97°F.

C. PZR Pressure is Red and reads 1735 PSIA.

D. PZR Level is Yellow and reads 75".

Answer: A

Answer Explanation:

A Correct-Per Ol-50A, on a loss of data or a missing sensor, the indications associated with the bad SPDS information will turn magenta. Although a CET reading of 536F is expected following a trip, the fact that the indication is magenta should trigger the Operator to evaluate alternate indications to make a more accurate assessment.

B. Incorrect-Per Ol-50A, on a loss of data or a missing sensor, the indications associated with the bad SPDS information will turn magenta. A SCM reading of 97F and green is plausible if the Operator is used to 100% SCM values in the 50-60F range, and anything outside of this range could be a suspect indication. SCM will go to 90-100F normally on a trip.

C. Incorrect-Per Ol-50A, on a loss of data or a missing sensor, the indications associated with the bad SPDS information will turn magenta. A pressure of 1735 PSIA and red is plausible if the Operator associates red color with danger in using the indication.

D. Incorrect-Per Ol-50A, on a loss of data or a missing sensor, the indications associated with the bad SPDS information will turn magenta. A level of 75" and yellow is plausible if the Operator associates yellow color with caution in using the indication.



Question 56 Info '

Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 0.00


'

Topic: PIC Indications on SPDS durinq a Trip RO Importance: 3.4 SRO Importance: 3.7 Cognitive level (High or


Tier/Group 212

KIA Info 002 Reactor Coolant System (RCS) A4 Ability to manually operate and/or monitor in the control room: A4.08 Safety parameter display systems

RO 3.4 Importance

Technical System Description 058 (Reactor Protection References System)


Learning Identify the color coding scheme of the SPDS Objective critical safety function boxes and parameters

and determine when safety function parameters are invalid or at a reduced logic

Question New Source




Comments None



I Task I 094.008 Monitor the SPDS RCS Pressure and Inventory Control safety function I

57 ID: Q106562

Unit-2 is opertating at 100% with the following initial conditions:

• 21 Charging Pump is selected as the running pump


• 24 4Kv bus faults • The reactor is manually tripped

Which ONE of the following is the initial condition of Charging Pumps in EOP-0?

A. ALL Charging Pumps are running.

B. ONLY 21 and 23 Charging Pumps are running.

C. ONLY 21 and 22 Charging Pumps are running.

D. ONLY 21 Charging Pumps is running.

Answer: B


OPERATIONS


Answer Explanation:

A. Incorrect-All Charging Pumps running is plausible if the Operator believes the 2B DG will repower the 24 4Kv bus after a bus fault and repower the 22 Charging Pump. However, the 2B DG will not close in on a bus with a fault and 22 Charging Pump will not have power.

B. Correct-Normal Charging Pump power lineup is 23 aligned to 21 480V bus. Both 21 and 23 Charging Pump have power. Only 22 Pump would have no power due to loss of 24 4Kv bus. On trip, all Charging Pumps get a signal to start due to low PZR level experienced during the trip. Since 21 and 23 Pumps have power, they would initially be running based on PZR level control system. In addition, the loss of 24 4Kv bus equates to the loss of 2Y10, which will provide a signal to start all Charging Pumps as well.

C. Incorrect-Only 21 and 22 Charging Pump running is plausible if the Operator believes the 2B DG will repower the 24 4Kv bus after a bus fault and restore power to 22 Charging Pump.However, the 2B DG will not close in on a bus with a fault and 22 Charging Pump will not have power. 23 Charging Pump not running is plausible if the Operator believes the normal power supply to 23 Charging Pump is from the B-Train, like SW, SRW, and CCW. Normal Charging Pump power lineup is 23 aligned to 21 480V bus. Both 21 and 23 Charging Pump have power. Only 22 Pump would have no power due to loss of 24 4Kv bus. On trip, all Charging Pumps get a signal to start due to low PZR level experienced during the trip. Since 21 and 23 Pumps have power, they would initially be running based on PZR level control system. In addition, the loss of 24 4Kv bus equates to the loss of 2Y10, which will provide a signal to start all Charging Pumps as well.

D. Incorrect-Only 21 Charging Pump running is plausible if the Operator believes the normal power supply to 23 Charging Pump is from the B-Train, like SW, SRW, and CCW. Normal Charging Pump power lineup is 23 aligned to 21 480V bus. Both 21 and 23 Charging Pump have power. Only 22 Pump would have no power due to loss of 24 4Kv bus. On trip, all Charging Pumps get a signal to start due to low PZR level experienced during the trip. Since 21 and 23 Pumps have power, they would initially be running based on PZR level control system. In addition, the loss of 24 4Kv bus equates to the loss of 2Y10, which will provide a signal to start all Charging Pumps as well.

Page: 178 of 235 13 April 2016



,,


Topic: Charqinq Pump Response to Loss of Power RO Importance: 3.1 SRO Importance: 3.2 Cognitive level (High or


Tier/Group 212

KIA Info 011 Pressurizer Level Control System (PZR LCS) K2 Knowledge of bus power supplies to the following: K2.01 Charging pumps

RO 3.1 Importance

Technical Ol-27D-2, AOP-71, EOP-0 References


Learning Given an electrical bus malfunction evaluate Objective the effect on Charging Pumps operation

Question New Source




Comments None



I Task 1201.014 Verify the PIG Safety Function is satisfied

.. ID: Q106563

Given Provided Reference

Unit-1 is at 100% when the following occurs:

• 11A Low Pressure Feedwater Heater (LPFWH) Level (1-Ll-1446) on 1C03 indicates pegged High

• No abnormal Plant Computer alarms have been received • FWH LVL HI alarm on 1C03 is not in alarm • White light above 11A FWH Level on 1C03 is dimly lit • Turbine is still online

Which ONE of the following is the expected plant response?

A. The 11A LPFWH High-Level Dump should go full open and the Turbine should remain online.

B. The 11A LPFWH High-Level Dump should remain shut and the Turbine should remain online.

C. The 11A LPFWH High-Level Dump should remain shut and the Turbine should trip.

D. The 11A LPFWH High-Level Dump should go full open and the Turbine should trip.

Answer: B

I


OPERATIONS


Answer Explanation:

A. Incorrect-The High-Level dump going full open is plausible if the Operator believes that a single detector is providing the signal to the 1 C03 indication and the High-Level Dump CV. Based on 60701SH0001 and the separation of detectors sending signals to CR level indication, Control Valve control, and Turbine trip functions, only L T-1146 has failed high. Since this LT only feeds 1-Ll-1446 on 1 C03, there will be no other plant impact. The LPFWH Hi-Level dump is normally shut at 100% and would remain shut. Since an actual high level does not exist (as this would have yielded 1C03 alarms, a bright white light on 1C03, and a Plant Computer alarm), the turbine should remain on line.

B. Correct-Based on 60701 SH0001 and the separation of detectors sending signals to CR level indication, Control Valve control, and Turbine trip functions, only L T-1146 has failed high. Since this LT only feeds 1-Ll-1446 on 1 C03, there will be no other plant impact. The LPFWH Hi-Level dump is normally shut at 100% and would remain shut. Since an actual high level does not exist (as this would have yielded 1C03 alarms, a bright white light on 1C03, and a Plant Computer alarm), the turbine should remain on line.

C. Incorrect-The Turbine tripping is plausible if the Operator believes that a single detector is providing the signal to the 1 C03 indication and the Turbine trip function. Based on 60701 SH0001 and the separation of detectors sending signals to CR level indication, Control Valve control, and Turbine trip functions, only L T-1146 has failed high. Since this LT only feeds 1-Ll-1446 on 1C03, there will be no other plant impact. The LPFWH HiLevel dump is normally shut at 100% and would remain shut. Since an actual high level does not exist (as this would have yielded 1C03 alarms, a bright white light on 1C03, and a Plant Computer alarm), the turbine should remain on line.

D. Incorrect-The High-Level dump going full open and the Turbine tripping is plausible if the Operator believes that a single detector is providing the signal to the 1 C03 indication, the High-Level Dump CV, and the Turbine trip function. Based on 60701 SH0001 and the separation of detectors sending signals to CR level indication, Control Valve control, and Turbine trip functions, only L T-1146 has failed high. Since this LT only feeds 1-Ll-1446 on 1 C03, there will be no other plant impact. The LPFWH Hi-Level dump is normally shut at 100% and would remain shut. Since an actual high level does not exist (as this would have yielded 1C03 alarms, a bright white light on 1C03, and a Plant Computer alarm), the turbine should remain on line.

Page: 181 of 235 13 April 2016




.·

Topic: LPFW HiQh Level Indications-Trip RO Importance: 2.7 SRO Importance: 2.8 Cognitive level (High or


Tier/Group 212

KIA Info 016 Non-Nuclear Instrumentation System (NNIS) KS Knowledge of the operational implication of the following concepts as they apply to the NNIS: KS.01 Separation of control and protection circuits

RO 2.7 Importance

Technical 60701 SH0001, 1 C03 ALM References

References 60701 SH0001 provided

Learning Recall the different conditions that will trip the Objective Unit 1 main turbine

Question New Source






Comments None

Task 047.008 Return a feedwater heater to service



59 ll:>,,:.·Q106564 •.. Points: 1.00

Which ONE of the following could be an indication of the core becoming uncovered?

OPERATIONS

A A CET temperature of 535°F with RCS pressure at 1100 PSIA

B. A CET temperature of 570°F with RCS pressure at 1550 PSIA

C. A CET temperature of 590°F with RCS pressure at 1350 PSIA

D. A CET temperature of 605°F with RCS pressure at 1700 PSIA

Answer: C

Answer Explanation:

A Incorrect-Per EOP-5, indication of core uncovery is superheated conditions. Conditions are subcooled. Plausible if Operator reads Steam Tables incorrectly.

B. Incorrect-Per EOP-5, indication of core uncovery is superheated conditions. Conditions are subcooled. Plausible if Operator reads Steam Tables incorrectly.

C. Correct-Per EOP-5, indication of core uncovery is superheated conditions. Conditions are superheat.

D. Incorrect-Per EOP-5, indication of core uncovery is superheated conditions. Conditions are subcooled. Plausible if Operator reads Steam Tables incorrectly.



Topic: Indications of Core Uncovery RO Importance: 3.6 SRO Importance: 3.8 Cognitive level (High or


I Tier/Group 1212 I

Page: 184 of 235 13 April 2016


K/A Info 017 In-Core Temperature Monitor System (ITM) A3 Ability to monitor automatic operation of the ITM system including: A3.01 Indications of normal, natural, and interrupted circulation of RCS

RO 3.6 Importance

Technical EOP-5, Steam Tables References

References Steam Tables provided

Learning Given references, plant conditions and/or Objective RCS parameter values, determine subcooled

margin during EOP-5

Question New Source




Comments None

Task 201.060 Maintain RCS flow verification


60


ID: Q20349 Points: 1.00

Unit-1 is operating at 100% power with the following conditions:

• 11, 12, and 13 Condensate Pumps are all initially running • 11 Condensate Pump trips • AOP-3G is implemented

Which ONE of the following describes 1) the effect this will have on the secondary systems and 2) what initial steps should be taken to mitigate the consequences?

A. 1) Main Generator output lowers. 2) Reduce power to restore Tcold to program.

B. 1) Condensate header pressure lowers. 2) Place the Hotwell level control in manual and bypass the Condensate Demineralizers and Precoat Filters.

C. 1) 12 and 13 Condensate Pumps cavitate. 2) Reduce power to maintain Condensate header flow less than 8,000 GPM.

D. 1) Feed Pump suction pressure initially lowers and then rises. 2) Verify a Condensate Booster Pump automatically started.

Answer: B


OPERATIONS


Answer Explanation:

A. Incorrect 1) False. Lowering Main Generator output is plausible since output lowers by 20 MWe on a loss of a Heater Drain Pump. For a loss of a running Condensate pump, MWe output does not change substantially. 2) False. Restoring Tcold to program is plausible since this is action for a loss of load (AOP-7F). The Operator may believe a loss of load has occurred due to the loss of 20 MWe output. However, the HOP Trip induced Main Generator output loss is due to a loss of efficiency and not a mismatch between the primary and the secondary. Reactor power will not need to be adjusted.

B. Correct 1) True. Loss of a running Condensate pump will result in a lower condensate header pressure and a lower SGFP suction pressure. 2) True. AOP-3G actions to increase header pressure include shutting the Hotwell MakeUp and Dump CVs by placing 1-LIC-4405 in manual with a 50% output, bypassing the Precoat system, and bypassing the Condensate Demin system.

C. Incorrect 1) True. Per Ol-11A, 2 Condensate Pumps running above 70% power will result in cavitation and increased impeller wear. 2) False. Restoring Condensate header flow <8000 GPM is plausible since this is an action on a loss of Condensate Pumps. However, this action is required when only 1 Condensate Pump remains. For this question, only 11 Condensate Pump was lost, leaving both 12 and 13 Condensate Pumps running.

D. Incorrect 1) False. Feedwater (FW) suction pressure will initially lower, but will not rise until manual actions are taken. Suction pressure will not rise until AOP-3G actions to increase header pressure (shutting the Hotwell Make-Up and Dump CVs by placing 1-LIC-4405 in manual with a 50% output, bypassing the Precoat system, and bypassing the Condensate Demin system) are taken. FW pressure rising is plausible if the Operator believes the Condensate Booster Pump will automatically start. 2) False. Verifying a Condensate Booster Pump automatically starts is plausible since this is an AOP-3G action for a loss of a Condensate Booster Pump. AOP-3G does not direct the start of a Condensate Booster Pump when just one Condensate Pump trips.

Page: 187 of 235 13 April 2016




Topic: Loss of Condensate Pump RO Importance: 2.6 SRO Importance: 2.8 Cognitive level (High or


Tier/Group 212

KIA Info 056 Condensate System A2 Ability to (a) predict the impacts of the following malfunctions or operations on the Condensate System; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: A2.04 Loss of condensate pumps

RO 2.6 Importance



Learning Identify the actions taken upon the failure of Objective any of the following pumps for all Reactor

Power Levels. • SGFP • Condensate Booster Pump • Condensate Pump • Heater Drain Pump

Question New Source






Comments None

Task 202.036 Respond to a failure of a pump with Reactor Power~ 5%


61


ID: Q106565 Points: 1.00


• Mode 6 • Containment Purge operating

Which ONE of the following could result in the automatic isolation of Containment Purge?

A. High Rad trash move on the 1 O' of Containment.

B. Irradiated fuel assembly is dropped and damaged in the Upender.

C. Outside air temperature drops to 55°F without Plant Heating in service.

D. Reactor Vessel Head is moved from over the Vessel to the Head Lay Down Area with the Containment Area Radiation Monitors in Level Cal.

Answer: B

Answer Explanation:

A. Incorrect-Per 1C08 ALM, possible causes of a CRS include a rupture of a fuel element during refueling. Per 61058ASH0001, a CRS actuation will automatically isolate purge. Movement of High Rad trash is plausible if the Operator believes high radiation anywhere in Containment will result in a CRS. The Containment Area RMS are located on the 69' and would not be affected by movement of material on the 10'.

B. Correct-Per 1 C08 ALM, possible causes of a CRS include a rupture of a fuel element during refueling. Per 61058ASH0001, a CRS actuation will automatically isolate purge.

C. Incorrect-Per 1C08 ALM, possible causes of a CRS include a rupture of a fuel element during refueling. Per 61058ASH0001, a CRS actuation will automatically isolate purge. Outside air temperature is plausible since there is a temperature limit in 01-36 to ensure Plant Heating is in service and aligned to Containment Purge with Containment temperatures <45F, and Purge is to be manually secured if Containment Temperature remains <45F. The Operator could confuse guidance in Ol-22B for Aux Building Ventilation where Plant Heating should be aligned when temperatures are <60F and the Aux Building Supply Fan trips if supply temperature drops to 40F.

D. Incorrect-Per 1 C08 ALM, possible causes of a CRS include a rupture of a fuel element during refueling. Per 61058ASH0001, a CRS actuation will automatically isolate purge. Movement of the RV Head is plausible since this will increase radiation levels on the 69' and near the Containment Area RMS. However, with the RMS being in Level Cal, this would prevent actuation of CRS and would prevent automatic isolation of Containment Purge. 01-35 provides guidance on when Level Cal is used to prevent actuation of CRS.





Topic: Conditions Causinq CRS RO Importance: 3.4 SRO Importance: 3.4 Cognitive level (High or


Tier/Group 212

KIA Info 029 Containment Purge System (CPS) A 1 Ability to predict and/or monitor changes in parameters to prevent exceeding design limits) associated with operating the Containment Purge System controls including: A 1 . 02 Radiation levels

RO 3.4 Importance

Technical 01-35, 1C08 ALM, 61058ASH0001 References


Learning Given any ESFAS alarm condition assess Objective the impact on plant operation

Question New Source






Comments None

Task 048.018 Test CRS "A" & "B" isolation of Cntmt Purge system


62


ID: Q106567 ... Points: 1.00


• Mode 3 • RCS is at 532°F and 2250 PSIA • Main Feedwater is in service • 11 AFW Pump is running for testing • AFW is feeding both 11 and 12 S/Gs at 50 GPM • 11 and 12 S/Gs are at O" and steady

A Wide Range Level Transmitter on 11 SIG, 1-LT-1114A, fails low.

Which ONE of the following is the expected plant response to the failed level transmitter?

A. Actual 11 SG level would remain steady; Actual 12 SG level would remain steady.

B. Actual 11 SG level would remain steady; Actual 12 SG level would rise.

C. Actual 11 SG level would rise; Actual 12 SG level would remain steady.

D. Actual 11 SG level would rise; Actual 12 SG level would rise.

Answer: A

Answer Explanation:

A. Correct-LT-1114A provides level indication at 1C04 and provides level input to AFAS. A transmitter failing low will result in the PAMS Channel A reading -400" on 1 C04. The LT does not provide an input to the MFW system, so MFW will not respond to an indicated low condition. AFAS will not initiate since only one LT has failed low. AFAS needs 2 of 4 to indicate < -170" before it initiates. L T-1114A failing low would therefore have no impact on the SIG level control and the levels would remain steady at O".

B. lncorrect-12 SG level rising is plausible if Operator believes SG level control system will start raising 12 SG level since 11 has gone low and is anticipating an increase in feed. This is not the case since L T-1114A only provides level indication at 1 C04 and level input to AFAS and L T-1114A is not part of the SG level control system.

C. lncorrect-11 SG level rising is plausible if Operator believes SG level control system will start raising 11 SG level due to indicated level failing low. This is not the case since LT-1114A only provides level indication at 1C04 and level input to AFAS and L T-1114A is not part of the SG level control system.

D. Incorrect-Both SG levels increasing is plausible if the Operator believes AFAS has initiated because of the low level. AFAS needs 2 of 4 LT to indicate< -170" before it initiates. LT-1114A failing low would not result in an AFAS and the levels would remain steady at O".





Topic: Effect of L T-1114A Failing Low RO Importance: 2.6 SRO Importance: 3.0 Cognitive level (High or


Tier/Group 212

KIA Info 035 Steam Generator System (S/GS) K6 Knowledge of the effect of a loss or malfunction on the following will have on the S/GS: K6.03 SIG level detector

RO 2.6 Importance

Technical 1 C03 ALM, 1 C04 ALM, 610702SH0004 References


Learning Explain the initiating plant conditions and Objective predict the AFAS response actions for the

following: • AFAS Start • AFAS Block • AFW Pipe Rupture • SIG High Level • AFW No Flow

Question New Source






Comments None

Task 036. 001 Fill the SI Gs from 1 (2)C04

JO: .Q106568:' . Points: 1.00


• Startup from RFO in progress • Reactor Power is 10% and steady • RCS temperature is 532°F and steady • SIG Pressures are 900 PSIG and steady • OP-3 is in progress

Which ONE of the following is 1) the reason the setpoint on the TBV Controller, 2-PIC-4056, is lowered before continuing with the startup and 2) why is this done?

OPERATIONS

A. 1) To open TBVs and lower SIG pressures. 2) This will lower RCS temperature and prevent the SIG Safeties from lifting.

B. 1) To open TBVs and lower SIG pressures. 2) This will allow better RCS temperature control when raising Generator Load.

C. 1) To shut the TBVs and raise SIG pressures. 2) This allows Reactor Power to rise while minimizing the volume of Liquid RadWaste created.

D. 1) To shut the TBVs and raise SIG pressures. 2) This allows Reactor Power to rise while limiting ASI.

Answer: B

Page: 195 of 235 13 April 2016

OPERATIONS


Answer Explanation:

A Incorrect 1) True. Lowering the setpoint on the TBV controller will open the TBVs and lower SG pressure. 2) False. Lowering RCS temperature and lowering SIG pressures to provide more margin to the SIG Safety Lift setpoints is plausible since this will be true, but this is not the reason the controller output is lowered per OP-3. The SIG pressures are well below the SIG Safety Lift setpoints when RCS temperature is on program and when Reactor Power is rising without the Main Generator paralleled yet. Per OP-3, lowering the TBV setpoint allows the TBVs to modulate further open as Reactor Power increases and modulate further closed as Main Generator load is increased. This will help limit RCS temperature swings due to steam demand changes and will limit magnitude on Reactor Power swings caused by temperature changes.

B. Correct 1) True. Lowering the setpoint on the TBV controller will open the TBVs and lower SG pressure. 2) True. Per OP-3, lowering the TBV setpoint allows the TBVs to modulate further open as Reactor Power increases and modulate further closed as Main Generator load is increased. This will help limit RCS temperature swings due to steam demand changes and will limit magnitude on Reactor Power swings caused by temperature changes.

C. Incorrect 1) False. Shutting the TBVs and raising SIG pressure is plausible if the Operator believes lowering the setpoint is the same as lowering the manual output of the controller. Lowering the manual output from the controller would shut the TBVs. Lowering the setpoint will open the TBVs, not shut them. 2) False. Reactor power rising to minimize Liquid RadWaste is plausible if the Operator determines that less RCS waste will be diverted to the Waste System since the volume of DI will not be added to the RCS to increase power. While this is a response of rising reactor power without use of DI, it is not the reason the TBV controller setpoint is lowered.

D. Incorrect 1) False. Shutting the TBVs and raising SIG pressure is plausible if the Operator believes lowering the setpoint is the same as lowering the manual output of the controller. Lowering the manual output from the controller would shut the TBVs. Lowering the setpoint will open the TBVs, not shut them. 2) False. Reactor power rising to limit impact on ASI is plausible since Reactor power changes without the use of CEAs minimizes the ASI swing with a +MTC. Withdrawing CEAs to raise power will cause ASI to move to upper part of the core, which is same direction it is moving as power is being increased with a +MTC condition like that exiting a RFC. While this is a response of rising reactor power without use of CEAs, it is not the reason the TBV controller setpoint is lowered.

Page: 196 of 235 13 April 2016




Topic: TBV Operations durinq +MTC Startup RO Importance: 2.6 SRO Importance: 2.7 Cognitive level (High or

H Low): NRC KJA Info, References:

Tier/Group 212

KJA Info 041 Steam Dump System (SDS)/Turbine Bypass Control K4 Knowledge of SOS design feature(s) and/or interlock(s) which provide for the following: K4.16 Low main steam pressure

RO 2.6 Importance

Technical OP-2 References


Learning Given plant conditions and/or parameters Objective evaluate the operation of the TBVs/ADVs for

normal and emergency operating conditions

Question New Source






Comments None

Task 093.047 Start and load Unit 2 main turbine qenerator


64


ID: Q95402 Points: 1.00

The Miscellaneous Waste Monitor Tank (MWMT) is being discharged with the following initial conditions:

• MWMT Pump is operating • Discharge flowpath is to Unit-2 with O-MWS-103 and O-MWS-105 both open

The following transient then occurs:

• LIQUID WASTE DISCH RMS monitor, O-RIC-2201, alarms high • Control Room panel 1C22 window annunciator D-3.2, LIQUID WASTE DISCH, has alarmed • LIQUID WASTE DISCH valves, O-MWS-2201-CV and O-MWS-2202-CV, are open

Which ONE of the following is the first expected operator response?

A. Place the handswitches, O-HS-2201 and O-HS-2202, on 1 C33 to shut.

B. Stop the MWMT pump being used to discharge the MWMT.

C. Ensure valves O-MWS-103 and O-MWS-105 are shut to isolate the Unit-2 SG Slowdown overboard discharge path.

D. Continue discharge of MWMT using the procedure for O-RE-2201 not available and energized.

Answer: A

Answer Explanation:

A. Correct-Per 1 C22-ALM, RMS Alarm Manual, this is the appropriate action. Verify means to make it happen if it hasn't. In this case, placing the handswitches for O-MWS-2201-CV and O-MWS-2202-CV in close would cause the valves to shut, terminating the accidental liquid waste release.

B. Incorrect-Stopping the MWMT Pump is plausible since this will stop the MWMT discharge, but this will not isolate the flowpath. Stopping the pump is action specified by AOP-68, Accidental Liquid Waste Release, if the discharge CVs fail to shut when handswitches placed to shut position.

C. Incorrect-Shutting MWS-103/105 is plausible since this will isolate the MWMT discharge flowpath. However, this action is specified by AOP-68, Accidental Liquid Waste Release, if the discharge CVs fail to shut when handswitches placed to shut position.

D. Incorrect-Continuing discharge is plausible since Operator may interpret indications as the RMS fails high rather than the process is causing the high alarms. Continuing the discharge is action per Alarm Manual due to an RMS failure. However, the stated conditions are indicative of an actual high alarm, not an RMS failure.




System ID: 95402 User-Defined ID: Q95402 Cross Reference Number: LOl-032-1

Topic: Liquid Waste Disch CVs fail to shut on high RMS alarm durinq discharqe of MWMT


H Low): Comments:

Tier/Group 2/2

KIA Info 068 Liquid Radwaste 2.2 Equipment Control 2.2.44 Ability to interpret control room indications to verify the status and operation of a system, and understand how operator actions and directives affect plant and system conditions.

RO 4.2 Importance

Technical AOP-68, Accidental Liquid Waste Release References 1 C22 Alarm Response Manual Window D32

Ol-17C-4, MWMT Operations


Learning Determine the automatic actions upon a high Objective alarm on O-RIC-2201







Comments None

Task 202.047 Respond to an uncontrolled liquid waste release

Additional K/A Match Info 059 Accidental Liquid RadWaste Release I 9 AK2 - Knowledge of the interrelations between the Accidental Liquid Radwaste Release and the following: AK2.01 - Radioactive-liquid monitors





65


ID: 0106573 Points: 1.00


• Fire in the Control Room develops requiring Control Room evacuation • AOP-9A-1 is implemented • Red Bus deenergizes • Black Bus deenergizes • Diesel Fire Pump fails to start

Which ONE of the following is the impact, if any, with the Diesel Fire Pump failure?

A There is no suction source available for the 13 AFW Pump.

B. There is no impact as the Electric Fire Pump will be operating.

C. No makeup is available to the 11 and 21 Condensate Storage Tanks.

D. There is a loss of all pressurized air.

Answer: C

Answer Explanation:

A Incorrect-No suction source available to the 13 AFW Pump is plausible if the Operator is considering the actions from the EOP-Attachments which align fire hoses to the suction of the 13 AFW Pump. These actions would not be taken in AOP-9A as suction would be from the CSTs and Fire Protection would provide the makeup to the CST.

B. Incorrect-No impact is plausible if the Operator believes the LOOP only affects the U-1 power supply to the Electric Fire Pump and the pump can be realigned to its alternate power source. The Electric Fire Pump is powered from either 12A or 23A 480V busses. Both of these busses will lose power during a LOOP and AOP-9A does not reenergize the busses.

C. Correct-The Electric Fire Pump is powered from either 12A or 23A 480V busses. Both of these busses will lose power during a LOOP and AOP-9A does not reenergize. With the Diesel Fire Pump failing to start, there is a loss of pressurized fire water. This will prevent the SRW Head Tank, CC Head Tank, 11 and 21 CSTs from being filled from the Fire Protection system during AOP-9A.

D. Incorrect-Loss of pressurized air is plausible to the Operator since 01-19 provides guidance to cool the Instrument Air Compressors with the Fire System. This only applies to a loss of SRW, however. AOP-9A would utilize the SWACs to supply air on Unit-1.





Topic: Loss of Fire Protection Impact on Redundant Shutdown Capabilities

RO Importance: 2.7 SRO Importance: 3.2 Cognitive level (High or L Low): NRG K/A Info, References:

Tier/Group 2/2

K/A Info 086 Fire Protection System (FPS) K3 Knowledge of the effect that a loss or malfunction of the Fire Protection System will have on the following: K3.01 Shutdown capability with redundant equipment

RO 2.7 Importance



Learning Operate the AFW system components as Objective required to maintain RCS temperatures and

SG levels to support AOP-9A plant shutdown

Question New Source






Comments None

Task 013.024 Start and operatre a diesel driven fire pump

ID: Q106575'.'

Given Provided Reference:


• Downpower in progress for RFO • Downpower started one hour ago • Power is currently 73% and lowering

Which ONE of the following is 1) the expected CEA positions at 73% and 2) the expected power indications on Delta-T Power and Linear Range Nuclear Instruments (LRNls)?

A. 1) Group 5 CEAs at 77 .25" 2) Delta-T Power-73%; LRNI Lower-69%; LRNI Upper-73%.

B. 1) Group 5 CEAs at 77.25" 2) Delta-T Power-73%; LRNI Lower-73%; LRNI Upper-69%.

C. 1) Group 5 CEAs at 84.0". 2) Delta-T Power-73%; LRNI Lower-69%; LRNI Upper-73%.

D. 1) Group 5 CEAs at 84.0". 2) Delta-T Power-73%; LRNI Lower-73%; LRNI Upper-69%.

Answer: B


OPERATIONS


Answer Explanation:

A Incorrect 1) True. Per the provided ReMA, Attachment 6 and Figure 2 indicate that CEAs should reach target position of 77.25" one hour into the shutdown, at 73% 2) False. LRNI Lower of 69% and a LRNI Upper of 73% are plausible if the Operator

calculates ASI as U-L/U+L. The expected ASI (L-U/L +U) is positive one hour into the shutdown per provided ReMA Figure 3. A positive ASI means more power is being generated in bottom part of core than in top. Delta-T power will be the most accurate indication during the downpower. The LRNI will read lower than actual power due to the RCS temperature lowering as the downpower progress. A LRNI Lower reading 73% and a LRNI Upper reading 69% yields an ASI of 0.03 (73-69/73+69), which is close to Figure 3 expectations.

B. Correct 1) True. Per the provided ReMA, Attachment 6 and Figure 2 indicate that CEAs should reach target position of 77.25" one hour into the shutdown, at 73%. 2) True. The expected ASI (L-U/L +U) is positive one hour into the shutdown per provided ReMA Figure 3. A positive ASI means more power is being generated in bottom part of core than in top. Delta-T power will be the most accurate indication during the downpower. The LRNI will read lower than actual power due to the RCS temperature lowering as the downpower progress. A LRNI Lower reading 73% and a LRNI Upper reading 69% yields an ASI of 0.03 (73-69/73+69), which is close to Figure 3 expectations.

C. Incorrect 1) False. A Group 5 target of 84" is plausible if the Operator uses the Group 5 position at 45 minutes from Figure 2. Per the provided ReMA, Attachment 6 and Figure 2 indicate that CEAs should reach target position of 77.25" one hour into the shutdown, at 73%. 2) False. LRNI Lower of 69% and a LRNI Upper of 73% are plausible if the Operator calculates ASI as U-L/U+L. The expected ASI (L-U/L +U) is positive one hour into the shutdown per provided ReMA Figure 3. A positive ASI means more power is being generated in bottom part of core than in top. Delta-T power will be the most accurate indication during the down power. The LRNI will read lower than actual power due to the RCS temperature lowering as the downpower progress. A LRNI Lower reading 73% and a LRNI Upper reading 69% yields an ASI of 0.03 (73-69/73+69), which is close to Figure 3 expectations.

D. Incorrect 1) False. A Group 5 target of 84" is plausible if the Operator uses the Group 5 position at 45 minutes from Figure 2. Per the provided ReMA, Attachment 6 and Figure 2 indicate that CEAs should reach target position of 77.25" one hour into the shutdown, at 73%. 2) True. The expected ASI (L-U/L +U) is positive one hour into the shutdown per provided ReMA Figure 3. A positive ASI means more power is being generated in bottom part of core than in top. Delta-T power will be the most accurate indication during the downpower. The LRNI will read lower than actual power due to the RCS temperature lowering as the downpower progress. A LRNI Lower reading 73% and a LRNI Upper reading 69% yields an ASI of 0.03 (73-69/73+69), which is close to Figure 3 expectations.

Page: 205 of 235 13 April 2016




Topic: ReMA Expectations and Indications Durinq a Downpower RO Importance: 4.3 SRO Importance: 4.6 Cognitive level (High or


Tier/Group 3/NA

KIA Info 2.1 Conduct of Operations 2.1.37 Knowledge of procedures, guidelines, or limitations associated with reactivity management

RO 4.3 Importance

Technical OP-AA-300, OP-AP-300-1003, OP-3 References

References OP-AP-300-1003 Attachment 1, Reactivity provided Maneuver Approval Cover Page

OP-AP-300-1003 Attachment 2, Reactivity Maneuver Guidance Sheet

Learning Recall why Linear Range Nuclear Objective Instruments (LRNI) are inaccurate during any

power change

Question New Source






Comments None

Task 203.019 Perform rapid power reduction

OPERATIONS Page 207 of 235 13April2016


ID: Q106576~:,\ · Points: 1.0()

21 STM LINE RUPTURE alarm annunciates on 2C03.

Which ONE of the following indications validate the alarm?

OPERATIONS

A. 21 SIG pressure reads 700 PSIA and 22 SIG pressure reads 700 PSIA.

B. 21 SIG pressure reads 650 PSIA and 22 SIG pressure reads 550 PSIA.

C. 21 SIG pressure reads 500 PSIA and 22 SIG pressure reads 620 PSIA.

D. Containment Pressure is reading 2.9 PSIG.

Answer: C

Answer Explanation:

A. Incorrect-SIG pressure less than 705 is plausible if Operator confuses SGIS setpoint with STEAM LINE RUPTURE alarm. Per 2C03 ALM C-57, alarm is caused by S/Gs having 115 PSID between them.

B Incorrect-SIG pressure differential of 100 PSID is plausible if Operator confuses TM/LP CH PRE-TRIP alarm with STEAM LINE RUPTURE alarm setpoint. Per 2C03 ALM C-57, alarm is caused by S/Gs having 115 PSID between them. SIG differential are reversed as well, as 21 SIG pressure should be lower than 22 SIG pressure for a 21 STEAM LINE RUPTURE alarm.

C. Correct-Per 2C03 ALM C-57, alarm is caused by S/Gs having 115 PSID between them.

D. Incorrect-High Containment Pressure is plausible if Operator believes steam leak will cause a high enough Containment pressure increase to actuate SIAS. Per 2C03 ALM C-57, alarm is caused by S/Gs having 115 PSID between them.

Page 208 of 235 13 April 2016




Topic: 21 STM LINE RUPTURE Alarm Response RO Importance: 4.3 SRO Importance: 4.3 Cognitive level (High or


Tier/Group 3/NA

KIA Info 2.1 Conduct of Operations 2.1.45 Ability to identify and interpret diverse indications to validate the response of another indication

RO 4.3 Importance

Technical 2C03 ALM References


Learning Explain the initiating plant conditions and Objective predict the AFAS response actions for the

following: • AFAS Start • AFAS Block • AFW Pipe Rupture • SIG High Level • AFW No Flow







Comments Modified from 074958. 074958 used on NRC 2010 SRO Exam.

Task 048.034 Notes failure of AFAS block to 11/12 SIG blocks AFAS manually prior to EOP-0 exit


68


. 10: 0106584 Points: 1.00

U-1 is operating with the following initial conditions:

• Time is 0100 • RCS pressure is 540 PSIA • RCS temperature is 250°F • A cooldown is in progress per OP-5, Plant Shutdown From Hot Standby to Cold Shutdown • Aux Spray, 1-CV-517, is opened to continue depressurization

Which ONE of the following is the lowest allowable RCS Pressure that ensures PZR Cooldown rate limits are not exceeded?

A. 380 PSIA at 0200.

B. 190 PSIA at 0200.

C. 120 PSIA at 0200.

D. 50 PSIA at 0200.

Answer: D

Answer Explanation:

A. Incorrect-Per OP-5, the PZR cooldown rate limit is 200F/hr. At time 0100, PZR temperature is 475F (Saturation temperature at 540 PSIA.) At time 0200, PZR temperature could be 275F. Pressure at 50 PSIA equates to 280F. 50 PSIA/280F would not exceed the cooldown rate limit 380 PSIA is plausible if the Operator uses RCS cooldown rate limit on Unit-1 of 40F/hr (380 PSIA/438F.)

B. Incorrect-Per OP-5, the PZR cooldown rate limit is 200F/hr. At time 0100, PZR temperature is 475F (Saturation temperature at 540 PSIA.) At time 0200, PZR temperature could be 275F. Pressure at 50 PSIA equates to 280F. 50 PSIA/280F would not exceed the cooldown rate limit 190 PSIA is plausible if the Operator uses RCS cooldown rate limit on Unit-2 of 100F/hr (190 PSIA/377F.)

C. Incorrect-Per OP-5, the PZR cooldown rate limit is 200F/hr. At time 0100, PZR temperature is 475F (Saturation temperature at 540 PSIA.) At time 0200, PZR temperature could be 275F. Pressure at 50 PSIA equates to 280F. 50 PSIA/280F would not exceed the cooldown rate limit 120 PSIA is plausible if the Operator uses wrong column in steam tables and applies 200F/hr limit to pressure instead of to pressure (540 PSIA-200 PSIA=340 PSIA, 120 PSIA/341F.)

D. Correct-Per OP-5, the PZR cooldown rate limit is 200F/hr. At time 0100, PZR temperature is 475F (Saturation temperature at 540 PSIA.) At time 0200, PZR temperature could be 275F. Pressure at 50 PSIA equates to 280F. 50 PSIA/280F would not exceed the cooldown rate limit





Topic: PZR Cooldown Limits RO Importance: 4.6 SRO Importance: 4.1 Cognitive level (High or


Tier/Group 3/NA

KIA Info 2.2 Equipment Control 2.2.2 Ability to manipulate the console controls as required to operate the facility between shutdown and designated power levels

RO 4.6 Importance

Technical OP-5 (Plant Shutdown From Hot Standby to References Cold Shutdown), Steam Tables

References Steam Tables provided

Learning Recall and apply the General Precautions of Objective OP-5

Question New Source




Comments None


69


1~jT_a_sk~~~j_2_03_.0_2_3_E_n_te_r_M_od_e_4~~~~~~~'1

ID: Q106585 .

Haza1·dous 1----Energy Sou1·ce

l\:IOV-300

CKV-400

CV-200

HV-500 ,. Which ONE of the following valves should be used as an isolation point to protect workers from the Hazardous Energy Source?

A. RV-100: Tagged installed.

B. CKV-400: Tagged installed.

C. CV-200: HS is tagged shut, CV is tagged shut with gag installed, Gag installation verified by cycling air isolation.

D. MOV-300: HS is tagged shut, MOV handwheel is tagged shut, MOV breaker is left closed for position indication verification.

Answer: C


OPERATIONS


Answer Explanation:

A. Incorrect-Per OP-CE-109-101, relief valves should not be used as protection isolation points unless the RV is gagged or restrained. Plausible to the Operator since the RV would normally be shut.

B. Incorrect-Per OP-CE-109-101, check valves should not be used as protection isolation points unless other isolation is practical. The figure depicts a manual valve just downstream of the checkvalve and could be used for protection. Plausible to the Operator since the checkvalve is oriented to stop flow from the hazardous energy source.

C. Correct-Incorrect-Per OP-CE-109-101, air operated valves (POVs) that fail open can be used as protection isolation points if the POV is held in the shut position by a gagging device. The integrity of the gag must be verified by isolating air once the gag is installed. Air is then restored to the POV.

D. Incorrect-Per OP-CE-109-101, MOVs can be used as protection isolation points when valve HS is tagged, the breaker supplying power to the MOV is opened, and the valves handwheel is tagged. Plausible to the Operator since the MOV is tagged in shut position. Power is not removed however, which does not meet OP-CE-109-101 requirements.

Page: 214 of 235 13 April 2016




Topic: Knowledqe of Taqqinq Procedures RO Importance: 4.1 SRO Importance: 4.3 Cognitive level (High or


Tier/Group 3/NA

KIA Info 2.2 Equipment Control 2.2.13 Knowledge of tagging and clearance procedures

RO 4.1 Importance

Technical OP-CE-109-101 (Clearance and Tagging) References


Learning Apply the Requirements of N0-1-112, Safety Objective Tagging

Question New Source




Comments None

Task 204.140 Verify the adequacy of a safety tag out

OPERATIONS Page: 215 of235 13 April 2016

70


ID: Q25936 Points: 1.00


• Reactor startup is in progress. • The reactor is not critical with the Regulating Group CEAs at the Upper Bound CEA Position

on the ECC form.

Which ONE of the following actions must be taken?

A. Maintain all regulating CEAs at the upper bound position, borate to increase RCS boron concentration by 200 ppm, and recalculate the ECC.

B. Insert all Regulating CEAs to the lower bound position and recalculate the ECC.

C. Maintain all Regulating CEAs at the upper bound position and notify Reactor Engineering.

D. Insert regulating CEAs to the ZPDIL position and notify Reactor Engineering.

Answer: D

Answer Explanation:

A. Incorrect-Per OP-2, when CEAs have reached the Upper Bound and the reactor is not critical, the CEAs are inserted to the Zero Power PDIL CEA Position and Reactor Engineering is notified. Actions to borate while maintaining current CEA position are plausible to Operator since the boration is action required if the reactor goes critical below Zero Power PDIL.

B. Incorrect-Per OP-2, when CEAs have reached the Upper Bound and the reactor is not critical, the CEAs are inserted to the Zero Power PDIL CEA Position and Reactor Engineering is notified. Action to return the CEAs to the lower bound is plausible to the Operator since actions are required to insert Regulating CEAs to the Lower Computer Stop if criticality is not achieved by the Lower Bound.

C. Incorrect-Per OP-2, when CEAs have reached the Upper Bound and the reactor is not critical, the CEAs are inserted to the Zero Power PDIL CEA Position and Reactor Engineering is notified. Action to maintaining current CEA position is plausible to the Operator since the answer also includes the required actions to notify Reactor Engineering.

D. Correct-Per OP-2, when CEAs have reached the Upper Bound and the reactor is not critical, the CEAs are inserted to the Zero Power PDIL CEA Position and Reactor Engineering is notified.




System ID: 25936 User-Defined ID: Q25936 Cross Reference Number: LOR-206-11

Topic: Action for criticality above Upper Bound RO Importance: 4.5 SRO Importance: 4.4 Cognitive level (High or


Tier/Group 3/NA

KIA Info 2.2 Equipment Control 2.2.1 Ability to perform pre-startup procedures for the facility, including operating those controls associated with plant equipment that could affect reactivity

RO 4.5 Importance

Technical OP-2 (Plant Startup From Hot Standby to References Minimum Load)


Learning Given reactor startup conditions, evaluate if Objective reactor criticality will occur outside ECC

tolerance limits and identify the required actions when: • Below the ECC Lower CEA Bound, or • Above the ECC Upper CEA Bound







Comments None

Task 203.135 Enter Mode 2


Cognitive Level

COMPREHENSION


11


ID: L92909 Points: 1.00

Unit-1 was operating at 100% power when the following transient occurred: • Loss of Offsite Power • Steam Generator Tube Rupture (SGTR) • The operators implemented the appropriate Optimal Recovery procedure • The affected SIG has been identified • THoT is 516°F (slowly lowering) Which ONE of following is the reason the Optimal Recovery Procedure directs a cooldown of Thot less than 515°F?

A. Minimizes the differential pressure across the break thereby reducing the leakrate.

B. Establishes natural circulation cooling as soon as possible during the event.

C. Minimizes radiation release to the environment via the affected SIG Main Steam Safety valves.

D. Prevents dilution of the RCS by maintaining SIG pressure lower than RCS pressure.

Answer: C

Answer Explanation:

A. Incorrect-Plausible to the Operator since lowering of temperature is done in EOP-5 to support the depressurization of the RCS. Per the EOP-6 Technical Basis document: The initial cooldown is done prior to isolating the affected SIG. This action reduces the risk of challenging the steam generator safety valves of the affected SIG after it is isolated.

B. Incorrect-Plausible to the Operator since natural circ conditions will exist due to the LOOP. A cooldown to 515F is not necessary to establish natural circulation conditions

C. Correct-Per the EOP-6 Technical Basis document: The initial cooldown is done prior to isolating the affected SIG. This action reduces the risk of challenging the steam generator safety valves of the affected SIG after it is isolated.

D. Incorrect-Plausible to the Operator since uncontrolled dilution is always a concern when shutdown. EOP-6 accounts for the potential flow from the SIG to the RCS by requiring additional boron/SOM. Backflow from the SIG to the RCS is an available method for controlling affected SIG level.




Svstem ID: 92909 User-Defined ID: L92909 Cross Reference Number:

Topic: Basis for cooldown to < 515F prior to isolating affected SIG RO Importance: 3.8 SRO Importance: 4.3 Cognitive level (High or


Tier/Group 3/NA

KIA Info 2.3 Radiation Control 2.3.11 Ability to control radiation releases

RO 3.8 Importance

Technical EOP-6, Steam Generator Tube Rupture References


Learning Recall the strategy and the basis for the Objective major actions performed in EOP-6, Steam

Generator Tube Rupture, and what actions are required if safety functions are in jeopardy of being lost





Comments None



!Task 1201.065 Identify, Isolate and Confirm the affected Steam Generator I

ID: .. Q11)6594··· .... Points: 1.00


• 100% power • Steam Generator Blowdown (SGBD) is overboard at 100 GPM per Steam Generator


• Condenser Off-Gas RMS alarms on Plant Computer • Letdown lowers from 38 GPM to 35 GPM and is steady at 35 GPM • AOP-2A is implemented for a Steam Generator Tube Leak

Which ONE of the following 1) locations should Radiation Safety be monitoring for potential radiological changes and 2) why?

OPERATIONS

A. 1) 5' East Penetration Room. 2) SGBD IX flowrate will be increased.

B. 1) Water Treatment Area. 2) Condensate flow through the Demineralizers will be maximized.

C. 1) Aux Building Roof. 2) ADVs will open when the manual trip is inserted.

D. 1) 45' Auxiliary Building near the SGBD Tank. 2) SGBD will be maximized.

Answer: B

Page: 221 of 235 13 April 2016

OPERATIONS


Answer Explanation:

A. Incorrect 1) False. Radiation surveys in the 5' East Penetration Room is plausible to the Operator since this area contains SGBD piping. If SGBD were to be maximized, then this area could potentially see radiation level increases. However, AOP-2A has a step to secure SGBD. 2) False. Maximizing SGBD flow is plausible to the Operator since these are actions if the SIG were to be experience high secondary contamination (such as Chlorides, Sodium, etc.) The conditions given indicate a SIG Tube Leak exists and is within the capacity of a Charging Pump. AOP-2A has a step to secure SGBD.

B. Correct 1) True. AOP-2A has a step to direct Radiation Safety to begin taking surveys in the Turbine Building and the Water Treatment Area. This will ensure Radiologically Controlled Areas are established as necessary. 2) True.The conditions given indicate a SIG Tube Leak exists and is within the capacity of a Charging Pump. This is determined by letdown flow lowering from 38 to 35 GPM. AOP-2A directions would include maximizing flow through the Condensate demineralizers.

C. Incorrect 1) False. AOP-2A does not have a step to direct Radiation Safety to begin taking surveys on the Aux Building roof, just the Turbine Building and the Water Treatment Area. Surveys on the Aux Building roof are plausible if the Operator believes the reactor will be manually tripped and opening of the ADVs is possible on the trip. However, a controlled shutdown would be performed for a SGTR within the capacity of a Charging Pump and a manual trip would not be inserted. 2) False. The ADVs opening is plausible since these actions are possible for a SIG tube leak exceeding the capacity of a charging pump. The conditions given indicate a SIG Tube Leak exists and is within the capacity of a Charging Pump. This is determined by letdown flow lowering from 38 to 35 GPM. Since the leakrate is within the capacity of Charging Pump, only a controlled shutdown would be performed and a manual trip would not be inserted. The ADVs would not open for a controlled shutdown.

D. Incorrect 1) False. Radiation surveys in the 45' SGBD Area is plausible to the Operator since this area contains SGBD piping and the SGBD tank. If SGBD were to be maximized, then this area could potentially see radiation level increases. However, AOP-2A has a step to secure SGBD. 2) False. Maximizing SGBD flow is plausible to the Operator since these are actions if the SIG were to be experience high secondary contamination (such as Chlorides, Sodium, etc.) The conditions given indicate a SIG Tube Leak exists and is within the capacity of a Charging Pump. For these conditions, AOP-2A has a step to secure SGBD.

Page 222 of 235 13 April 2016




Topic: Radiation Hazards Associated with a Steam Generator Tube Leak



Tier/Group 3/NA

KIA Info 2.3 Radiation Control 2.3.14 Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or emergency conditions or activities

RO 3.4 Importance



Learning Given plant conditions indicating an RCS Objective leak, the license candidate shall quantify

leakage then direct and/or implement the applicable actions to mitigate the event in progress per AOP-2A

Question New Source






Comments Considered a KIA match since Operator must know location where radiation hazards will develop when directing Radiation Safety to respond

Task 202.015 Respond to RCS leakage exceeding capacity of one charqinq pump, modes 1 & 2


73


ID: Q106613 Points: 1.00

Unit-2 was operating at 100% power when the following transient occurred:

• Reactor was manually tripped • Multiple events are occurring • EOP-8 was implemented • You were assigned to identify all the success paths except for Reactivity and Pressure and

Inventory Control • VA-1 and CE-3 were identified as Met • HR-2 and RLEC-2 were identified as Not Met

Which ONE of the following is when you should first notify the Unit Supervisor of success path status?

A When VA-1 was identified as Met.

B. When HR-2 was identified as Not Met.

C. When RLEC-2 was identified as Not Met.

D. When CE-3 was identified as Met.

Answer: B

Answer Explanation:

A lncorrect-EOP-8 Success Paths would be evaluated in their hierarchy order-VA then HR then CE then RLEC. Per OP-CA-103-102-1001, the US is informed of success path status as soon as a success path is identified as Not Met. HR-2 would be the first success path identified as Not Met. VA-1 is plausible to the Operator since Safety Functions are reported as they are completed in EOP-0.

B. Correct-EOP-8 Success Paths would be evaluated in their hierarchy order-VA then HR then CE then RLEC. Per OP-CA-103-102-1001, the US is informed of success path status as soon as a success path is identified as Not Met. HR-2 would be the first success path identified as Not Met.

C. lncorrect-EOP-8 Success Paths would be evaluated in their hierarchy order-VA then HR then CE then RLEC. PerOP-CA-103-102-1001, the US is informed of success path status as soon as a success path is identified as Not Met. HR-2 would be the first success path identified as Not Met. CE-3 is plausible to the Operator since it is the highest numbered Success Path out of VA, HR, CE, and RLEC.

D. lncorrect-EOP-8 Success Paths would be evaluated in their hierarchy order-VA then HR then CE then RLEC. Per OP-CA-103-102-1001, the US is informed of success path status as soon as a success path is identified as Not Met. HR-2 would be the first success path identified as Not Met. RLEC-2 is plausible to the Operator since RLEC is Not Met.





Topic: EOP-8 Roles and Responsibilities RO Importance: 4.0 SRO Importance: 4.6 Cognitive level (High or

H Low):



NRC KIA Info, References: Tier/Group 3/NA

KIA Info 2.4 Emergency Procedures/Plan 2.4.13 Knowledge of crew roles and responsibilities during EOP usage

RO 4.0 Importance

Technical OP-CA-103-102-1001 (Strategies for References Successful Transient Mitigation)


Learning Given plant conditions recognize the success Objective paths and order of their priority

Question New Source




Comments None

Task 201.090 Establish Core and RCS Heat Removal by SIG Heat Sink with SIS operation (HR-2)


74


ID: Q106615 Points: 1.00


• Reactor tripped on Loss of Load • Several Main Steam Safety Valves lifted • U-1 MAIN STM PIPING RM 315 Fire Alarm is received on 1C24B

Which ONE of the following actions should be taken during EOP-0 and why?

A Dispatch the Fire and Safety Watch to verify a fire does not exist and prevent an unnecessary Emergency Action Level declaration.

B. Dispatch the Operations Technical Advisor to open the Main Steam Penetration Room Watertight doors to clear the Fire Alarm.

C. Dispatch the Fire Brigade Leader to establish an Hourly Fire Watch to meet TRM requirements.

D. Dispatch the Auxiliary Building Operator to secure the MSIV ventilation fans to limit fire impacts.

Answer: A


OPERATIONS


Answer Explanation:

A Correct-When the Main Steam Safety valves lift, Operating Experience has shown that the MSIV Room Fire Alarm (Window SP-54) will actuate due to the small amount of steam emitted in the room (as a result of the Safety Valve design.) Per 1C24B ALM, response actions include dispatching an operator and/or the Fire and Safety Watch to investigate the cause. Since the alarm is likely caused by humidity change in room given the plant conditions, evaluating the room is free of fire effects will ensure the Shift Manager does not unnecessarily declare an Unusual Event for a Fire not extinguished within 15 minutes (as a result of a fire alarm that exists for >15 minutes but fire conditions were not evaluated locally within 15 mins.)

B. Incorrect-When the Main Steam Safety valves lift, Operating Experience has shown that the MSIV Room Fire Alarm (Window SP-54) will actuate due to the small amount of steam emitted in the room (as a result of the Safety Valve design.) Per 1C24B ALM, response actions include dispatching an operator and/or the Fire and Safety Watch to investigate the cause. Dispatching the Operations Technical Advisor is plausible to the Operator since these are actions for an actual fire condition. The Operations Technical Advisor would not be dispatched until an actual fire condition existed and ERPIP-3.0 was activated for a real fire. Opening the watertight doors is plausible to the Operator since the alarm is likely caused by humidity change in room given the plant conditions, and opening the doors will restore conditions to clear the alarm. However, opening the doors, which would include TRM entry, would not be taken during EOP-0.

C. Incorrect-When the Main Steam Safety valves lift, Operating Experience has shown that the MSIV Room Fire Alarm (Window SP-54) will actuate due to the small amount of steam emitted in the room (as a result of the Safety Valve design.) Per 1C24B ALM, response actions include dispatching an operator and/or the Fire and Safety Watch to investigate the cause. Dispatching the Fire Brigade Leader is plausible to the Operator since these are actions for an actual fire condition. The Fire Brigade Leader would not be dispatched until an actual fire condition existed and ERPIP-3.0 was activated for a real fire. Establishing an hourly fire watch is plausible to the Operator since these are actions associated with the TRM for a hanging fire alarm. However, these actions would not be taken in EOP-0.

D. Incorrect-When the Main Steam Safety valves lift, Operating Experience has shown that the MSIV Room Fire Alarm (Window SP-54) will actuate due to the small amount of steam emitted in the room (as a result of the Safety Valve design.) Per 1C24B ALM, response actions include dispatching an operator and/or the Fire and Safety Watch to investigate the cause. Securing the MSIV ventilation fans is plausible to the Operator since these are actions for an actual fire condition. However, securing ventilation would not be performed until an actual fire condition existed and ERPIP-3.0 was activated for a real fire.

Page: 229 of 235 13 April 2016




Topic: Response to MSIV Room Fire Alarm RO Importance: 3.3 SRO Importance: 3.7 Cognitive level (High or

H Low):



NRG KIA Info, References: Tier/Group 3/NA

KIA Info 2.4 Emergency Procedures/Plan 2.4.25 Knowledge of fire protection procedures

RO 3.3 Importance

Technical 1 C24B ALM (Fire Systems Alarm Manual), References ERPIP-3.0 (Immediate Actions)

Ol-32A (Auxiliary Feedwater System)


Learning Respond to a fire in the plant per ERPIP-3.0 Objective Question New Source




Comments None

Task 323.005 Investigate the source of a fire alarm: specifically smoke detector, duct detector, pull station, or water flow alarm


75


ID: Q106635 P~ints:.1.00

The following conditions have occurred:

• An Emergency Action Level was declared at 0200 • The ERO Notification System (ERONS) was activated at 0205 • The Shift Manager provided you with the Initial Notification Form at 0210 and directed you to

notify the State and Local agencies • The dedicated Offsite Agency Telephone is not working

Which ONE of the 1) communication systems should be used and 2) what is the latest time the State and Local agencies should be notified?

A. 1) Any available phone 2)0215

B. 1) Any available phone 2)0225

C. 1) Radio Communications Console 2) 0215

D. 1) Radio Communications Console 2)0225

Answer: A


OPERATIONS


Answer Explanation:

A. Correct-1) Per EP-CE-114-100, if the dedicated Offsite Agency Telephone is not available, then the next preferred method of communications is the use of any available operating phone. 2) State and Local agencies must be notified within 15 mins of a declaration. Since the declaration was at 0200, the agencies must be notified by 0215.

B. lncorrect-1) Per EP-CE-114-100, if the dedicated Offsite Agency Telephone is not available, then the next preferred method of communications is the use of any available operating phone. 2) State and Local agencies must be notified within 15 mins of a declaration. Since the declaration was at 0200, the agencies must be notified by 0215. 0225 is plausible to the Operator if the 15 min timeline is added to the time when the Initial Notification Form was given to the Communicator.

C. lncorrect-1) Per EP-CE-114-100, if the dedicated Offsite Agency Telephone is not available, the next preferred method of communications is any available operating phone. Using the Radio Communications Console is plausible to the Operator since this is the preferred method after all phone systems are unavailable. 2) State and Local agencies must be notified within 15 mins of a declaration. Since the declaration was at 0200, the agencies must be notified by 0215.

D. lncorrect-1) Per EP-CE-114-100, if the dedicated Offsite Agency Telephone is not available, the next preferred method of communications is any available operating phone. Using the Radio Communications Console is plausible to the Operator since this is the preferred method after all phone systems are unavailable. 2) State and Local agencies must be notified within 15 mins of a declaration. Since the declaration was at 0200, the agencies must be notified by 0215. 0225 is plausible to the Operator if the 15 min timeline is added to the time when the Initial Notification Form was given to the Communicator.

Page: 233 of 235 13 April 2016




Topic: Initial Notification Timelines and Backup Communication Systems


L Low):




KIA Info 2.4 Emergency Procedures/Plan 2.4.43 Knowledge of emergency communications systems and techniques

RO 3.2 Importance

Technical EP-CE-114-100 (Emergency Notifications) References


Learning Notify offsite agencies of Emergency Objective Classification

Question New Source




Comments None

Task 204.111 Perform duties as Emergency Communicator


1

EXAMINATION ANSWER KEY LOI 2015 NRC SRO Exam (June 2016)

ID: Q106535 Points: 1.00

Unit-1 is at 50% power when the following occurs:

• Large Break LOCA • Loss of Offsite Power • 1 A DG fails to start • OC DG is emergency started • 12 SRW Pump trips immediately on overload but is not noted during EOP-0 • 13 SRW Pump does not automatically start due to a 12 SRW Pump breaker switch

malfunction • EOP-5 is implemented • Containment Pressure is 30 PSIG and steady • Containment Temperature is 240°F and very slowly rising

Which ONE of the following directions should be given to an Auxiliary Operator and 2) why?

A. 1) Isolate SRW to 13 and 14 Containment Air Coolers. 2) To allow the start of 13 SRW Pump and restore cooling to the 1 B DG.

B. 1) Throttle SRW to 13 and 14 Containment Air Coolers. 2) To allow the start of 13 SRW Pump and a controlled restoration of Containment Cooling.

C. 1) Locally trip the 1 B DG fuel racks. 2) To prevent damage to the 1 B DG.

D. 1) Throttle SRW to 11 and 12 Containment Air Coolers and align OC DG disconnects to 11 4Kv bus. 2) To allow the start of 11 SRW Pump without waterhammer occurring in the SRW system.

Answer: A

Answer Explanation:

A. Correct 1) True. On a LOCA/LOOP with loss of 1 A DG and 12 SRW Pump failure to start, there are no SRW pumps operating. With SRW idle and a large LOCA occurring, a potential voiding concern exists with the CACs. Since Containment Pressure has exceeded 25 PSIG with SRW idle, voids in both SRW headers are possible. EOP-5 directs that if CSAS has actuated (which it will at >4.25 PSIG) and 11 SRW header is idle, 11 SRW pump is placed in PTL and the TSC consulted for guidance on restoration of the SRW header. If 12 header is also idle with high Containment Pressure, then SRW is isolated to 13/14 CACs and 13 SRW is started to at least restore cooling to the 1 B DG. 2) True. If 12 header is also idle with high Containment Pressure, then SRW is isolated to 13/14 CACs and 13 SRW is started to at least restore cooling to the 18 DG.


OPERATIONS


B. Incorrect 1) False -On a LOCA/LOOP with loss of 1A DG and 12 SRW Pump failure to start, there is no SRW pumps operating. With SRW idle and a large LOCA occurring, a potential voiding concern exists with the CACs. Since Containment Pressure has exceeded 25 PSIG with SRW idle, voids in both SRW headers are possible. EOP-5 directs that if CSAS has actuated (which it will at >4.25 PSIG) and 11 SRW header is idle, 11 SRW pump is placed in PTL and the TSC consulted for guidance on restoration of the SRW header. If 12 header is also idle with high Containment Pressure, then SRW is isolated to 13/14 CACs and 13 SRW is started to at least restore cooling to the 1 B DG. Throttling SRW is plausible if Operator believes systematic recovery of CACs is similar to recovery of RCPs ( when lower seal temperatures exceed 280F) to prevent thermal shock. 2) False. Restoration of SRW to prevent waterhammer is plausible since this is similar action to that taken when restoring CCW to the RCPs. Controlled restoration of containment cooling is plausible since containment parameters indicate that restoration of Containment Cooling would lower pressures and temperatures.

C. Incorrect 1) False. On a LOCA/LOOP with loss of 1A DG and 12 SRW Pump failure to start, there is no SRW pumps operating. With SRW idle and a large LOCA occurring, a potential voiding concern exists with the CACs. Since Containment Pressure has exceeded 25 PSIG with SRW idle, voids in both SRW headers are possible. EOP-5 directs that if CSAS has actuated (which it will at >4.25 PSIG) and 11 SRW header is idle, 11 SRW pump is placed in PTL and the TSC consulted for guidance on restoration of the SRW header. If 12 header is also idle with high Containment Pressure, then SRW is isolated to 13/14 CACs and 13 SRW is started to at least restore cooling to the 1 B DG. Tripping the fuel racks is plausible if Operator does not believe that SRW pumps can operate below minimum flow at all (with CACs isolated) and that the DG must be secured since cooling won't be restored. EOP-5 has a note that allows the restoration of flow to only the DG in order to maintain electrical power. 2) False. Preventing damage is plausible if the Operator believes no SRW flow can be established to cool the DG.

D. Incorrect 1) False. On a LOCA/LOOP with loss of 1A DG and 12 SRW Pump failure to start, there is no SRW pumps operating. With SRW idle and a large LOCA occurring, a potential voiding concern exists with the CACs. Since Containment Pressure has exceeded 25 PSIG with SRW idle, voids in both SRW headers are possible. EOP-5 directs that if CSAS has actuated (which it will at >4.25 PSIG) and 11 SRW header is idle, 11 SRW pump is placed in PTL and the TSC consulted for guidance on restoration of the SRW header. If 12 header is also idle with high Containment Pressure, then SRW is isolated to 13/14 CACs and 13 SRW is started to at least restore cooling to the 1 B DG. Plausible if Operator considers options to restore power to 11 4Kv bus, but EOP-5 does not direct this as an action. 2) False. Restoration of SRW by throttling flow to the CACs is plausible since this systematic response is similar to restoring CCW to an RCP with elevated temperatures. However, EOP-5 directs that if CSAS has actuated (which it will at >4.25 PSIG) and 11 SRW header is idle, 11 SRW pump is placed in PTL and the TSC consulted for guidance on restoration of the SRW header.

Page: 2 of 80 13 April 2016




Topic: SRO ONLY CAC Strategy During LOCA with no SRW RO Importance: 3.8 SRO Importance: 4.0 Cognitive level (High or


Tier/Group 1 /1

KIA Info EPE011 La~eBreakLOCA 2.4 Emergency Procedures I Plan 2.4.35 Knowledge of local auxiliary operator tasks during an emergency and the resultant operational effects

SRO 4.0 Importance

Technical EOP-5, EOP-5 Technical Bases References


Learning Given a value of containment pressure, Objective determine if a SRW header may be returned

to service

Question New Source




Comments None


EXAMINATION ANSWER KEY LOI 2015 NRG SRO Exam (June 2016)

1

201.004 Direct the implementation of Optimal I Recovery Procedures (EOPs 1-7)

Unit-1 is performing a coast down per PSTP-10 with the following conditions:

• RFO to begin in 2 days • Reactor Power is 99% and very slowly lowering • RCS Tcold is 539°F and steady

A transient occurs with the following indications:

• RCS temperatures are lowering • AOP-7K is implemented

Which ONE of the following is (1) when a Reactor Trip would be ordered and (2) the expected plant response during the transient?

OPERATIONS

A (1) Trip the reactor if Tavg reached 537°F (2) Power would be lowering before the trip and RCS pressure would be lower than normal in EOP-0

B. (1) Trip the reactor if Tcold reached 525°F (2) Power would be rising before the trip and PZR level would be higher than normal in EOP-0

C. ( 1) Trip the reactor if Tcold reached 520°F (2) Power would be lowering before the trip and SG Pressures would be lower than normal in EOP-0

D. (1) Trip the reactor if Tcold reached 515°F (2) Power would be rising before the trip and subcooling would be higher than normal in EOP-0

Answer: D


OPERATIONS


Answer Explanation:

A lncorrect-1 )Trip criteria in AOP-7K includes Tcold lowering to 51 SF or a Reactor trip is imminent. A Tavg of 537 with power >90% will result in Tcold approaching the 515F. The 537F mark is plausible since this is the trip criteria in AOP-2A during a SGTR. 2)Power would initially be rising before the trip occurred. The core is at EOC based on information that RFO will be starting in two days, resulting in a large -MTC. As RCS temperature lowers due to the leak in Containment, Reactor Power will begin to rise because of the -MTC. Power lowering is a plausible answer if the Operator mistakes a shutdown for an RFO with a startup from an RFO and its impact on MTC. A startup from a RFO will have a +MTC, and power would lower as temperature was lowered. Following the trip, RCS temperature will continue to lower since a steam leak into containment will continue to remove heat from the RCS. Temperature impact on SCM will be larger than lowering pressure and SCM will be larger than normal in EOP-0.

B lncorrect-1 )Trip criteria in AOP-7K includes Tcold lowering to 51 SF or a Reactor trip is imminent. A Tcold of 525F as the trip parameter is plausible if the Operator utilizes the 525F lower limit of EOP-0 as the trigger. 2)Power would initially be rising before the trip occurred. The core is at EOC based on information that RFO will be starting in two days, resulting in a large -MTC. As RCS temperature lowers due to the leak in Containment, Reactor Power will begin to rise because of the -MTC. Following the trip, RCS temperature will continue to lower since a steam leak into containment will continue to remove heat from the RCS, resuling in lowering PZR level. Rising PZR level is plausible if the Operator believes that SIAS actuation on high Containment pressure would start all 3 Charging Pumps and isolate letdown. The temperature impact would have a greater impact than the capacity of eves, so PZR level would be lower than normal in EOP-0.

C. lncorrect-1 )Trip criteria in AOP-7K includes Tcold lowering to 51 SF or a Reactor trip is imminent. A Tcold of 520F as the trip parameter is plausible if the Operator believes a trip is imminent. 2)Power would initially be rising before the trip occurred. The core is at EOC based on information that RFO will be starting in two days, resulting in a large -MTC. As RCS temperature lowers due to the leak in Containment, Reactor Power will begin to rise because of the -MTC. Power lowering is a plausible answer if the Operator mistakes a shutdown for an RFO with a startup from an RFO and its impact on MTC. A startup from a RFO will have a +MTC, and power would lower as temperature was lowered. Following the trip, RCS temperature will continue to lower since a steam leak into containment will continue to remove heat from the RCS, resulting in lowering RCS temperature and lowering SG pressures.

D. Correct-1 )Trip criteria in AOP-7K includes Tcold lowering to 51 SF. 2)Power would initially be rising before the trip occurred. The core is at EOC based on information that RFO will be starting in two days, resulting in a large -MTC. As RCS temperature lowers due to the leak in Containment, Reactor Power will begin to rise because of the -MTC. Following the trip, RCS temperature will continue to lower since a steam leak into containment will continue to remove heat from the RCS.





Topic: SRO ONLY AOP-7K Trip Criteria RO Importance: 4.6 SRO Importance: 4.7 Cognitive level (High or H Low): NRC KIA Info, References:

Tier/Group 1/1

KIA Info APE 040 Steam Line Rupture AA2. Ability to determine and interpret the following as they apply to the Steam Line Rupture: AA2.02 Conditions requiring a reactor trip

SRO 4.7 Importance

Technical AOP-7K References


Learning Given an overcooling event in progress, Objective determine and implement the correct actions

to mitigate the event in accordance with plant operating procedures

Question New Source






Comments None

Task 202.102 Respond to an Overcooling Event in Modes 1 or 2

ID: Q106~41 ' PQints: 1.00

Unit-1 is operating at 100% when the following occurs:

• Loss of Offsite Power • 11 and 12 MSIVs are shut during Turbine Trip • 13 AFW Pump is started during Core and Heat Removal

Which ONE of the following is the expected loop temperature indications and ADV response during the EOP-0 Wrap-Up Brief?

OPERATIONS

A. Tcold is 517°F, Thot is 562°F, and only one ADV indicates intermediate.

B. Tcold is 528°F, Thot is 553°F, and both ADVs indicate intermediate.

C. Tcold is 530°F, Thot is 575°F, and both ADVs are being operated locally and 9 turns open.

D. Tcold is 532°F, Thot is 534°F, and the ADV Controller, 1-HIC-4056, is in Manual with an output of 30%.

Answer: B


OPERATIONS


Answer Explanation:

A Incorrect-On a trip with natural circulation established, D-T with both SGs available to remove heat is expected to be 20-25F. With the TBVs shut due to loss of power to the MSRs, the ADVs will modulate to control Tavg based on the RRS input. With Tcold of 528F and a That of 553F (reflecting normal simulator response), Tavg will be 540F and the ADVs would be partially open to control temperature. A Tcold of 517F and That of 562F is plausible if Operator believes that ADV controller responds like TBV controller (where one valve fully opens before the next valve starts to open) and the D-T associated with only a single SG as the heat sink (since only one ADV is open) would be 40-50F.

B. Correct-On a trip with natural circulation established, D-T with both SGs available to remove heat is expected to be 20-25F. With the TBVs shut due to loss of power to the MSRs, the ADVs will modulate to control Tavg based on the RRS input. With Tcold of 528F and a That of 553F (reflecting normal simulator response), Tavg will be 540F and the ADVs would be partially open to control temperature.

C. Incorrect-On a trip with natural circulation established, D-T with both SGs available to remove heat is expected to be 20-25F. With the TBVs shut due to loss of power to the MSRs, the ADVs will modulate to control Tavg based on the RRS input. A Tcold of 530F and That of 575F is plausible if Operator believes that 40-50F is expected D-T rather than 20-25F even though both SGs are available for heat removal. Operating the ADVs locally is plausible if the Operator believes the ADV controller will not be repowered following the LOOP. Since the controller is powered from 1Y09, the controller will momentarily deenergize on the LOOP but repower once the 1A DG reenergizes the 11 4Kv bus. 9 turns locally open represents 30% open.

D. Incorrect-On a trip with natural circulation established, D-T with both SGs available to remove heat is expected to be 20-25F. With the TBVs shut due to loss of power to the MSRs, the ADVs will modulate to control Tavg based on the RRS input. A Tcold of 532F and That of 534F is plausible if Operator remembers normal D-T with RCPs still in service. With RCPs off due to the LOOP, D-T is expected to be 20-25F. The ADV controller output is expected to be high enough to ensure that both ADVs are modulating temperature, where 30% is a typical simulator response.





Topic: SRO ONLY Temperature Response on LOOP RO Importance: 4.3 SRO Importance: 4.3 Cognitive level (High or


Tier/Group 1/1

KIA Info APE 056 Loss of Offsite Power AA2. Ability to determine and interpret the following as they apply to the Loss of Offsite Power: AA2.32 Transient trend of coolant temperature toward no-load T-ave

SRO 4.3 Importance

Technical EOP-0, EOP-2 References


Learning Recall the core and plant parameters Objective response to a Loss of Forced Circulation

(LOFC)

Question New Source






Comments None


10: Qt,,4~78 Points: 1'.00

Given the following conditions on Unit-1:

• 1C13, K-19, 11 SRW HEAD TK LVL alarm is received and momentarily clears • 11 SRW Head Tank Level Control Valve, 1-SRW-1579-CV, is intermittently opening and

closing • 1C10-ALM, J-21, CNTMT NORMAL SUMP LVL HI alarm is received • The crew enters the appropriate procedure to address the condition

Which ONE of the following describes 1) the source of the leak, and 2) the required actions in accordance with procedures and Technical Specifications?

OPERATIONS

A. 1) 11 or 12 CNTMT CLRs 2) Isolate one cooler at a time and determine Head Tank level response; Leave the leaking cooler isolated; NO Technical Specification action is required if only 1 CNTMT CLR is isolated.

B. 1) 11 or 12 CNTMT CLRs 2) Isolate BOTH coolers and verify Head Tank level stabilizes; Determine the leaking cooler by placing them in service one at a time and isolate the leaking cooler; ONE Train of CNTMT Cooling must be declared inoperable.

C. 1) 13 or 14 CNTMT CLRs 2) Isolate one cooler at a time and determine Head Tank level response, leave the leaking cooler isolated; NO Technical Specification action is required if only 1 CNTMT CLR is isolated.

D. 1) 13or14 CNTMT CLRs 2) Isolate BOTH coolers and verify Head Tank level stabilizes; Determine the leaking cooler by placing them in service one at a time and isolate the leaking cooler; ONE Train of CNTMT Cooling must be declared inoperable.

Answer: B

Page: 1 o of 80 13 April 2016

OPERATIONS


Answer Explanation:

A lncorrect-11 SRW Head Tank Level is affected and is associated with 11 SRW Header which supplies 11/12 Cntmt Clrs. AOP-7B directs isolation of the coolers as a pair, unisolating one at a time. Isolating one cooler at a time is plausible if Operator believes Ol-5A will be used in parallel to remove the cooler since Ol-5A removes the CAC one at a time. Removal of one cooler removes a TS Train from service. Not entering TS LCO is plausible if Operator believes both 11/12 Cooler must be removed before train is declared OOS since minimum safety function for containment cooling is still met with 3 CACs remaining in service.

B. Correct-11 SRW Head Tank Level is affected and is associated with 11 SRW Header which supplies 11 /12 Cntmt Clrs. AOP-7B directs isolation of the coolers as a pair, unisolating one at a time. Isolation of one cooler removes a TS Train from service.

C. lncorrect-11 SRW Head Tank Level is affected and is associated with 11 SRW Header which supplies 11/12 Cntmt Clrs. 13/14 Coolers are plausible if Operator believes U-1 Service Water is cross connected (like U-2 in the Turbine Building) and believes 13/14 Coolers could be affected (since Turbine Building isolation CVs are still open.) AOP-7B directs isolation of the coolers as a pair, unisolating one at a time. Isolating one cooler at a time is plausible if Operator believes Ol-5A will be used in parallel to remove the cooler since Ol-5A removes the CAC one at a time. Removal of one cooler removes a TS Train from service. Not entering TS LCO is plausible if Operator believes both 11/12 Cooler must be removed before train is declared OOS since minimum safety function for containment cooling is still met with 3 CACs remaining in service.

D. lncorrect-11 SRW Head Tank Level is affected and is associated with 11 SRW Header which supplies 11 /12 Cntmt Clrs. 13/14 Coolers are plausible if Operator believes U-1 Service Water is cross connected (like U-2 in the Turbine Building) and believes 13/14 Coolers could be affected (since Turbine Building isolation CVs are still open.) AOP-7B directs isolation of the coolers as a pair, unisolating one at a time. Removal of one cooler removes a TS Train from service.

Page: 11 of 80 13 April 2016




Topic: SRO ONLY DiaQnosis of and response to a CAC SRW leak RO Importance: 4.3/2.9 SRO Importance: 4.4/3.5 Cognitive level (High or


Tier/Group 1/1

KIA Info APE 062 Loss of Nuclear Service Water 2.1 Conduct of Operations 2.1.23 Ability to perform specific system and integrated plant procedures during all modes of plant operation

SRO 4.4 Importance

Technical AOP-7B, TS 3.6.6 References


Learning Differentiate between a loss of Service water Objective flow and a Service Water leak (rupture)


Question NRC 2008 SRO Exam History



Comments None


.5


I Task 1202.065 Respond to a loss of Service Water I in modes 1 or 2

Additional KIA Match Info APE 062 Loss of Nuclear Service Water AA2. Ability to determine and interpret the following as they apply to the Loss of Nuclear Service Water: AA2.01 Location of a leak in the SWS


Cognitive Level

COMPREHENSION

JQ: Q,106546 Points: 1.00

Unit-2 is operating at 100% power when the following occurs:

• 21 13Kv Feeder Breaker 252-2104 trips open • 2 CEAs remain full out after the plant trips and Alternate Actions for Reactivity are completed • EOP-1 is implemented • RCS temperature is 532°F and steady • PZR level is 180" and rising • ST A reports that the RCS Pressure and Inventory Safety Function is not currently being met

Which ONE of the following actions should be directed?

A Maintain all three Charging Pumps running and increase BIAS on the LID THROTTLE VLV CONTR, 2-HIC-110.

B. Secure the one running Charging Pump and allow the second Backup Charging Pump to cycle on deviation per Ol-2A.

C. Maintain one Charging Pump running and restore letdown per Ol-2A.

D. Immediately exit EOP-1 and implement EOP-8.

Answer: C


OPERATIONS


Answer Explanation:

A. lncorrect-Boration is required until RCS reaches 2300 PPM due to the two stuck CEAs. Ensuring one Charging Pump is running will ensure boration is in progress. Running all three Charging Pumps is plausible if Operator just considers Reactivity Alternate Actions of starting all available Charging Pumps. As PZR level rises above 160", the PZR level control system will secure the backup Charging Pumps on deviation. With the loss of 24 4Kv bus momentarily, 2Y10 was lost for -10 seconds and letdown isolated. To restore PZR level, letdown can be returned to service using Ol-2A and this will restore PZR level trend back to <180".lncreasing BIAS is plausible if the Operator did not recognize the loss of letdown and considered it to remain in service.

B. lncorrect-Boration is required until RCS reaches 2300 PPM due to the two stuck CEAs. Ensuring one Charging Pump is running will ensure boration is in progress. Allowing PZR level to cycle on deviation is plausible if Operator forgets that boration is maintained and utilizes steps in AOP-71 for a loss of 2Y10.

C. Correct-Boration is required until RCS reaches 2300 PPM due to the two stuck CEAs. Ensuring one Charging Pump is running will ensure boration is in progress. With the loss of 24 4Kv bus momentarily, 2Y10 was lost for -10 seconds and letdown isolated. To restore PZR level, letdown can be returned to service using Ol-2A and this will restore PZR level trend back to <180".

D. lncorrect-Boration is required until RCS reaches 2300 PPM due to the two stuck CEAs. Ensuring one Charging Pump is running will ensure boration is in progress. With the loss of 24 4Kv bus momentarily, 2Y10 was lost for -10 seconds and letdown isolated. To restore PZR level, letdown can be returned to service using Ol-2A and this will restore PZR level trend back to <180". Exiting EOP-1 is plausible if Operator considers not meeting the ISFSC for EOP-1 to be exit criteria. While this is procedurally allowed, the action would not be a proper use of resources and plant response would be further delayed with the implementation of EOP-8. OP-CA-103-102-1001 allows a 15-minute window to attempt to restore the PIC Safety Function to within limits and remain in EOP-1.

Page: 14 of 80 13 April 2016




Topic: SRO ONLY EOP-1 Actions for High PZR Level and 2 Stuck CEAs



Tier/Group 1/1

K/A Info CE EPE E02 Reactor Trip Recovery EA2. Ability to determine and interpret the following as they apply to the (Reactor Trip Recovery) EA2.2 Adherence to appropriate procedures and operation within the limitations in the facility*s license and amendments

SRO 4.0 Importance

Technical EOP-0, EOP-1, AOP-71, Ol-2A References


Learning Evaluate if the required boration is met from Objective EOP-0 based on CEA status and charging

pump(s) operating

Question New Source






Comments None

Task 201.019 Maintain RCS Boron concentration

ID: Q106556 .. Points: 1.0()'


• 11 Feed Reg Valve, 1-FW-1111-CV, failed shut • Plant automatically tripped on low SIG level • EOP-0 was implemented • Main Feedwater was lost due to a Condensate Header Rupture • 23 AFW Pump was OOS • Both 11 and 12 AFW Pumps tripped on overspeed and cannot be reset • 11 S/G level is -220" and lowering • 12 S/G level is -130" and lowering • The CRO reports motor amps for 13 AFW Pump are oscillating • The STA reports that AFW flow on SPDS indicates 0 GPM • 13 AFW Pump breaker indicates closed • Both Motor Train Flow Control Valves (1-AFW-4525-CV and 1-AFW-4535-CV) indicate full

open

Which ONE of the following actions 1) should be directed and 2) what EOP should be implemented?

OPERATIONS

A 1) Lower setpoint on both Motor Train Flow Control Valves and throttle shut both CVs 2) Implement EOP-6

B. 1) Place both Motor Train Flow Control Valves in Manual and throttle shut both CVs 2) Implement EOP-1.

C. 1) Monitor 11 /12 S/G level for positive trends 2) Implement EOP-8

D. 1) Place 13 AFW in PTL 2) Implement EOP-3

Answer: D

Page: 16 of 80 13 April 2016

OPERATIONS


Answer Explanation:

A lncorrect-1) Indications provided indicate AFW pump suction blockage. The Flow Control Valves are normally in Auto with a setpoint of 150 GPM. When AFW flow is established, the CVs will indicate intermediate as they throttle to control flow. With no AFW flow indicated on SPDS and CVs full open (trying to control flow at 150 GPM but can't since there is no actual flow), 13 AFW pump indications are of a cavitating pump. N0-1-200 allows for urgent actions to protect personnel or equipment, which in this case is to protect 13 AFW Pump and place it in PTL. Throttling shut the FCVs is plausible if the Operator believes that the FCVs being full open is the cause of 13 AFW cavitating (due to excessive flow.) The combination of lowering SG levels and no indicated AFW flow indicate that there is no actual AFW flow. 2) EOP-3 would be implemented since no feed is available. EOP-6 is plausible if the Operator believes the S/G level mismatch is due to a SGTR.

B. lncorrect-1) Indications provided indicate AFW pump suction blockage. The Flow Control Valves are normally in Auto with a setpoint of 150 GPM. When AFW flow is established, the CVs will indicate intermediate as they throttle to control flow. With no AFW flow indicated on SPDS and CVs full open (trying to control flow at 150 GPM but can't since there is no actual flow), 13 AFW pump indications are of a cavitating pump. N0-1-200 allows for urgent actions to protect personnel or equipment, which in this case is to protect 13 AFW Pump and place it in PTL. Throttling shut the FCVs is plausible if the Operator believes that the FCVs being full open is the cause of 13 AFW cavitating (due to excessive flow.) The combination of lowering SG levels and no indicated AFW flow indicate that there is no actual AFW flow. 2) EOP-3 would be implemented since no feed is available. EOP-1 is plausible if the Operator believes that EOP-0 Core and RCS Heat Removal is met since at least one SIG has a level of >-170". The EOP-0 Diagnostic Flowchart would direct EOP-3, however, when there is no feed available.

C. lncorrect-1) Indications provided indicate AFW pump suction blockage. The Flow Control Valves are normally in Auto with a setpoint of 150 GPM. When AFW flow is established, the CVs will indicate intermediate as they throttle to control flow. With no AFW flow indicated on SPDS and CVs full open (trying to control flow at 150 GPM but can't since there is no actual flow), 13 AFW pump indications are of a cavitating pump. N0-1-200 allows for urgent actions to protect personnel or equipment, which in this case is to protect 13 AFW Pump and place it in PTL. Monitoring S/G trends is plausible if Operator believes that 13 AFW pump is providing flow since the FCVs are full open. The combination of lowering SG levels and no indicated AFW flow indicate that there is no actual AFW flow. 2) EOP-3 would be implemented since no feed is available. EOP-8 is plausible since this is always an option if a condition is unknown or multiple events are in progress. EOP-3 would be the more efficient strategy versus the implementation of EOP-8.

D. Correct-1) Indications provided indicate AFW pump suction blockage. The Flow Control Valves are normally in Auto with a setpoint of 150 GPM. When AFW flow is established, the CVs will indicate intermediate as they throttle to control flow. With no AFW flow indicated on SPDS and CVs full open (trying to control flow at 150 GPM but can't since there is no actual flow), 13 AFW pump indications are of a cavitating pump. N0-1-200 allows for urgent actions to protect personnel or equipment, which in this case is to protect 13 AFW Pump and place it in PTL. 2) EOP-3 would be implemented since no feed is available.

Page: 17 of 80 13 April 2016




Topic: SRO ONLY Entry conditions for EOP-3 RO Importance: 4.5 SRO Importance: 4.7 Cognitive level (High or


Tier/Group 1 /1

KIA Info CE EPE E06 Loss of Feedwater 2.4 Emergency Procedures I Plan 2.4.4 Ability to recognize abnormal indications for system operating parameters that are entry-level conditions for emergency and abnormal operating procedures.

SRO 4.7 Importance

Technical EOP-0, N0-1-200 References


Learning Given plant conditions and the Diagnostic Objective Flowchart, determine the correct procedure

to implement following safety function assessment in EOP-0

Question New Source





7


Comments None

Task 201.002 Perform Diagnostics Actions of EOP-0

ID: Q106557


• 95% Power • End of Cycle

A plant transient occurs and the RO reports the following:

• Power has increased to 98% and appears to be stabilizing at 98% • Tcold has increased and is continuing to rise • Generator Output has increased and is continuing to rise

Which ONE of the following 1) events is in progress and 2) what procedure should be implemented?

A. 1) An uncontrolled rod withdrawal. 2) AOP-1 B, CEA Malfunction

B. 1) A steam line break. 2) AOP-7K.

C. 1) An excessive feed. 2) AOP-3G, Malfunction of Main Feedwater System.

D. 1) A loss of load. 2) AOP-7F, Loss of Load.

Answer: A


OPERATIONS


Answer Explanation:

A. Correct 1) As CEA withdraws, this will raise reactor power. Since steam demand has not initially changed, RCS temperatures would start to rise and this would raise pressures in both SGs. This would result in increasing steam flow through the Main Turbine and Main Generator output would increase. 2) CEA malfunctions are addressed by AOP-1 B.

B. Incorrect 1) A steam leak is plausible since this would result in power increasing due to the lowering temperature. However, per the initial conditions, temperature is rising. 2) AOP-7K is plausible since this is the appropriate procedure for a steam leak. Since a steam leak is not present, AOP-7K would not be correct.

C. Incorrect 1) An excessive feed is plausible since this would result in power increasing due to the lowering temperature. However, per the initial conditions, temperature is rising. 2) AOP-3G is plausible since this is the appropriate procedure for an excessive feed. Since an excessive feed is not present, AOP-3G would not be correct.

D. Incorrect 1) A loss of load is plausible since this would result in temperature increasing due to the lowering steam demand. However, a loss of load would not result in increased steam demand to the Main Turbine and Main Generator output would actually lower. 2) AOP-7F is plausible since this is the appropriate procedure for a loss of load. Since a loss of load is not present, AOP-7F would not be correct.

Page: 20 of 80 13 April 2016




Topic: SRO ONLY Continuous CEA Withdrawal Indications RO Importance: 4.4 SRO Importance: 4.6 Cognitive level (High or


Tier/Group 1/2

KIA Info APE 001 Continuous Rod Withdrawal AA2. Ability to determine and interpret AA2.05 Uncontrolled rod withdrawal, from available indications

SRO 4.6 Importance

Technical AOP-18, 1C05-ALM, USFAR Chapter 14 References


Learning Recognize the conditions which require a Objective Reactor trip during the implementation of

AOP-18

Question New Source




Comments None



Task 202.008 Respond to CEA(s) Misaligned by 15" or more


8


ID: 0106560 '.Points: 1.00

Unit-1 was operating at 100% when the following transient occurred:

• Reactor was manually tripped • All WRNI indications are lost

Which ONE of the following combinations ensure that Reactivity Control is met during the EOP?

A SUR is zero with One CEA NOT fully inserted.

B. SUR is negative with Two CEAs NOT fully inserted.

C. SUR is zero with Boration in progress using Two Charging Pumps.

D. SUR is negative with Boration in progress using One Charging Pump.

Answer: D

Answer Explanation:

A Incorrect-Per the EOP-0 SFSC, WRNI is expected to be lowering and SUR should be negative following a trip. A SUR of zero is plausible if the Operator applies the Final Safety Function parameters in EOP-1 which allows SUR to be negative or zero. However, EOP-0 requires that SUR be negative. Boration is not required since only one CEA is not fully inserted.

B. Incorrect-Per the EOP-0 SFSC, WRNI is expected to be lowering and SUR should be negative following a trip. Boration is required only if more than one CEA is stuck out. Two CEAs not fully inserted is plausible since RC-1 in EOP-8 allows multiple CEAs to remain stuck as long as WRNI power is <1 OE-4% and SUR is negative or zero. However, in EOP-0, boration must be in progress if more than one CEA fails to insert.

C. Incorrect-Per the EOP-0 SFSC, WRNI is expected to be lowering and SUR should be negative following a trip. A SUR of zero is plausible if the Operator applies the Final Safety Function parameters in EOP-1 which allows SUR to be negative or zero. However, EOP-0 requires that SUR be negative. Boration in progress would satisfy Reactivity regardless of CEA status.

D. Correct-Per the EOP-0 SFSC, WRNI is expected to be lowering and SUR should be negative following a trip. Boration in progress would satisfy Reactivity regardless of CEA status.





Topic: SRO ONLY Loss of WRNI Indications and Determining Reactivity Status in EOP-0



Tier/Group 1/2

KIA Info APE 032 Loss of Source Range Nuclear Instrumentation 2.4 Emergency Procedures I Plan 2.4.21 Knowledge of the parameters and logic used to assess the status of safety functions, such as reactivity control, core cooling and heat removal, reactor coolant system integrity, containment conditions, radioactivity release control, etc.

SRO 4.6 Importance



Learning Given a plant condition requiring a reactor Objective trip the license candidate will demonstrate an

understanding of the strategy, basis and operator actions of EOP-0 to direct or implement the procedural steps, including warnings, notes, and cautions

Question New Source



9




Comments None ·

Task 201.009 Verify the RC Safety Function is satisfied

P(>ints: 1.00


• Complete Core Off-Load to the Spent Fuel Pool in progress • Both doors of the Personnel Air Lock (PAL) are open • One door of the Emergency Air Lock (EAL) is shut • The PAL has an active Containment Closure Deviation Sheet • The Containment Outage Door (COD) is currently closed • The COD has an active Containment Closure Deviation Sheet

Which ONE of the following 1) would require entry into a LCO and 2) what would be the action required?

A 1) The PAL Watch enters Containment to inspect the EAL. 2) Verify Refueling Pool Water Level is 2'.. 23 ft above the top of the irradiated fuel assemblies seated in the Reactor Vessel immediately.

B. 1) The PAL Watch enters Containment to inspect the EAL. 2) Suspend movement of irradiated fuel assemblies within containment immediately.

C. 1) The COD is opened by the COD Watch. 2) Verify Refueling Pool Water Level is 2'.. 23 ft above the top of the irradiated fuel assemblies seated in the Reactor Vessel immediately.

D. 1) The COD is opened by the COD Watch. 2) Suspend movement of irradiated fuel assemblies within containment immediately.

Answer: B


OPERATIONS


Answer Explanation:

A. Incorrect- 1) TS 3.9.3 allows the PAL to be open under administrative controls during refueling. N0-1-200 and N0-1-114 provide those administrative controls, which includes a PAL door watch stationed at the PAL. If the Watch were to leave the PAL area, TS 3.9.3 LCO would not be met. 2) Condition A of TS 3.9.3 requires that movement of irradiated fuel in the containment stop immediately. This prevents loss of Containment and potential Fuel Handling Accident occurring at same time. Verifying RFP water level is plausible since this is a SR action in TS 3.9.6 that would also limit impact of a Fuel Handling Accident occurring at the same time.

B. Correct- 1) TS 3.9.3 allows the PAL to be open under administrative controls during refueling. N0-1-200 and N0-1-114 provide those administrative controls, which includes a PAL door watch stationed at the PAL. If the Watch were to leave the PAL area, TS 3.9.3 LCO would not be met. 2) Condition A of TS 3.9.3 requires that movement of irradiated fuel in the containment stop immediately. This prevents loss of Containment and potential Fuel Handling Accident occurring at same time.

C. Incorrect- 1) TS 3.9.3 allows the PAL to be open under administrative controls during refueling. N0-1-200 and N0-1-114 provide those administrative controls, which includes a PAL door watch stationed at the PAL. If the Watch were to leave the PAL area, TS 3.9.3 LCO would not be met. Opening the COD is plausible if the Operator believed the COD had to be closed during fuel moves instead of just having the ability to be closed. 2) Condition A of TS 3.9.3 requires that movement of irradiated fuel in the containment stop immediately. This prevents loss of Containment and potential Fuel Handling Accident occurring at same time. Verifying RFP water level is plausible since this is a SR action in TS 3.9.6 that would also limit impact of a Fuel Handling Accident occurring at the same time.

D. Incorrect- 1) TS 3.9.3 allows the PAL to be open under administrative controls during refueling. N0-1-200 and N0-1-114 provide those administrative controls, which includes a PAL door watch stationed at the PAL. If the Watch were to leave the PAL area, TS 3.9.3 LCO would not be met. Opening the COD is plausible if the Operator believed the COD had to be closed during fuel moves instead of just having the ability to be closed. 2) Condition A of TS 3.9.3 requires that movement of irradiated fuel in the containment stop immediately. This prevents loss of Containment and potential Fuel Handling Accident occurring at same time.

Page: 26 of 80 13 April 2016




Topic: SRO ONLY 1-Hour ContainmenURefueling TS Action Statements



Tier/Group 1/2

KIA Info APE 036 Fuel Handling Incidents 2.2 Equipment Control 2.2.39 Knowledge of less than or equal to one hour Technical Specification action statements for systems

SRO 4.5 Importance

Technical N0-1-200, N0-1-114. TS 3.9.3 References


Learning Recall the conditions and requirements for Objective performing the Core Alteration Checklist for

fuel movement

Question New Source



10CFR 55.43(b )(2) Part 55 Content



Comments None

Task 203.012 Direct the performance of core alterations in accordance with Refueling Procedure


10


ID: Q74956 Points: 1 ;OO

Unit-1 is at 100% power with the following conditions:

• Main Generator output is slowly lowering • T coLD remains stable at 547.8° F

Which ONE of the following reports from crew members would indicate the cause of the lowering Main Generator output and what procedure should be implemented?

A. Main Turbine Control Valve-1 indicates 40% open; Implement AOP-7E.

B. TBV 1-MS-3944-CV indicates 100% open; Implement AOP-7K.

C. Suction CV on an idle CAR is not fully shut; Implement AOP-7G.

D. Main Generator Cold Gas temperature is 51°C; Implement AOP-7E.

Answer: C

Answer Explanation:

A. Incorrect-Plausible since MTCV going shut will lower Main Generator output. However, MTCV normal position at 100% power is approximately 60%. If the valve closed to 40% open, T coLD would rise.

B. Incorrect-Plausible since a TBV opening would rob steam from the Main Turbine and Generator output would lower. However, a TBV failing open would cause T coLD to lower.

C. Correct-Indications of lowering vacuum are usually a lowering MWe output with no change in primary power/secondary steam demand. The open suction CV on an idle CAR has caused Condenser Vacuum to lower and is addressed in AOP-7G.

D. Incorrect-Main Generator Cold gas temperature rising is plausible since changing gas temperature could affect Generator efficiency. However, temperature only rising to 51 C will not change MWe load.





Topic: SRO ONLY AOP-7G, Loss of Condenser Vacuum. RO Importance: 2.4/4.5 SRO Importance: 2.7/4.7 Cognitive level (High or


Tier/Group 1/2

KIA Info APE 051 Loss of Condenser Vacuum AA2. Ability to determine and interpret the following as they apply to the Loss of Condenser Vacuum: AA2.01 Cause for low vacuum condition

SRO 2.7 Importance



Learning Given an Loss of or lowering Condenser Objective Vacuum, determine the correct actions to

mitigate the event in accordance with plant operating procedures






11


Comments None

Task 202.087 Respond to condenser vacuum reduction

Additional KIA Match Info 2.4 Emergency Procedures/Plan 2.4.4 - Ability to recognize abnormal indications of system operating parametes that are entry-level conditions for emergency and abnormal operating procedures.

, JD: Q106566


• Mode 6 • RCS Level 43' • 21 LPSI Pump operating • Shutdown Cooling (SDC) flow is 3000 GPM


• RCS level lowers to 36.8' and then steadies at 36.7' • The CRO reports that 21 LPSI is cavitating • The appropriate short-term actions are taken per the AOP

Which ONE of the following is the 1) impact to SDC following the short-term actions and 2) what recovery actions are required once RCS level is restored?

A. 1) SDC flow is lower but not lost. 2) Immediately raise flow to normal.

B. 1) SDC flow is lower but not lost. 2) Vent the SDC Return header before restoring flow to normal.

C. 1) SDC flow is lost. 2) Throttle SDC flow with the SDC TEMP CONTRO valve, 2-Sl-657-CV, before restarting a LPSI pump.

D. 1) SDC flow is lost. 2) Vent the LPSI Pump casings before restarting a LPSI pump.

Answer: D


OPERATIONS


Answer Explanation:

A Incorrect 1) False. AOP-38 directs that both LPSI Pumps be placed in PTL if level reaches 36.8'. With the LPSI pumps secured, SOC will be secured. Lower flow is plausible since AOP directs that flowrate be lowered to 1800 GPM if RCS level lowers below 37.6'. 2) True. Raising flowrate will be required once RCS level is restored to normal.

8. Incorrect 1) False. AOP-38 directs that both LPSI Pumps be placed in PTL if level reaches 36.8'. With the LPSI pumps secured, SOC will be secured. Lower flow is plausible since AOP directs that flowrate be lowered to 1800 GPM if RCS level lowers below 37.6'. 2) False. Venting the SOC header is plausible to the Operator since cavitation in the LPSI pumps was seen. Attachment 3 of AOP-38 will require the venting of the SOC header if air is suspected of being trapped in the SOC return header. However, the SOC return header is on the bottom of the 12 Hot leg. The bottom of the hot leg elevation is 35.58'. The conditions given indicate that level would have remained above 35.58' and air in the SOC header would not be expected.

C. Incorrect 1) True. AOP-38 directs that both LPSI Pumps be placed in PTL if level reaches 36.8'. With the LPSI pumps secured, SOC will be secured. 2) False. Throttling the flowpath is plausible since CV-306 is only 5% open when a LPSI pump is restarted. CV-657 is the temperature control valve, however, and this valve is initially shut when restoring flow. CV-306 is opened and then CV-657 is adjusted to restore temperature as desired.

0. Correct 1) True. AOP-38 directs that both LPSI Pumps be placed in PTL if level reaches 36.8'. With the LPSI pumps secured, SOC will be secured. 2) True. Attachment 3 of AOP-38 will require the venting of both LPSI pump casings if air is suspected in the pumps. Since the pump was cavitating, air is suspected, and the pump is vented before restarting it.

Page: 32 of 80 13 April 2016




Topic: SRO ONLY AOP-38 for LPSI Cavitation Due to RCS Level Lowerinq



Tier/Group 2/1

KIA Info 005 Residual Heat Removal System (RHRS) A2 Ability to (a) predict the impacts of the following malfunctions or operations on the RHRS, and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: A2.03 RHR pump/motor malfunction

SRO 3.1 Importance



Learning Differentiate between loss of SOC events Objective caused by:

• LPSI pump breaker trip, or • LPSI pump cavitation (loss of RCS level;

closure of one or both SOC return valves}, or

• Loss of component cooling and/or saltwater systems

Question New Source






Comments Considered SRO question since SRO would determine sections of AOP-38 that need to be performed to restore SDC.

Task 202.022 Respond to a complete loss of SDC with RCS open and the refueling pool not available



ID: 0106569 Points: 1.00


• Large Secondary transient caused by a loss of all Main Feedwater • At time 0100, SIG levels are -50" and lowering • Direction is given to the RO to trip the Reactor • The RO depresses both Manual REACTOR TRIP buttons on 1C05 • Reactor power is still -100%

Which ONE of the following 1) immediate actions should be taken and 2) if an Alert was declared at 0110, what is the latest time the NRC can be notified?

OPERATIONS

A. 1) Open the four blue-handled breakers on 1C17. 2) 0210

B. 1) Open the four blue-handled breakers on 1C17. 2)0500

C. 1) Deenergize 12 and 13 4Kv busses. 2) 0210

D. 1) Deenergize 12 and 13 4Kv busses. 2)0500

Answer: A

Page: 35 of 80 13 April 2016

OPERATIONS

EXAMINATION ANSWER KEY LOI 201S NRC SRO Exam (June 2016)

Answer Explanation:

A. Correct 1) True. Per 1 COS ALM 0-0S, if a PROT CH TRIP alarm is in and the Reactor did not trip (ATWS), EOP-0 is implemented. Per EOP-0 and OP-CA-103-102-1001, the RO is expected to immediately open the four blue handled breakers on 1C17 to deenergize the MG sets (by deenergizing the 12A and 13A 480V busses.) 2) True. An ATWS/Alert declaration at time 0110 means the NRC must be notified as soon as possible, but no later than 0210. EP-AA-114 requires the NRC be notified within an hour of the initial classification of an event.

B. Incorrect 1) True. Per 1 COS ALM 0-0S, if a PROT CH TRIP alarm is in and the Reactor did not trip (ATWS), EOP-0 is implemented. Per EOP-0 and OP-CA-103-102-1001, the RO is expected to immediately open the four blue handled breakers on 1C17 to deenergize the MG sets (by deenergizing the 12A and 13A 480V busses.) 2) False. 0500 is plausible if the Operator believes only the reportability timeline associated with LS-AA-1110 SAF1 .6 applies. The event occurred at 0100, and a 4-hour report would be required by OSOO. However, an ATWS/Alert declaration at time 0110 means the NRC must be notified as soon as possible, but no later than 0210. EP-AA-114 requires the NRC be notified within an hour of the initial classification of an event.

C. Incorrect 1) False. Oeenergizing the 12 and 13 4Kv busses is plausible since this will also deenergize the 12A and 13A 480V busses, which is same result as opening the four blue handled breakers. However, this this is not per guidance in EOP-0. 2) True. An ATWS/Alert declaration at time 0110 means the NRC must be notified as soon as possible, but no later than 0210. EP-AA-114 requires the NRC be notified within an hour of the initial classification of an event.

0. Incorrect 1) False. Deenergizing the 12 and 13 4Kv busses is plausible since this will also deenergize the 12A and 13A 480V busses, which is same result as opening the four blue handled breakers. However, this this is not per guidance in EOP-0. 2) False. OSOO is plausible if the Operator believes only the reportability timeline associated with LS-AA-1110 SAF1 .6 applies. The event occurred at 0100, and a 4-hour report would be required by OSOO. However, an ATWS/Alert declaration at time 0110 means the NRC must be notified as soon as possible, but no later than 0210. EP-AA-114 requires the NRC be notified within an hour of the initial classification of an event.

Page: 36 of 80 13 April 2016




Topic: SRO ONLY Directinq ATWS Actions RO Importance: 4.6 SRO Importance: 4.4 Cognitive level (High or


Tier/Group 2/1

KIA Info 012 Reactor Protection System 2.4 Emergency Procedures I Plan 2.4.49 Ability to perform without reference to procedures those actions that require immediate operation of system components and controls

SRO 4.4 Importance

Technical EOP-0, OP-CA-103-102-1001, 1C05 ALM, References LS-AA-1110 Reportable Event SAF 1.6 (RPS

Actuation), EP-AA-114 References None provided

Learning Recall the Operator Actions for Verifying a Objective Reactor Trip per EOP-0 Reactivity Control

Question New Source




Comments None



'.I Task 1201.009 Verify the RC Safety Function is satisfied I


13


ID: Q1~6574 Points: 1.00


• Plant startup in progress per OP-1 • RCS temperature is 220°F and slowly rising • 11 and 13 Containment Air Coolers (CACs) are the only running CACs and both are in High

Speed • Containment temperature is 65°F and slowly rising


• Component Engineering reports that vibration readings on 11 CAC are significantly higher than normal

• 11 CAC is declared inoperable • 11 CAC is secured per Ol-5A

Which ONE of the following is 1) the system impact on the plant startup and 2) what actions are required?

A. 1) Startup could not continue. 2) Cooldown to <200°F per OP-5.

B. 1) Startup could not continue. 2) Maintain current RCS conditions per OP-1.

C. 1) Startup could continue per OP-1. 2) Invoke TS 3.0.4 prior to exceeding 300°F.

D. 1) Startup could continue per OP-1. 2) RCS pressure would be limited to <1750 PSIA.

Answer: C


OPERATIONS


Answer Explanation:

A. Incorrect 1) False. Having to secure the startup is plausible if the Operator believes securing the CAC per Ol-5A and declaring the CAC inoperable isolates SRW to the CAC and affects the SRW system (which is required in Modes 1-4). TS 3.6.6 for the CACs only applies in Modes 1-3. 2) False. Shutdown to Mode 5 is plausible if the Operator believes Mode 4 is no longer allowed. However, TS 3.6.6 for the CACs only applies in Modes 1-3.

B. Incorrect 1) False. Having to secure the startup is plausible if the Operator believes securing the CAC per Ol-5A and declaring the CAC inoperable isolates SRW to the CAC and affects the SRW system (which is required in Modes 1-4). TS 3.6.6 for the CACs only applies in Modes 1-3. 2) True. Maintaining current conditions is always a choice during a plant startup provided TS or other plant conditions are not forcing a shutdown.

C. Correct 1) True. TS 3.6.6 allows the use of TS 3.0.4, allowing startup with a piece of inoperable equipment. 2) True. TS 3.0.4 would allow a Mode change provided requirements of 3.0.4.b were completed and TS 3.6.6 was entered once Mode 3 was reached.

D. Incorrect 1) True. TS 3.6.6 allows the use of TS 3.0.4, allowing startup with a piece of inoperable equipment. 2) False. Limiting RCS pressure to <1750 PSIA is plausible since TS 3.6.6, which applies to both the CACs and Containment Spray, limits Containment Spray to <1750 PSIA if Containment Spray is inoperable. However, there is no pressure limit associated with the CACs in TS 3.6.6.

Page 40 of 80 13 April 2016




Topic: SRO ONLY CAC Failure During Plant Startup RO Importance: 2.3 SRO Importance: 2.6 Cognitive level (High or


Tier/Group 2/1

KIA Info 022 Containment Cooling System (CCS) A2 Ability to (a) predict the impacts of the following malfunctions or operations on the CCS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: A2.02 Fan motor vibration

SRO 2.6 Importance

Technical Ol-5A, 1C19 ALM, TS 3.6.6, TS 3.0.4 References


Learning Determine when to declare Mode 3 from Objective Mode 4 and what requirements initially need

to be met

Question New Source




.14



Comments None

Task 203.006 Complete OP-6 Mode 3 Checklist

ID:Q106577 Points: 1.00

Given Provided Reference


• Power is at 90% • RCS Tcold reads 544.5°F • The Control Room Operator is preparing to raise load on the Turbine

Which ONE of the following 1) directions, if any, should be given to the Reactor Operator and 2) what is the expected plant response when Turbine load is raised?

A. 1) No directions to the RO are necessary. 2) Main Steam Header pressure as read on 2-PIC-4056 will lower and Turbine First Stage Pressure as read on 2C02 will rise.

B. 1) Raise Reactor Power. 2) Main Steam Header pressure as read on 2-PIC-4056 will lower and Turbine First Stage Pressure as read on 2C02 will rise.

C. 1) No directions to the RO are necessary. 2) Main Steam Header pressure as read on 2-PIC-4056 will rise and Turbine First Stage Pressure as read on 2C02 will lower.

D. 1) Raise Reactor Power. 2) Main Steam Header pressure as read on 2-PIC-4056 will rise and Turbine First Stage Pressure as read on 2C02 will lower.

Answer: B


OPERATIONS


Answer Explanation:

A Incorrect 1) Using OP-3 Figure 1, given conditions result in Tcold being 1.9F off program initially. When Turbine Load is raised, this will result in temperature lowering even further. Per OP-3 Appendix C, power should be raised using CEAs or boration to restore Tcold to within 2F of program. No direction to the RO is plausible if the Operator believes raising steam demand will raise Reactor Power and will restore Tcold to within 2F of program. 2) As steam demand increases, steam flow will rise and pressure in the Main Steam system will lower. Main Steam Header pressure on 2-PIC-4056 will lower from 900 PSIG (at 0% power) to 800 PSIG (at 100%) as steam flow increases. Turbine First Stage Pressure is directly proportional to Turbine Load. As load is raised, First Stage Pressure will rise.

B. Correct 1) Using OP-3 Figure 1, given conditions result in Tcold being 1.9F off program initially. When Turbine Load is raised, this will result in temperature lowering even further. Per OP-3 Appendix C, power should be raised using CEAs or boration to restore Tcold to within 2F of program. 2) As steam demand increases, steam flow will rise and pressure in the Main Steam system will lower. Main Steam Header pressure on 2-PIC-4056 will lower from 900 PSIG (at 0% power) to 800 PSIG (at 100%) as steam flow increases. Turbine First Stage Pressure is directly proportional to Turbine Load. As load is raised, First Stage Pressure will rise.

C. Incorrect 1) Using OP-3 Figure 1, given conditions result in Tcold being 1.9F off program initially. When Turbine Load is raised, this will result in temperature lowering even further. Per OP-3 Appendix C, power should be raised using CEAs or boration to restore Tcold to within 2F of program. No direction to the RO is plausible if the Operator believes raising steam demand will raise Reactor Power and will restore Tcold to within 2F of program. 2) As steam demand increases, steam flow will rise and pressure in the Main Steam system will lower. Main Steam Header pressure on 2-PIC-4056 will lower from 900 PSIG (at 0% power) to 800 PSIG (at 100%) as steam flow increases. Turbine First Stage Pressure is directly proportional to Turbine Load. As load is raised, First Stage Pressure will rise. Main Steam Header pressuring rising is plausible if the Operator believes Main Steam Header Pressure is directly proportional to steam flow. Turbine First Stage Pressure lowering is plausible if the Operator believes a higher steam flow will result in a lower First Stage Pressure.

D. Incorrect 1) Using OP-3 Figure 1, given conditions result in Tcold being 1.9F off program initially. When Turbine Load is raised, this will result in temperature lowering even further. Per OP-3 Appendix C, power should be raised using CEAs or boration to restore Tcold to within 2F of program. 2) As steam demand increases, steam flow will rise and pressure in the Main Steam system will lower. Main Steam Header pressure on 2-PIC-4056 will lower from 900 PSIG (at 0% power) to 800 PSIG (at 100%) as steam flow increases. Turbine First Stage Pressure is directly proportional to Turbine Load. As load is raised, First Stage Pressure will rise. Main Steam Header pressuring rising is plausible if the Operator believes Main Steam Header Pressure is directly proportional to steam flow. Turbine First Stage Pressure lowering is plausible if the Operator believes a higher steam flow will result in a lower First Stage Pressure.

Page: 43 of 80 13 April 2016




Topic: SRO ONLY Turbine Load Increase at 90% with Tcold Off Program



Tier/Group 2/1

K/A Info 039 Main and Reheat Steam System (MRSS) A2 Ability to (a) predict the impacts of the following malfunctions or operations on the MRSS; and (b) based on predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: A2.05 Increasing steam demand, its relationship to increases in reactor power

SRO 3.6 Importance

Technical OP-3-2 (Normal Power Operation) References Ol-43A-2 (Main Turbine and

Generator/Exciter Operation)

References OP-3-2 Figure 1 (Temperature Program provided Curve)

Learning Given plant conditions and a power Objective maneuver in progress evaluate if the

continuously applicable steps of Appendix C or D are being met

Question New Source






Comments None

Task 203.002 Direct increasing/decreasing load on Main Turbine Generator and implement required actions for various power levels


15


ID: 0106581


• Power is at 100% • 11 AFW Pump Testing per STP 0-5A 11-1, Auxiliary Feedwater System Quarterly

Surveillance Test, is on the schedule

Which ONE of the following are 1) the required actions to support 11 AFW Pump testing and 2) what actions, if any, are necessary if a valid AFAS were to occur during the STP?

A. 1) Lower power to 95%, then commence 11 AFW Pump testing. 2) An Equipment Operator will need to close the AFW Pump Room Watertight Doors.

B. 1) Lower power to 95%, then commence 11 AFW Pump testing. 2) An Equipment Operator will need to open 11 AFW Pump discharge valve.

C. 1) Maintain -100% power, station a dedicated operator in the AFW Pump Room with direct communications with a dedicated operator in the Control Room, then commence 11 AFW Pump testing. 2) No action required as 12 AFW Pump will be aligned for Auto.

D. 1) Maintain -100% power, station a dedicated operator in the AFW Pump Room with direct communications with a dedicated operator in the Control Room, then commence 11 AFW Pump testing. 2) Dedicated operator will need to open 11 AFW Pump discharge valve.

Answer: D


OPERATIONS


Answer Explanation:

A. lncorrect-1) Lowering power to 95% is plausible if the Operator believes AFW flow is established to the S/Gs. STP 0-73H is the only STP when the AFW pumps will be initiating flow to the S/Gs. A load drop to 95% is required for STP 0-73H. During STP 0-5A-1, the AFW pumps are run but are not discharging to the S/Gs as the pump discharge valve is shut during the testing. Reactor power is monitored while the steam driven AFW pumps are running, but a large power decrease is not required since cold AFW will not be feeding the SGs. Per TS 3.7.3 note, a dedicated operator shall be stationed with the AFW pump taken OOS administratively. The dedicated operator shall be in direct communication with the dedicated operator in the Control Room. 2) Shutting the watertight doors is plausible if the Operator believes the doors need to be shut whenever the pumps are running during an AFAS. Per STP 0-5A-1, the doors only need to be shut if Turbine Building ventilation is lost while the AFW pumps are running.

B. lncorrect-1) Lowering power to 95% is plausible if the Operator believes AFW flow is established to the S/Gs. STP 0-73H is the only STP when the AFW pumps will be initiating flow to the S/Gs. A load drop to 95% is required for STP 0-73H. During STP 0-5A-1, the AFW pumps are run but are not discharging to the S/Gs as the pump discharge valve is shut during the testing. Reactor power is monitored while the steam driven AFW pumps are running, but a large power decrease is not required since cold AFW will not be feeding the SGs. Per TS 3.7.3 note, a dedicated operator shall be stationed with the AFW pump taken OOS administratively. The dedicated operator shall be in direct communication with the dedicated operator in the Control Room. 2) The TBO opening the pump discharge valve is plausible if the Operator needs to get a valve operated in the Turbine Building. Per TS 3.7.3, the dedicated watch would be stationed and would be the operator performing this evolution.

C. lncorrect-1) During STP 0-5A-1, the AFW pumps are run but are not discharging to the S/Gs as the pump discharge valve is shut during the testing. Reactor power is monitored while the steam driven AFW pumps are running, but a large power decrease is not required since cold AFW will not be feeding the SGs. Per TS 3.7.3 note, a dedicated operator shall be stationed with the AFW pump taken OOS administratively. The dedicated operator shall be in direct communication with the dedicated operator in the Control Room. 2) No action is plausible if the Operator believes 12 AFW is aligned for Auto while 11 testing is conducted. MS is aligned to only one steam driven AFW pump at a time. With 11 running, 12 would be in a standby status and would not automaticity start on an AFAS.

D. Correct-1) During STP 0-5A-1, the AFW pumps are run but are not discharging to the S/Gs as the pump discharge valve is shut during the testing. Reactor power is monitored while the steam driven AFW pumps are running, but a large power decrease is not required since cold AFW will not be feeding the SGs. Per TS 3.7.3 note, a dedicated operator shall be stationed with the AFW pump taken OOS administratively. The dedicated operator shall be in direct communication with the dedicated operator in the Control Room. 2) Per TS 3.7.3, the dedicated watch would be stationed and would be responsible for restoring AFW from the test configuration to its operational configuration when required.

Page: 47 of 80 13 April 2016




Topic: SRO ONLY AFW LCO Requirements/Actions During Testin~



Tier/Group 2/1

KIA Info 061 Auxiliary I Emergency Feedwater (AFW) System 2.2 Equipment Control 2.2.38 Knowledge of conditions and limitations in the facility license.

SRO 4.5 Importance

Technical STP 0-5A11-1 (Auxiliary Feedwater System References Quarterly Surveillance Test)

STP 0-73H-1 (AFW Pump Large Flow Test), TS 3.7.3


Learning Given a plant or system condition and the Objective Technical Specifications, be able to apply the

appropriate action requirements

Question New Source





EXAMINATION ANSWER KEY LOI 2015 NRC SRO Exam {June 2016)

Comments None

Task 204.094 Determine and apply Tech Spec requirements


16


ID: Q106582 Points: 1.00


• Multiple alarms are received 1 COS • Indications at 1C15 Reactor Protection System Channel A are as pictured below

Which ONE of the following is 1) the probable cause of the 1 COS alarms and 2) what direction should be provided to the Control Room Operator?

A. 1) Delta-T Power has failed low. 2) Within an hour, bypass RPS Channel A Trip Units 1,2, 7,8, 10 per 01-6.

B. 1) A Linear Range Nuclear Instrument has failed low. 2) Within an hour, bypass RPS Channel A Trip Units 1,2,7,8, 10 per 01-6.

C. 1) Delta-T Power has failed low. 2) Within an hour, bypass RPS Channel A Trip Units 7 and 10 only per 01-6.

D. 1) A Linear Range Nuclear Instrument has failed low. 2) Within an hour, bypass RPS Channel A Trip Units 7 and 10 only per 01-6.

Answer: B


OPERATIONS


Answer Explanation:

A. lncorrect-1) Delta-T failing low is plausible if the Operator evaluates the NI-Delta T meter on the RPSCIP drawer incorrectly. Delta-T power failing low would actually cause the NI-Delta T meter to peg to the right. Based on NI drawer indications, the LRNI Lower Instrument has failed low. 2) Per N0-1-200, RPS Trip Units 1,2,7,8,10 are bypassed for LRNI failure.

B. Correct-1) Based on NI drawer indications, the LRNI Lower Instrument has failed low. 2) Per N0-1-200, RPS Trip Units 1,2,7,8,10 are bypassed for LRNI failure.

C. lncorrect-1) Delta-T failing low is plausible if the Operator evaluates the NI-Delta T meter on the RPSCIP drawer incorrectly. Delta-T power failing low would actually cause the NI-Delta T meter to peg to the right. Based on NI drawer indications, the LRNI Lower Instrument has failed low. 2) Bypassing only Trip Units 7, 10 is plausible if the Operator only bypasses those Trip Units that are actively tripped and does not consider inoperability impacts to the other Trip Units. Per N0-1-200, RPS Trip Units 1,2,7,8, 10 are bypassed for LRNI failure.

D.lncorrect-1) Based on NI drawer indications, the LRNI Lower Instrument has failed low. 2) Bypassing only Trip Units 7, 10 is plausible if the Operator only bypasses those Trip Units that are actively tripped and does not consider inoperability impacts to the other Trip Units. Per N0-1-200, RPS Trip Units 1,2,7,8,10 are bypassed for LRNI failure.

Page: 51 of 80 13 April 2016




Topic: SRO ONLY Loss of LRNI and TS Requirements RO Importance: 3.9 SRO Importance: 4.6 Cognitive level (High or


Tier/Group 212

KIA Info 015 Nuclear Instrumentation System 2.2 Equipment Control 2.2.42 Ability to recognize system parameters that are entry-level conditions for Technical Specifications

SRO 4.6 Importance

Technical TS 3.3.1, N0-1-200 (Control of Shift References Activities)


Learning Given a plant or system condition and the Objective Technical Specifications, be able to apply the

appropriate action requirements

Question New Source




Comments Considered SRO since SRO will determine Trip Units that need to be bypassed


17


'Task 1204.094 Determine and apply Tech Spec requirements I

Roints: 1.00

A Waste Gas discharge is in progress with the following conditions:

• 11 WGDT is being released to Unit-2 Plant Vent at 45 SCFM • Waste Gas Release Permit Flow limit is 50 SCFM

Air is lost to WG DISCH B/U ISOL, O-CV-2192.

Which ONE of the following is 1) the expected impact on the Waste Gas discharge and 2) what is the first action required in 01-178?

A 1) Waste Gas discharge flowrate will go to zero. 2) Shut the WG DISCH HOR PSR/FLOW CONTR, O-WGS-2191-PCV.

B. 1) Waste Gas discharge flowrate will rise. 2) Shut the WG DISCH HOR PSR/FLOW CONTR, O-WGS-2191-PCV.

C. 1) Waste Gas discharge flowrate will go to zero. 2) Shut the WG DISCH ISOL, O-WGS-2191-CV.

D. 1) Waste Gas discharge flowrate will rise. 2) Shut the WG DISCH ISOL, O-WGS-2191-CV.

Answer: A


OPERATIONS


Answer Explanation:

A. Correct 1) True. On a loss of air to WGS-2192-CV, the valve will fail shut. This will secure the Waste Gas discharge and flow will go to zero. 2) True. Per 01-178, if one of the WG DISCH ISOL valves, either WGS-2191-CV or WGS-2192-CV, goes shut during a release, then WG DISCH HDR PSR/FLOW CONTR, O-WGS-2191-PCV is shut. Per the Caution note in 01-178, this is done to prevent the discharge header RV from lifting.

8. Incorrect 1) False. On a loss of air to WGS-2192-CV, the valve will fail shut. This will secure the Waste Gas discharge and flow will go to zero. A rising Waste Gas discharge flowrate is plausible to the Operator if they determine a loss of air to WGS-2192-CV will fail the valve open with the valve position being throttled to control discharge flowrate. 2) True. Per 01-178, if one of the WG DISCH ISOL valves, either WGS-2191-CV or WGS-2192-CV, goes shut during a release, then WG DISCH HDR PSR/FLOW CONTR, O-WGS-2191-PCV is shut. Per the Caution note in 01-178, this is done to prevent the discharge header RV from lifting.

C. Incorrect 1) True. On a loss of air to WGS-2192-CV, the valve will fail shut. This will secure the Waste Gas discharge and flow will go to zero. 2) False. Shutting O-WGS-2191-CV is plausible since this in an action taken in 01-178 to secure from the Waste Gas discharge. However, this is not the first action taken. Per 01-178, if one of the WG DISCH ISOL valves, either WGS-2191-CV or WGS-2192-CV, goes shut during a release, then WG DISCH HDR PSR/FLOW CONTR, O-WGS-2191-PCV is shut first. Per the Caution note in 01-178, this is done to prevent the discharge header RV from lifting.

D. Incorrect 1) False. On a loss of air to WGS-2192-CV, the valve will fail shut. This will secure the Waste Gas discharge and flow will go to zero. A rising Waste Gas discharge flowrate is plausible to the Operator if they determine a loss of air to WGS-2192-CV will fail the valve open with the valve position being throttled to control discharge flowrate. 2) False. Shutting O-WGS-2191-CV is plausible since this in an action taken in 01-178 to secure from the Waste Gas discharge. However, this is not the first action taken. Per 01-178, if one of the WG DISCH ISOL valves, either WGS-2191-CV or WGS-2192-CV, goes shut during a release, then WG DISCH HDR PSR/FLOW CONTR, O-WGS-2191-PCV is shut first. Per the Caution note in 01-178, this is done to prevent the discharge header RV from lifting.

Page: 54 of 80 13 April 2016




Topic: SRO ONLY Waste Gas Discharge Secures Unexpectedly Actions



Tier/Group 2/2

KIA Info 071 Waste Gas Disposal System (WGDS) A2 Ability to (a) predict the impacts of the following malfunctions or operations on the Waste Gas Disposal System ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: A2.06 Supply failure to the isolation valve

SRO 2.5 Importance

Technical 01-178 (Waste Gas System), 60735SH0001 References


Learning Determine the appropriate actions per Objective technical specifications and procedures for

Waste Gas, Reactor Coolant Waste and/or Miscellaneous Waste Processing System parameters

Question New Source



18




Comments None

Task 069.001 Release Waste Gas

Po(llt$: 1.00


• Large grass influx occurs • 13 Circulating Water Pump (CWP) is secured due to high Travelling Screen D/P • Condenser vacuum begins lowering

Which ONE of the following is 1) when a manual Reactor Trip must be ordered per the appropriate AOP and 2) what actions are required in EOP-0 if the Condenser vacuum continues to lower?

A. 1) 23.5 inches-Hg. 2) Shift heat removal method to the ADVs if the Condenser vacuum lowers below 22.5 inches-Hg.

B. 1) 25 inches-Hg. 2) Shift heat removal method to AFW if the Condenser vacuum lowers below 22.5 inches-Hg.

C. 1) If 12 CWP is also secured due to high Travelling Screen DIP. 2) Shift heat removal method to the ADVs if the Condenser vacuum lowers below 22.5 inches-Hg.

D. 1) If 14 CWP is also secured due to high Travelling Screen DIP. 2) Shift heat removal method to AFW if the Condenser vacuum lowers below 22.5 inches-Hg.

Answer: A


OPERATIONS


Answer Explanation:

A Correct 1) True. Given the conditions, AOP-7G and AOP-7L are implemented. Both procedures require manual Reactor Trip if condenser vacuum lowers to 23.5 inches-Hg. 2) True. Per AOP-7G, the TBVs are locked out if condenser vacuum lowers to 22.5 inches-Hg. Heat removal would be shifted to the ADVs.

B. Incorrect 1) False. Given the conditions, AOP-7G and AOP-7L are implemented. 25 inches-Hg is plausible to the Operator since a down power is required if vacuum lowers below 25 inches-Hg. Both procedures require manual Reactor Trip if condenser vacuum lowers to 23.5 inches-Hg. 2) False. Shifting heat removal to AFW when vacuum lowers below 22.5" is plausible since 22.5" is the setpoint where the TBVs are locked out. The Steam Generator Feed Pumps (SGFPs) trip at a vacuum of 20 inches-Hg and would continue to operate with a vacuum of 22.5". Isolating the TBVs would not shutdown the SGFPs.

C. Incorrect 1) False. Given the conditions, AOP-7G and AOP-7L are implemented. 12 CWP being secured is plausible to the Operator since two CWPs being secured is trip criteria in AOP-7L, but the two secured CWPs must be associated with the same waterbox. Since 13 CWP was the first CWP to be secured, only a downpower is required if 12 CWP is secured. 2) True. Per AOP-7G, the TBVs are locked out if condenser vacuum lowers to 22.5 inches-Hg. Heat removal would be shifted to the ADVs.

D. Incorrect 1) True. Given the conditions, AOP-7G and AOP-7L are implemented. 14 CWP being secured would require a manual reactor trip since AOP-7L directs a trip when two secured CWPs are associated with the same waterbox. 2) False. Shifting heat removal to AFW when vacuum lowers below 22.5" is plausible since 22.5" is the setpoint where the TBVs are locked out. The Steam Generator Feed Pumps (SGFPs) trip at a vacuum of 20 inches-Hg and would continue to operate with a vacuum of 22.5". Isolating the TBVs would not shutdown the SGFPs.

Page: 57 of 80 13 April 2016




Topic: SRO ONLY AOP-7L and Condenser Vacuum Setpoints to Trip/TBVs



Tier/Group 212

KIA Info 075 Circulating Water System A2 Ability to (a) predict the impacts of the following malfunctions or operations on the circulating water system; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: A2.03 Safety features and relationship between condenser vacuum, turbine trip, and steam dump

SRO 2.7 Importance

Technical AOP-7G, AOP-7L, EOP-1 References


Learning Identify three (3) major plant components Objective that are affected by a loss of condenser

vacuum, and how they impact plant operation

Question New Source






Comments None

Task 202.087 Respond to condenser vacuum reduction

Points: 1.00

U-1 is operating at 100% power when the following transient occurs:

• Reactor trips • All CEAs are fully inserted. • All electrical busses are energized from their normal power supplies. • Pressurizer level is 88 inches and lowering slowly • Pressurizer pressure is 1875 PSIA and lowering slowly • T AVG is 530°F and lowering slowly • ADVs and TBVs are shut • RCS subcooling is 100°F and rising slowly • Main Feedwater is being supplied to both steam generators • 11 S/G level is -90 inches and lowering slowly • 12 S/G level is -50 inches and rising • Containment pressure is 1.5 PSIG and rising slowly • 11 Main Steam Line Radiation Monitor reads 3.6 E+2 uci/cc • 12 Main Steam Line Radiation Monitor reads 1.2 E+3 uci/cc

Which ONE of the following EOPs must be implemented?

A. EOP-4, Excess Steam Demand Event

B. EOP-5, Loss of Coolant Accident

C. EOP-6, Steam Generator Tube Rupture

D. EOP-8, Functional Recovery Procedure

Answer: D


OPERATIONS


Answer Explanation:

A. Incorrect-Plant conditions indicate a steam leak or an RCS leak is removing decay heat. Containment parameters indicate the location of this leak. A SG tube leak also exists on 12 SG based on MSLRM indications. Based on a tube leak occurring with either an RCS leak or a steam leak in containment, EOP-8 would be implemented. EOP-4 is plausible to the Operator since temperature is lowering and PZR level is lowering.

B. Incorrect-Plant conditions indicate a steam leak or an RCS leak is removing decay heat. Containment parameters indicate the location of this leak. A SG tube leak also exists on 12 SG based on MSLRM indications. Based on a tube leak occurring with either an RCS leak or a steam leak in containment, EOP-8 would be implemented. EOP-5 is plausible to the Operator due to Containment indications.

C. Incorrect-Plant conditions indicate a steam leak or an RCS leak is removing decay heat. Containment parameters indicate the location of this leak. A SG tube leak also exists on 12 SG based on MSLRM indications. Based on a tube leak occurring with either an RCS leak or a steam leak in containment, EOP-8 would be implemented. EOP-6 is plausible to the Operator due to MSLRM and SIG level indications.

D. Correct-Plant conditions indicate a steam leak or an RCS leak is removing decay heat. Containment parameters indicate the location of this leak. A SG tube leak also exists on 12 SG based on MSLRM indications. Based on a tube leak occurring with either an RCS leak or a steam leak in containment, EOP-8 would be implemented.

Page: 60 of 80 13 April 2016




Topic: SRO ONLY Select appropriate optimal/functional recovery procedure



Tier/Group 3/NA

KIA Info EPE 007 - Reactor Trip EA2 Ability to determine or interpret the following as they apply to a reactor trip: EA2.06 Occurrence of a reactor trip

SRO 4.4 Importance

Technical EOP-0 (Post Trip Immediate Actions) References


Learning Given plant conditions and the Diagnostic Objective Flowchart, determine the correct procedure

to implement following safety function assessment in EOP-0







Comments None

Task 201.002 Perform Diagnostics Actions of EOP-0


20


ID: L 103279 i< Points: 1.00

Unit-1 is in MODE 6 with the following conditions:

• Movement of irradiated fuel is in progress. • A safety tagger requests the Fuel Handling Supervisor (FHS) to independently verify a safety

tagged component located on the 69 Ft. containment elevation.

Which ONE of the following describes the restrictions, if any, placed on the FHS performing the Independent Verification (IV)?

A The FHS may not perform the IV under any circumstances.

B. The FHS may perform the IV at any time during the movement of irradiated fuel.

C. The FHS may perform the IV only during suspension of movement of irradiated fuel.

D. The FHS may perform the IV only if the Refueling Machine Operator is a licensed operator.

Answer: C

Answer Explanation:

The FHS may leave the refueling machine if a licensed operator is the RFM operator. The FHS may tour the 69 Foot area of containment. The FHS supervises refueling operations with no other concurrent duties.

A Incorrect-The designated FHS may perform the IV if fuel handling activities have been suspended. Plausible to the Operator since FHS is to have no other duties while moving fuel.

B. lncorrect-N0-1-200, Step 5.1.B.a. states that during fuel loading, transferring or handling, an SRO shall be designated the FHS and further states: The FHS shall directly supervise from the Containment 69 Foot elevation and have no other concurrent duties. Therefore, the IV cannot be performed by the FHS unless fuel handling activities are suspended. Plausible to the Operator since performing a tagout IV can be an Operator task.

C. Correct-N0-1-200, Step 5.1.B.a. states that during fuel loading, transferring or handling an SRO shall be designated the FHS and further states: The FHS shall directly supervise from the Containment 69 Foot elevation and have no other concurrent duties. Therefore, the IV cannot be performed by the FHS unless fuel handling activities are suspended.

D. Incorrect-Having a licensed operator on the RFM allows the FHS to leave the RFM as long as FHS remains on the 69 Foot elevation of the Containment and focused solely on the fuel handling activities. The FHS cannot perform the IV unless fuel handling activities are suspended. Plausible to the Operator since performing a tagout IV can be an Operator task.




System ID: 103279 User-Defined ID: L103279 Cross Reference Number: LOI070A REFUELING EQUIP OPS

Topic: SRO ONLY Determine the limits of travel for the FHS RO Importance: 2.8 SRO Importance: 3.9 Cognitive level (High or

Low Low):




KIA Info 2.1 Conduct of Operations 2.1.40 Knowledge of refueling administrative requirements

SRO 4.3 Importance

Technical N0-1-200 (Control of Shift Activities) References


Learning Determine limits of travel for the FHS when Objective the RFM is being operated





Comments None

Task 204.090 Apply watchstander's role as part of the Control Room Team in Normal, Abnormal, Emerqency and ERPIP situations


21


ID: Q106617 Points: 1.00

A complex test procedure has the following risk assessment attributes:

• Has been classified as a Special Test or Evolution (STOE) • Has a Special Test Coordinator assigned

Which ONE of the following is the correct responsibility during the STOE process?

A. The Special Test Coordinator approves implementation of the test.

B. The Shift Manager determines the Cross-Discipline Review requiements for technical adequcy of the document.

C. The Shift Manager ensures plant conditions are maintained during performance of the test.

D. The Special Test Coordinator provides the Just-In-Time-Training prior to performance of the test.

Answer: C

Answer Explanation:

A. Incorrect-The STC approving implementation is plausible to the Operator since this is a responsibility listed in OP-AA-108-110. Per the procedure, Operations Shift Management approves implementation of the STOE.

B. Incorrect-The SM assuming overall responsibility of the conduct of the test is plausible to the Operator since this is a responsibility listed in OP-AA-108-110. Per the procedure, the Station Qualified Reviewer is assigned this responsibility.

C. Correct-Per OP-AA-108-110, Operations Shift Management ensures required plant conditions are maintained during performance of the STOE.

D. Incorrect-The STC providing the training is plausible to the Operator since this is a responsibility listed in OP-AA-108-110. Per the procedure, Training would be tasked with providing the training.





Topic: SRO ONLY Process for Conducting Special or Infrequent Tests


Low Low): NRC K/A Info, References:

Tier/Group 3/NA

K/A Info 2.2 Equipment Control 2.2.7 Knowledge of the process for conducting special or infrequent tests.

SRO 3.6 Importance

Technical OP-AA-108-110 (Evaluation of Special Tests References or Evolutions)


Learning Apply the Requirements of CNG-OP-1.01-Objective 1000, Conduct of Operations

Question New Source



10CFR 55.43(b )(3) Part 55 Content

Comments None



Task 204.091 Maintain situational awareness of the plant during Normal, Abnormal, Emergency and ERPIP operations


22


ID: L92932 Points: 1.00

Unit -1 was operating at 100% power when the following transient occurred:

• 1-RE-1752NBIC/D (11112113114 CAR Suction RAD MONs) alarmed and indicate a leakrate of 28 GPO and stable

• 1-Rl-4014 (Unit 1 SIG BID RMS) is elevated • 1-RIC-4095 (Unit 1 SIG BID Recovery RMS) is elevated • Chemistry sample results indicate a leakrate of 31 GPO and stable

Which ONE of the following actions are required?

A. Implement AOP-2A, Excessive RCS Leakage, and reduce power to <50% within an hour.

B. Implement 01-SA, Slowdown System, and secure SIG Slowdown.

C. Implement AOP-10, Abnormal Secondary Chemistry Conditions, and place Unit-1 Turbine Building Sump Pumps in STOP.

D. Implement AOP-2A, Excessive RCS Leakage, and trip the reactor.

Answer: C

Answer Explanation:

A. Incorrect-Implementing AOP-2A is plausible to the Operator since the AOP addresses a SIG tube leak. However, an RCS leak of 31 GPO (.02 GPM) is below the threshold for implementation of AOP-2A. AOP-10 specifies implementation of AOP-2A if the SG tube leakage reaches the operational limit of 50 GPO through any one SG.

B. Incorrect-Securing blowdown is plausible to the Operator since AOP-2A secures blowdown during a SIG tube leak. However, the SIG Slowdown System RMSs, while elevated, have yet to reach a value where manual or automatic action is required per plant procedure. The decision to secure Slowdown under these circumstances would be based on recommendations from Chemistry.

C. Correct-AOP-10, Attachment 1 (UNIT 1 ACTIONS FOR SG TUBE LEAKAGE GREATER THAN 5 GPO) is written to address SG tube leakage of between 5 GPO and 50 GPO. Attachment 1 requires the Unit-1 Turbine Building Sump pumps be placed in STOP.

D. Incorrect-Tripping the reactor is plausible to the Operator since these are actions in AOP-2A for a SG tube leak if the Operator mistakes 31 GPO for 31 GPM , which exceeds the capacity of a Charging Pump. However, an RCS leak of 31 GPO (.02 GPM) is below the threshold for implementation of AOP-2A.




System ID: 92932 User-Defined ID: L92932 Cross Reference Number: 2.3.5

Topic: SRO ONLY Given conditions for 31 gpd SGTL, identify AOP-10 as correct procedure to enter

RO Importance: SRO Importance: 2.9 Cognitive level (High or

High Low): NRC KIA Info, References:

Tier/Group 3/NA

KIA Info 2.3 Radiation Control 2.3.5 Ability to use radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc.

SRO 2.9 Importance

Technical 1C22 ALM (RMS Alarm Manual), AOP-10 References (Abnormal Secondary Chemistry Conditions),

AOP-2A (Excessive RCS Leakage)


Learning Identify the entry conditions for AOP-10, Objective Abnormal Secondary Chemistry Conditions







Comments None

Task 202.127 Respond to Unit 1 (2) Steam Generator tube leakage > 5 GPO but below the T.S. limit of 100 GPO



Licensed Operator Requal Training (LOR)

Operations Procedure References (from Nucleis)

AOP AOP-10 ABNORMAL SECONDARY CHEMISTRY CONDITIONS

System Designations

Abnormal Operating Procedures (AOPs)


23


ID: 0106634 · Points: 1.00

Which ONE of the following describes a responsibility of the Unit Supervisor when signing a Gaseous Release Permit?

A Enter termination criteria into the Plant Computer for high flow rate and high activity as indicated on the Release Permit.

B. Ensure required plant systems are in operation and the required plant configuration has been established.

C. Ensure radiation monitor alarm setpoints have been adjusted to the appropriate values.

D. Ensure pre-release source check and channel check has been completed.

Answer: B

Answer Explanation:

A Incorrect-Entering termination criteria into the computer is plausible since the SM/SRO would perform an IV that the termination criteria has properly been entered into the Plant Computer. However, entry of the data into the Plant Computer is the responsibility of the CRO.

B. Correct-Per CP-604, Attachment 4 or CP-645 Attachment 11, the SM/SRO signature verifies the release criteria are understood, that required plant systems are in operation and that the required plant configuration has been established.

C. Incorrect-Adjusting RMS setpoints is plausible to the Operator since the RMS Warning/Alert/Critical alarm values from the permit are inputted to the Plant Computer for termination criteria. However, RMS alarm setpoints are not adjusted during a discharge.

D. Incorrect-Pre-release source check and channel checks is plausible to the Operator since these are actions required and initialed for on the Waste Gas Permit. However, these checks are the responsibility of the CRO. Per CP-604, Attachment 4 or CP-645 Attachment 11, the SM/SRO signature verifies the release criteria are understood, that required plant systems are in operation and that the required plant configuration has been established.





Topic: SRO ONLY Approve a Gaseous Release Permit. RO Importance: 2.0 SRO Importance: 3.8 Cognitive level (High or


Tier/Group 3/NA

K/A Info 2.3 Radiation Control 2.3.6 Ability to approve release permits

SRO 3.8 Importance

Technical CP-604 (Radioactive Gaseous Waste References Permits)

CP-645 (Manual Liquid and Gaseous Radioactive Waste Release Permits)


Learning Determine when a Waste Gas Decay Tank Objective can be isolated and/or released





Comments Modified from Q74971. Q74971 used on NRC 2008 SRO Exam.

Task 069.001 Release Waste Gas

OPERATIONS Page: 73 of 80 13Apri12016

24


ID: 0106636 ·· n?e>ints: 1.00


• A Steam Generator (S/G) Tube Leak developed • The Reactor was manually tripped per the AOP • The Optimal EOP was implemented • The CRO is beginning the Optimal EOP step to isolate the affected S/G • A large RCS leak into Containment begins • The ST A reports that the Containment Environment Safety Function is no longer being met

Which ONE of the following strategies should be used to mitigate the transient?

A. Immediately exit the Optimal EOP and implement EOP-8 to complete the isolation of the affected S/G and address the RCS leak.

B. Use the Optimal EOP to complete the isolation of the affected S/G in parallel with transitioning to EOP-8 to address the RCS leak.

C. Remain in the Optimal EOP to first complete the isolation of the affected S/G and address the RCS leak, then transition to EOP-8.

D. Remain in the Optimal EOP to first complete the isolation of the affected S/G, then transition to EOP-8 to address the RCS leak.

Answer: B


OPERATIONS


Answer Explanation:

A. Incorrect-Isolating an affected SIG is a critical step to limit radiation release. Per OPCA-103-102-1001 EOP-8 implementation flowchart, the CRS is allowed to direct continued execution of required Optimal EOP steps while EOP-8 is implemented. An inprogress SIG isolation should continue while the EOP-8 Resource Assessment Table is evaluated and actions taken in EOP-8. This will ensure a more timely isolation of the SIG. Immediately exiting the Optimal EOP is plausible to the Operator since these are actions taken if critical Optimal EOP steps are not in progress.

B. Correct-Isolating an affected SIG is a critical step to limit radiation release. Per OP-CA-103-102-1001 EOP-8 implementation flowchart, the CRS is allowed to direct continued execution of required Optimal EOP steps while EOP-8 is implemented. An in-progress SIG isolation should continue while the EOP-8 Resource Assessment Table is evaluated and actions taken in EOP-8. This will ensure a more timely isolation of the SIG.

C. Incorrect-Isolating an affected SIG is a critical step to limit radiation release. Per OPCA-103-102-1001 EOP-8 implementation flowchart, the CRS is allowed to direct continued execution of required Optimal EOP steps while EOP-8 is implemented. An inprogress SIG isolation should continue while the EOP-8 Resource Assessment Table is evaluated and actions taken in EOP-8. This will ensure a more timely isolation of the SIG. Remaining in the Optimal to address the RCS leak is plausible since the same action is being done to isolate the SIG. However, since actions are not already in progress to address the RCS leak and two events are in progress, EOP-8 must be implemented to address the RCS leak.

D. Incorrect-Isolating an affected SIG is a critical step to limit radiation release. Per OPCA-103-102-1001 EOP-8 implementation flowchart, the CRS is allowed to direct continued execution of required Optimal EOP steps while EOP-8 is implemented. An inprogress SIG isolation should continue while the EOP-8 Resource Assessment Table is evaluated and actions taken in EOP-8. Remaining in the Optimal EOP to isolate the leak before implementing EOP-8 is plausible since two implementing procedures are normally not controlling an evolution.

Page: 75 of 80 13 April 2016




Topic: SRO ONLY EOP Mitigation Strategies RO Importance: 3.7 SRO Importance: 4.7 Cognitive level (High or


Tier/Group 3/NA

K/A Info 2.4 Emergency Procedures/Plan 2.4.6 Knowledge of EOP mitigation strategies

SRO 4.7 Importance

Technical OP-CA-103-102-1001 (Strategies for References Successful Transient Mitigation)


Learning Given situation(s) outside of the Optimum Objective Recovery Procedures the trainee will be able

to identify, understand the basis and take appropriate actions per plant operating procedures to mitigate the event(s)

Question New Source




Comments None



NRC K/A Info, References: Tier/Group 3/NA

K/A Info 2.4 Emergency Procedures/Plan 2.4.40 Knowledge of SRO responsibilities in emergency plan implementation

SRO 4.5 Importance

Technical EP-AA-112-100-F050 (Shift Emergency References Director Checklist), EAL-HOT EALS

References EAL-HOT EALS provided

Learning Determine appropriate emergency response Objective actions per the ERPIP while maintaining an

overview of plant conditions





Comments Form numbers no longer the same as in 2014.

Task 204.107 Notify offsite agencies of Emergency Classification


Documents

EXAMINATION ANSWER KEY - NRC: Home Page · • 24 4KV bus trips on a ground fault • The plant trips on Loss of Load • RCS pressure peaks at 2410 PSIA During EOP-0, the following