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Draft Written Exam. (Question #99 Is Omitted Because It Is Sensitive and Non-Public) (Folder 2)
	ML12072A356
Person / Time
Site:	FitzPatrick
Issue date:	01/13/2012
From:	Peter Presby; Division of Reactor Safety I
To:	; Entergy Nuclear Generation Co
	Jackson D
Shared Package
ML113070677	List: IR 05000333/2012301; ML113330864; ML120610564; ML12072A344; ML12072A351; ML12072A353; ML12072A356; ML12116A225; ML12116A242; ML12116A252; ML12116A259; ML12116A275; ML12116A290;
References
	TAC U01843
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Question #1 The plant is operating at 100% reactor power when the 'A' Reactor Recirculation pump trips.

How will RPV water level initially respond and what is the reason for this response?

RPV water level will ...

A. lower due to the lack of coolant velocity to sweep voids into the steam separator.

B. rise due to the addition of water into the downcomer by increased steam voiding.

C. lower due to the runback of the 'B' Reactor Recirculation pump to the #1 Speed Limiter.

D. rise due to the continuing addition of feedwater at 100% rated feedwater flow.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO Tier # 1 Group # 1 KIA # 295001 AK 2.01 Importance Rating 3.6/3.7 Proposed Question:

The plant is operating at 100% reactor power when the 'A' Reactor Recirculation pump trips.

How will RPV water level initially respond and what is the reason for this response?

RPV water level will ...

A. lower due to the lack of coolant velocity to sweep voids into the steam separator.

B. rise due to the addition of water into the downcomer by increased steam voiding.

C. lower due to the runback of the 'B' Reactor Recirculation pump to the #1 Speed Limiter.

D. rise due to the continuing addition of feedwater at 100% rated feedwater flow.

Proposed Answer:

B. rise due to the displacement of water into the downcomer by increased steam voiding.

Explanation (Optional):

A. Incorrect: lower due to the lack of coolant velocity to sweep voids into the steam separator. Plausible if the candidate does not understand the reactor physics and thermodynamic effects of a partial loss of recirc flow on RPV level.

B. Correct: rise due to the displacement of water into the downcomer by increased steam voiding. When a RWR pump trips, less cooling water flow is available to remove heat from the nuclear fuel in the core, therefore more steam voiding occurs resulting in less flow into the core from the jet pumps and thus a rise in downcomer level.

C. Incorrect: lower due to the runback of the 'B' Reactor Recirculation pump to the #1 Speed Limiter. Plausible if the candidate does not understand the reactor physics and thermodynamic effects of a partial loss of recirc flow on RPV level as well as the effect of a trip of one RWR pump on the remaining "running" RWR pump. With both feed pumps running at 100% power, total feed flow will not be < 20% of rated thus a 30%

runback will not occur.

D. Incorrect: rise due to the continuing addition of feedwater at 100% rated feedwater flow. Plausible if the candidate does not understand the reactor physics and thermodynamic effects of a partial loss of recirc flow on RPV level.

Technical Reference(s): SDLP-02H, Rev. 15 Proposed references to be provided to applicants during examination: None Learning Objective: SDLP-02H; 1.09.e Question Source: Bank # 21815 Modified Bank #

New Question History: Last NRC Exam Perry 2001 NRC ILO Exam Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 55.41 (7)

Comments:

Parent Question Taken from INPO Exam Bank - Perry 2001 ILO Exam The plant is operating at 100% reactor power when one Reactor Recirculation pump trips.

All systems respond as designed to this event. How will RPV water level initially respond and what is the reason for this response?

RPV water level will. ..

A. decrease due to the lack of coolant velocity to sweep voids into the steam separator.

B. increase due to the displacement of water into the downcomer by increased steam voiding.

C. decrease due to the run back of feedwater pumps to minimum speed.

D. Increase due to the continuing addition of feedwater at 100% rated feedwater flow.

Answer:

B. increase due to the displacement of water into the downcomer by increased steam voiding.

Question #2 The plant is operating at 100% power when the following occurs:

Transformers T2 and T3 trip due to high winds and flying debris

An Unusual Event is declared

Annunciator 09-8-2-24 "4160V BUS 10500 DEGRADED VOLTAGE TIMER INITIATED" alarms

EPIC alarm A-721 "SWGR BUS 12500 VOLT LOW" is in alarm at 588 VAC

No Operator actions are taken.

10 seconds after the annunciator is received you note the following indications:

10500 Bus voltage as indicated at the 09-8 panel is currently 3850 volts and steady.

Drywell pressure is 1.8 psig and steady.

RPV water level is 201 inches and steady.

Based on the above indications, which of the following is the correct action to be taken?

A. Raise 10500 bus voltage via the 71T-4 tap changer.

B. Raise 10500 bus voltage via the EDG A(C) voltage regulator(s).

C. Enter and execute AOP-18, "Loss of 10500 Bus."

D. Enter and execute AOP-59, "Loss of RPS Bus A Power."

ES-401 Sample Written Examination Form ES-401-S Question Worksheet Examination Outline Cross-

Reference:

Level RO Tier # 1 Group # 1 KIA # 295003 AK 2.01 Importance Rating 4.2/4.0 Proposed Question:

The plant is operating at 100% power when the following occurs:

Transformers T2 and T3 trip due to high winds and flying debris

An Unusual Event is declared

Annunciator 09-8-2-24 "4160V BUS '10500 DEGRADED VOLTAGE TIMER INITIATED" alarms

No Operator actions are taken.

30 seconds after the annunciator is received you note the following indications:

10500 Bus voltage as indicated at the 09-8 panel is currently 3850 volts and steady.

Drywell pressure is 1.8 psig and steady.

RPV water level is 201 inches and steady.

Based on the above indications, which of the following is the correct action to be taken?

A. Raise 10500 bus voltage via the 71T-4 tap changer.

B. Raise 10500 bus voltage via the EDG A(C} voltage regulator(s}.

C. Enter and execute AOP-18, "Loss of 10500 Bus."

D. Enter and execute AOP-59, "Loss of RPS Bus A Power."

Proposed Answer:

A. Raise 10500 bus voltage via the 71T-4 tap changer.

Explanation (Optional):

A. Correct: Raise 10500 bus voltage via the 71T -4 tap changer. Based on the conditions listed in the question stem, 10500 bus voltage is greater than 2975 volts and there is no indication that a LOCA is occurring. Therefore, the auto starts of the A and C EDGs have not yet occurred (occur at 45 seconds after the timer initiates). The 10304 and 10514 tie breakers will still be shut, therefore raising 10500 bus voltage shall be accomplished via the 71T-4 tap changer.

B. Incorrect: Raise 10500 bus voltage via the EDG A(C) voltage regulator(s). Based on the conditions listed in the question stem, 10500 bus voltage is greater than 2975 volts and there is no indication that a LOCA is occurring. Therefore, the auto starts of the A and C EDGs have not yet occurred (occur at 45 seconds after the timer initiates). The 10304 and 10514 tie breakers will still be shut, therefore raising 10500 bus voltage shall be accomplished via the 71T-4 tap changer not the EDG A(C) voltage regulators.

Plausible if the candidate does not correlate the conditions provided in the stem.

C. Incorrect: Enter and execute AOP-18, "Loss of 10500 Bus." Based on the conditions provided in the question stem, the 10500 bus has not been lost. Therefore the entry conditions for AOP-18 have not been met. Plausible is the candidate cannot recall the entry conditions for AOP-18.

D. Incorrect: Enter and execute AOP-59, "Loss of RPS Bus A Power." Based on the conditions provided in the question stem, the 10500 bus has not been lost and thus RPS Bus A has not been lost either. Therefore the entry conditions for AOP-59 have not been met. Plausible is the candidate cannot recall the entry conditions for AOP-59.

Technical Reference(s): ARP-09-8-2-24, AOP-18, AOP-59 Proposed references to be provided to applicants during examination: None Learning Objective: SDLP-71E; 1.12.a Question Source: New Question History: New Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 10CFR55.41 (1 0)

Comments:

Question #3 The plant was conducting a shutdown \ cooldown when the following occurred:

Drywell pressure: 3.1 psig, rising 1 psig/min

Rx Pressure: 460 psig. lowering 6 psig/min

Bus 10500 tripped on a high differential current bus fault Assuming no operator actions for the next 10 minutes, complete the following statement concerning the OVERALL trend of DC battery voltages:

The "B" LPCI MOV batteries {71BAT-3B} voltage trend will be to and the "A" Station Batteries (71SB-1A) voltage trend will be to _ _ __

71BAT-3B 71SB-1A A. Remain the same Lower B. Lower Remain the same C. Remain the same Remain the same D. Lower Lower

ES-401 Sample Written Examination Form ES-401*5 Question Worksheet Examination Outline Cross-

Reference:

Level RO/SRO Tier # 1 Group # 1 KIA # 295004 AA 2.03 Importance Rating 2.8/2.9 PARRAL.

Proposed Question: A, Lower, Lower The plant was conducting a shutdown \ cooldown when the following occurred:

Drywell pressure: 3.1 psig, rising 1 psig/min

Rx Pressure: 460 psig, lowering 6 psig/min

Bus 10500 tripped on a high differential current bus fault Assuming no operator actions for the next 10 minutes, complete the following statement concerning the OVERALL trend of DC battery voltages:

The "B" LPCI MOV batteries (71BAT-3B) voltage trend will be to and the "An Station Batteries (71SB-1A) voltage trend will be to _ _ __

71BAT-3B 71SB-1A A. Remain the same Lower B. Lower Remain the same C. Remain the same Remain the same D. Lower Lower Proposed Answer: D

Explanation (Optional):

10500 supplies the "A" Station Battery charger. The EOGs will auto start to resupply 10500 loads but the bus fault (high differential current) will cause the EOGs to trip off the bus. So the Station Batteries (without a charger) continue to supply loads (such as RHR logic).

10600 supplies the LPCI "B" battery charger. However it lost the charger because a LOCA signal was generated (OW press >2.7 psig) which is a system design feature. Normally.

voltage would remain steady since there are no active loads on the bus. However, at 450 psig the LPCI injection valve will open (since there is a LOCA signal) and lower the battery voltage. Only after 10 minutes can the operators manually place the battery charger back in service.

Technical Reference(s):

Proposed references to be provided to applicants during examination:

Learning Objective: SOLP-71B, EO 1.10 Question Source: New Question History: New Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 10CFR55.41 (1 0)

Comments:

Question #4 A reactor startup is in progress. Reactor power is currently 23%. The main turbine is paralleled to the grid when the following annunciator is received in the Control Room:

09-5-2-09 "MAIN TURB TRIP" Based on the conditions above, which of the following responses are expected with respect to the (1) Reactor and (2) Main Turbine Bypass Valves?

(1) Reactor (2) Main Turbine B~Qass Valves A. Scram Occurs Bypass valves open to mitigate the resultant reactor core reactivity addition B. Scram Occurs Bypass valves open to prevent mechanical damage to the reactor internal components C. No Scram Occurs Bypass valves open to mitigate the resultant reactor core reactivity addition D. No Scram Occurs Bypass valves open to prevent exceeding ASME Code pressure limits

ES-401 Sample Written Examination Form ES-401*5 Question Worksheet Examination Outline Cross-

Reference:

Level RO/SRO Tier # 1 Group # 1 KIA # 295005 AK 3.07 Importance Rating 3.8/3.8 Proposed Question:

A reactor startup is in progress. Reactor power is currently 23%. The main turbine is paralleled to the grid when the following annunciator is received in the Control Room:

09-5-2*09 "MAIN TURB TRIP" Based on the conditions above, which of the following responses are expected with respect to the (1) Reactor and (2) Main Turbine Bypass Valves?

(1) Reactor (2) Main Turbine Bypass Valves A. Scram Occurs Bypass valves open to mitigate the resultant reactor core reactivity addition B. Scram Occurs Bypass valves open to prevent mechanical damage to the reactor internal components C. No Scram Occurs Bypass valves open to mitigate the resultant reactor core reactivity addition D. No Scram Occurs Bypass valves open to prevent exceeding ASME Code pressure limits Proposed Answer:

C. No Scram Occurs Bypass valves open to mitigate the resultant reactor core reactivity addition

Explanation (Optional):

Based on the conditions given in the question stem, with Reactor Power less than 29%, as the main turbine trips a reactor scram WILL NOT occur. The main turbine bypass valves are designed to bypass steam equivalent to 25% reactor power to the main condenser.

Therefore, all bypass valves will open to mitigate the resultant reactor core reactivity addition due to the pressure rise in the RPV. The bypass valves are not designed to protect the reactor internal components from mechanical damage.

Technical Reference(s): SDLP-.94A, OP-9 Proposed references to be provided to applicants during examination: None Learning Objective: SDLP-94A, EO 1.09.b,c,f Question Source: New Question History: New Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 10CFR55.41 (5)

Comments:

Question #5 The reactor is operating at 95% power following 531 days of high power operations. The following events occur:

The reactor scrams.

The main turbine trips AND the 4160V distribution buses fail to transfer to reserve power.

APRMs are downscale.

Immediate actions per AOP-1 "Reactor Scram" have been completed by the At the Controls (ATC) operator.

Based on the conditions above and assuming no further operator action, which one of the following choices correctly describes the status of torus water temperature?

A. Torus water temperature is steady since there is no steam being discharged to the torus via HPCI or RCIC.

B. Torus water temperature is steady since continued steam generation in the reactor is being directed to the Main Condenser.

C. Torus water temperature is slowly lowering since the reactor scram has resulted in a cooldown of the Reactor Pressure Vessel (RPV).

D. Torus water temperature is slowly rising due to steam being discharged from the RPV via the Safety Relief Valves (SRVs).

ES-401 Sample Written Examination Question Worksheet Examination Outline Cross-

Reference:

Level RO/SRO Tier# 1 Group # 1 K/A# 295006 AK 3.03 Importance Rating 3.8/3.9 for Proposed Question:

The reactor is operating at 95% power following 531 days of high power operations. The following events occur:

The reactor scrams.

The main turbine trips AND the 4160V distribution buses fail to transfer to reserve power.

APRMs are downscale.

Immediate actions per AOP-1 "Reactor Scram" have been completed by the At the Controls (ATC) operator.

Based on the conditions above and assuming no further operator action, which one of the following choices correctly describes the status of torus water temperature?

A. Torus water temperature is steady since there is no steam being discharged to the torus via HPCI or RCIC.

B. Torus water temperature is steady since continued steam generation in the reactor is being directed to the Main Condenser.

C. Torus water temperature is slowly lowering since the reactor scram has resulted in a cooldown of the Reactor Pressure Vessel (RPV).

D. Torus water temperature is slowly rising due to steam being discharged from the RPV via the Safety Relief Valves (SRVs).

Proposed Answer:

D. Torus water temperature is slowly rising due to steam being discharged from the RPV via the Safety Relief Valves (SRVs).

Explanation (Optional):

Based on the conditions listed in the question stem, when the reactor scram occurs concurrent with the main turbine trip, reactor pressure will rise due to the auto closure of the main turbine stop valves. Since reactor power is at a high level, the main turbine bypass valves do not have enough capacity to mitigate the pressure rise in the RPV. Additionally, with the failure of the 4160V buses to transfer to reserve power on the main turbine trip, all 3 circulating water pumps will be lost requiring manual closure of the MSIV's per AOP-1 immediate actions. In this condition, the only way that RPV pressure rise can be mitigated is via auto opening of the safety relief valves, which discharge steam from the RPV to the torus. This will cause torus water temperature to rise. Depending on the lowest RPV water level experienced on the reactor scram, HPCI and RCIC may start up to restore and maintain RPV water level, however this would input steam to the torus which would also cause a rise in torus water temperature, thus making distractor 'A' incorrect.

Technical Reference(s): AOP-1,OP-68 Proposed references to be provided to applicants during examination: None Learning Objective: SDLP-36, EO 1.09; LP-AOP, EO 1.10 Question Source: New Question History: New Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 10CFR55.41 (5)

Comments:

Question #6 The Control Room was abandoned due to toxic gas. A reactor shutdown was commenced per the guidance of AOP-43 "Plant Shutdown from Outside the Control Room."

A reactor cooldown is currently in progress.

During the last 30 minutes, reactor pressure was lowered from 1013 psig to 666 psig.

If the cooldown continues at the SAME RATE ('F/hour) for the next 30 minutes, what will be the status of cooldown rate limits \ targets?

A. No limits \ targets will be exceeded.

B. Administrative limits \ targets will be exceeded; but no technical specifications will be violated.

C. Administrative limits \ targets and technical specifications will be exceeded; thirty minutes is the maximum time allowed to restore cooldown limits.

D. Administrative limits \ targets and technical specifications will be exceeded; one hour is the maximum time allowed to restore cooldown limits.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level ROISRO Tier # 1 Group # 1 KJA# 295016 AA 2.03 Importance Rating 4.3/4.4 Proposed Question:

The Control Room was abandoned due to toxic gas. A reactor shutdown was commenced per the guidance of AOP-43 "Plant Shutdown from Outside the Control Room."

A reactor cooldown is currently in progress.

During the last 30 minutes, reactor pressure was lowered from 1013 psig to 666 psig.

If the cooldown continues at the SAME' RATE ('F/hour) for the next 30 minutes, what will be the status of cooldown rate limits \ targets?

A. No limits \ targets will be exceeded.

B. Administrative limits \ targets will be exceeded; but no technical specifications will be violated.

C. Administrative limits \ targets and technical specifications will be exceeded; thirty minutes is the maximum time allowed to restore cooldown limits.

D. Administrative limits \ targets and technical specifications will be exceeded; one hour is the maximum time allowed to restore cooldown limits.

Proposed Answer:

B. Administrative limits \ targets will be exceeded; but no technical specifications will be violated.

Explanation (Optional):

Based on the conditions stated in the question stem the reactor has cooled down approximately 48 degrees Fahrenheit in 30 minutes. Therefore, if the same cooldown rate is maintained for the subsequent 30 minutes, the hourly cooldown rate would be approximately 96 degrees Fahrenheit per hour. This violates the administrative cooldown rate limits \ target of ST-26J which are listed as 85-95 degrees Fahrenheit per hour. The hourly cooldown rate does not exceed the Tech Spec LCO 3.4.9 limit of 100 degrees Fahrenheit per hour. The candidates will be required to take the data given in the question stem, convert the gage pressures to absolute and then look up the corresponding saturation temperatures using ASME steam tables and or AOP-1 Press-Temp Attachment 7 in order to calculate the cooldown rate.

Technical Reference(s): AOP-43, Tech Spec LCO 3.4.9, ST-26J Proposed references to be provided to applicants during examination: Steam Tables, AOP-1 Attachment 7 Learning Objective: SDLP-02A, EO 1.07.a, 1.08.a Question Source: Modified Bank -INPO #27888 Question History: Cooper 2003 ILO Exam Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 10CFR55.41 (10)

Comments:

Parent Question from INPO Exam Bank - Cooper 2003 ILO Exam Control room abandonment is required due to toxic gas in the control room. The CRS at the Alternate Shutdown Panel commences a reactor cooldown. The CRS lowers reactor pressure from 900 psig to 550 psig in 30 minutes.

If the cooldown continues at the SAME RATE (F/minute) for the next thirty minutes, what will be the status of cooldown rate limits?

a. No limits will be exceeded.

b. Administrative limits will be exceeded; but no technical specifications will be violated.

c. Administrative limits and technical specifications will be exceeded; thirty minutes is the maximum time allowed to restore cooldown limits.

d. Administrative limits and technical specifications will be exceeded; one hour is the maximum time allowed to restore cooldown limits.

Correct Answer:

c. Administrative limits and technical specifications will be exceeded; thirty minutes is the maximum time allowed to restore cooldown limits.

Question #7 The plant is operating at 100% power with three RBClC heat exchangers in service.

Reactor Building Closed loop Cooling pumps 'A' and 'C' are running supplying cooling to their respective loads.

The following annunciator is received at the 09-6 panel:

09-6-2-31 "RBC HDR PRESS lO" Which one of the following choices correctly describes a potential cause for this annunciator?

A. The supply breaker for l-14, 11402 has tripped.

B. A significant tube leak has developed from the 'B' Non-Regenerative Heat Exchanger.

C. Annunciator 09-6-2-22 "RBC MAKEUP TK lVl HI OR lO" is in alarm.

D. One RBC heat exchanger has been removed from service.

ES-401 Sample Written Examination Form ES-401*5 Question Worksheet Examination Outline Cross-

Reference:

Level RO/SRO Tier # 1 Group # 1 KfA# 295018 AA 2.03 Importance Rating 3.2/3.5 Proposed Question:

The plant is operating at 100% power with three RBCLC heat exchangers in service.

Reactor Building Closed Loop Cooling pumps 'A' and 'C' are running supplying cooling to their respective loads.

The following annunciator is received at the 09-6 panel:

09-6-2-31 "RBC HDR PRESS LO" Which one of the following choices correctly describes a potential cause for this annunciator?

A. The supply breaker for L-14, 11402 has tripped.

B. A significant tube leak has developed from the 'B' Non-Regenerative Heat Exchanger.

C. Annunciator 09-6-2-22 "RBC MAKEUP TK LVL HI OR LO" is in alarm.

D. One RBC heat exchanger has been removed from service.

Proposed Answer:

C. Annunciator 09-6-2-22 "RBC MAKEUP TK LVL HI OR LO" is in alarm.

Explanation (Optional):

A. Incorrect: The supply breaker for L-14, 11402 has tripped. L-14 supplies power to the

'B' RBCLC which is not running based on the conditions given in the question stem.

Thus, the power loss could not be the cause for the RBC header pressure drop.

B. Incorrect: A significant tube leak has developed from the 'B' Non-Regenerative Heat Exchanger. This tube leak would result in leakage into the RBCLC system since the RWCU system is operating at a higher pressure than RBCLC. Therefore, a drop in RBCLC header pressure would not occur.

C. Correct: A low level in the Makeup Tank would lower NPSH to the running pumps and result in low pump discharge pressure.

D. Incorrect: Removing a heat exchanger from service would affect heat loading on the system. Changing the heat load would affect temperature which would affect water density.

Technical Reference(s): OP-40, ARP 09-6-2-22, ARP 09-6-2-31 Proposed references to be provided to applicants during examination: None Learning Objective: SDLP-15, EO 1.14.c Question Source: New Question History: New Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 10CFR55.41(10)

Comments:

Question #8 The plant is operating at 85% power. Power ascension to rated power is in progress following maintenance that was conducted on the CRD Hydraulic system.

The Reactor Building NPO has just notified the Control Room that he has observed an air leak on the west side of RB 272'

Instrument and service air header pressure is 106 psig and lowering at a rate of 1 psig per minute.

Based on the indications above:

(1) How many Instrument Air Compressors are currently running? AND (2) Assuming the air leak rate remains constant, how long until the Service Air Header Auto Isolation Valve (39FCV-11 0) auto closes?

(1 ) (2)

A. 2 11 minutes B. 3 11 minutes C. 2 21 minutes D. 3 21 minutes

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO/SRO Tier# 1 Group # 1 KIA # 295019 AA 2.01 Importance Rating 3.5/3.6 Proposed Question:

The plant is operating at 85% power. Power ascension to rated power is in progress following maintenance that was conducted on the CRD Hydraulic system.

The Reactor Building NPO has just notified the Control Room that he has observed an air leak on the west side of RB 272'

Instrument and service air header pressure is 106 psig and lowering at a rate of 1 psig per minute.

Based on the indications above:

(1) How many Instrument Air Compressors are currently running? AND (2) Assuming the air leak rate remains constant, how long until the Service Air Header Auto Isolation Valve (39FCV-110) auto closes?

(1 ) (2)

A. 2 11 minutes B. 3 11 minutes C. 2 21 minutes D. 3 21 minutes Proposed Answer:

A. 2 11 minutes

Explanation (Optional):

Based on the conditions provided in the question stem, with instrument air header pressure at 106 psig, there should be a total of 2 air compressors running. Per AOP-12, the first standby air compressor starts at 107 psig and the second standby air compressor will not start until air header pressure lowers to 104 psig. With the air leak rate at 1 psig per minute it will take 11 minutes to lower to the auto isolation setpoint of 95 psig for the service air header auto isolation valve (39FCV-110). The breathing air header auto isolation valve auto close setpoint is at 85 psig (therefore it will take 21 minutes to reach the closure setpoint).

Technical Reference(s): AOP-12 Proposed references to be provided to applicants during examination: None Learning Objective: SDLP-39, EO 1.08.a,c, and 1.14.b Question Source: New Question History: New Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 10CFR55.41(10)

Comments:

Question #9 The plant is on day 10 of a Refueling Outage when a loss of shutdown cooling occurs. The Control Room operators enter and execute AOP-30 "Loss of Shutdown Cooling."

Per AOP-30, Reactor Water Cleanup flow rate is (1) _ _ _ _ _ _ _ _ _ in order to (2) _ _ _ _ _ __

ill A. lowered minimize reactor vessel inventory loss B. lowered minimize the possibility of thermal stratification in the RPV C. raised maximize heat removal rate through the regenerative heat exchanger D. raised maximize heat removal rate through the non-regenerative heat exchanger

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO/SRO Tier# 1 Group # 1 KIA # 295021 AK 3.04 Importance Rating 3.3/3.4 Proposed Question:

The plant is on day 10 of a Refueling Outage when a loss of shutdown cooling occurs. The Control Room operators enter and execute AOP-30 "Loss of Shutdown Cooling."

Per AOP-30, Reactor Water Cleanup flow rate is (1) _ _ _ _ _ _ _ _ _ _ in order to (2}._ _ _ _ _ _,

ill ill A. lowered minimize reactor vessel inventory loss B. lowered minimize the possibility of thermal stratification in the RPV C. raised maximize heat removal rate through the regenerative heat exchanger D. raised maximize heat removal rate through the non-regenerative heat exchanger Proposed Answer:

D. raised maximize heat removal rate through the non-regenerative heat exchanger

Explanation (Optional):

Per AOP-30 "Loss of Shutdown Cooling", immediate action E.4 directs control of reactor coolant temperature per Attachment 3. Attachment 3 directs establishing a RPV cooldown rate of less than 80 degree Fahrenheit per hour using any of the methods listed on . RWCU is one of the methods listed in Attachment 6. Therefore, on a loss of shutdown cooling, RWCU flow rate is raised in order to maximize heat removal from the RPV through the non-regenerative heat exchanger. Distractors A and B are incorrect because they direct lowering RWCU flow rate in order to change parameters other than reactor coolant temperature. Distractor C is incorrect since the regenerative heat exchanger is not rejecting heat from the reactor coolant to a separate cooling medium.

Technical Reference(s): AOP-30 Proposed references to be provided to applicants during examination: None Learning Objective: SDLP-10, EO 1.09.e Question Source: Bank -INPO Exam Bank 20028 Question History: Dresden - 2001 ILO Exam Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 10CFR55.41 (5)

Comments:

Parent Question from INPO Exam Bank - Dresden 20011LO Exam Following a loss of shutdown cooling, RWCU system flow rate is (1) _ _ _ _ _ _ __

if possible, in order to (2) _ _ _ _ _ _ _.

a. (1) increased (2) maximize heat removal rate through the non-regenerative heat exchanger

b. (1) reduced (2) minimize the possibility of thermal stratification

c. (1) increased (2) maximize heat removal rate through the regenerative heat exchanger

d. (1) reduced (2) minimize reactor vessel inventory loss.

Correct Answer:

a. (1) increased (2) maximize heat removal rate through the non-regenerative heat exchanger

Question #10 While operating in Mode 5 with the Mode Selector Switch in "REFUEL" the --:-----l(....!.1J...}- : - : - _ - :

in conjunction with (2) ensures that inadvertent criticality is not achieved while refueling.

ill m A. Rod Seq Sel Switch placed in the proper rod withdrawal sequence "WITHDRAW" loaded into the Rod Worth Minimizer (RWM)

B. Rod Seq Sel Switch placed in adherence to the approved fuel movement "WITHDRAW" plan during fuel moves

c. One-Rod-Out interlock adherence to the approved fuel movement plan during fuel moves D. One-Rod-Out interlock the proper rod withdrawal sequence loaded into the Rod Worth Minimizer (RWM)

ES*401 Sample Written Examination Form ES~401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO/SRO Tier # 1 Group # 1 KIA # 295023 AK 1.02 Importance Rating 3.2/3.6 Proposed Question:

While operating in Mode 5 with the Mode Selector Switch in "REFUEL" the --:---I(...:..1.l-}_ __

in conjunction with (2) ensures that inadvertent criticality is not achieved while refueling.

ill m A. Rod Seq Sel Switch placed in the proper rod withdrawal sequence "WITHDRAW" loaded into the Rod Worth Minimizer (RWM)

B. Rod Seq Sel Switch placed in adherence to the approved fuel movement "WITHDRAW" plan during fuel moves

c. One-Rod-Out interlock adherence to the approved fuel movement plan during fuel moves D. One-Rod-Out interlock the proper rod withdrawal sequence loaded into the Rod Worth Minimizer (RWM)

Proposed Answer:

c. One-Rod-Out interlock adherence to the approved fuel movement plan during fuel moves

Explanation (Optional):

Per Technical Specification bases Section 3.9.2, in order to prevent inadvertent criticality during Mode 5, the refuel position one-rod-out interlock ensures no more than one control rod may be withdrawn thus providing protection against prompt reactivity excursions.

Additionally, in conjunction with the one-rod-out interlock per Tech Spec bases Section 3.1.1 (specifically SR 3.1.1.1) an evaluation of each in vessel fuel movement during fuel loading is required to ensure adequate SDM is maintained during refueling. This evaluation ensures that the intermediate loading patterns are bounded by the safety analyses for the final core loading pattern. Even though the Sequence lVlode Select switch does physically allow withdrawal of control rods when in the "WITHDRAW" position it does not assist in prevention of inadvertent criticality. Also, even though the Rod Worth Minimizer enforces a prescribed rod withdrawal pattern in order to prevent significant core damage, it does not enforce the withdrawal pattern while performing refueling operations.

Technical Reference(s): Tech Spec Bases B 3.1.1 and B 3.9.2 Proposed references to be provided to applicants during examination: None Learning Objective: SDLP-03F, EO 1.09.d Question Source: Bank -INPO Exam Bank #24857 Question History: Duane Arnold 2002 [LO Exam Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 10CFR55.41 (8)

Comments:

Parent Question from INPO Exam Bank - Duane Arnold 2002 ILO Exam Refueling operations are in progress with the Mode Switch in "REFUEL".

Which of the following allows the withdrawal of a control rod AND ensures the reactor will remain shutdown?

a. The One-Rod permissive interlock and following the fuel moving plan during fuel moving.

b. The RSCS Group Selector switch selected to the correct sequence and adequate shutdown margin designed into the core.

c. Refueling Rod Block Interlocks and the proper rod withdrawal sequence loaded into the RWM.

d. The RSCS Mode Selector Switch placed in "WITHDRAW" and following the fuel moving plan during fuel moving.

Correct Answer:

a. The One-Rod permissive interlock and following the fuel moving plan during fuel moving.

Question #11 The plant has experienced a loss of coolant accident. The following plant conditions exist:

Orywell pressure: 6 psig.

Orywell temperature: 220'F.

Torus pressure: 7 psig

Torus Water temperature: 110'F If Orywell sprays were currently in service, the lowest Orywell pressure could go before being automatically secured would be (1) psig.

If Orywell sprays were not in service then sprays _~(.:.2L.)__ to be placed in service.

(1) (2)

A. o are NOT allowed B. o are allowed

c. 2.7 are NOT allowed O. 2.7 are allowed

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO/SRO Tier # 1 Group # 1 KJA# 295024 EA 1.17 Importance Rating 3.9/3.9 Proposed Question:

The plant has experienced a loss of coolant accident. The following plant conditions exist:

Drywell pressure: 6 psig.

Drywell temperature: 220'F.

Torus pressure: 7 psig

Torus Water temperature: 110'F If Drywell sprays were currently in service. the lowest Drywell pressure could go before being automatically secured would be (1) psig.

If Drywell sprays were not in service then sprays _ ........(. ; ; ;2......

>__ to be placed in service.

(1 ) (2)

A. 0 are NOT allowed B. 0 are allowed C. 2.7 are NOT allowed D. 2.7 are allowed Proposed Answer: C: 2.7. are NOT allowed

Explanation (Optional):

EOP-4 would have been entered when OW pressure reached 2.7 psig and OW temp reached 135 deg F. OW Spray could have been placed in service to maintain OW temp less than 309 deg F. Once in service, per EOP-4 direction, they must be secured before pressure drops below 0 psig. However, they will automatically secure if OW pressure reaches 2.7 psig. If not in service, they can only be started if the OW Spray Initiation Limit is met. Although OW spray uses Torus pressure (15 psig) as a bases to spray the OW, the OW Spray Initiation Limit is only based on OW parameters. Using OW parameters, the plant is in the unsafe region of the curve and can not spray.

Technical Reference(s): EOP-4, EOP-11 Proposed references to be provided to applicants during examination:

Learning Objective: MIT -301.11 E, EO 4.05 Question Source: New Question History: New Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 10CFR55.41(10)

Comments:

Question #12 The Plant is operating at 100% when the High Pressure Coolant Injection (HPCI) pump inadvertently initiates and injects into the RPV.

With no Operator action, which one of the following correctly states the response of the Plant with regards to (1) Reactor Power and (2) Main Turbine Control?

(1) Reactor Power (2) Turbine Control A. Lowers Turbine Control Valves close B. Rises Turbine Stop Valves close C. Lowers Turbine Intermediate Stop Valve close D. Rises Turbine Control Valves open

ES-401 Sample Written Examination Form ES*401*5 Question Worksheet Examination Outline Cross-

Reference:

Level RO/SRO Tier # 1 Group # 1 KIA # 295025 EK3.08 Importance Rating 3.5/3.5 Proposed Question:

The Plant is operating at 100% when the High Pressure Coolant Injection (HPCI) pump inadvertently initiates and injects into the RPV.

With no Operator action, which one of the following correctly states the response of the Plant with regards to (1) Reactor Power and (2) Main Turbine Control?

(1) Reactor Power (2) Turbine Control A. Lowers Turbine Control Valves close B. Rises Turbine Stop Valves close C. Lowers Turbine Intermediate Stop Valve close D. Rises Turbine Control Valves open Proposed Answer:

D Rises Turbine Control Valves open

Explanation: A HPCI initiation and injection at full power causes Rx power to rise due to cold water addition. Although the opening of the HPCI inlet valve causes a decrease in power, more voiding and a power decrease; it is compensated for by EHC control and the large positive reactivity of the cold water. As Rx power rises, Main Steam pressure rises. The Control Valves are not fully open at 100% power and will open further to the setpoint of EHC Load Limit (111%).

A. incorrect- If voiding was the primary reactivity event, then control valves would close as pressure dropped.

B. incorrect- power does rise but not enough to pick up the APRM scram signal which would then cause a turbine trip and the stop valves would close.

C. incorrect- If voiding was the primary reactivity event, then pressure would drop until the MSIVs closed (low pressure in Run isolation), scrammed the reactor and cause a turbine trip which would close the Intermediate Stop Valves.

D. correct-see above explanation Technical Reference(s): OP-9 Proposed references to be provided to applicants during examination: None Learning Objective: SDLP-23 1.14 SDLP-94C 1.05 Question Source: New Question History: New Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 10CFR55.41(5)

Comments:

Question #13 The plant is currently performing ST-4N "HPCI Quick Start, Inservice, and Transient Monitoring Test". Per the ST, the RHR system must be placed in Torus Cooling mode prior to testing the HPCI turbine.

Placing the RHR system in torus cooling is required to maintain torus water temperature below 105°F because continued HPCI operation ...

A. can result in mechanical damage to HPCI components.

B. can result in pump cavitation.

C. could result in damage to the torus T-quenchers.

D. could result in inadequate heat absorption capability of the torus.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO/SRO Tier # 1 Group # 1 KJA# 295026 EK 3.02 Importance Rating 3.9/4.0 Proposed Question:

The plant is currently performing ST-4N "HPCI Quick Start, Inservice, and Transient Monitoring Test". Per the ST, the RHR system must be placed in Torus Cooling mode prior to testing the HPCI turbine.

Placing the RHR system in torus cooling is required to maintain torus water temperature below 105°F because continued HPCI operation ...

A. can result in mechanical damage to HPCI components.

B. can result in pump cavitation.

C. could result in damage to the torus T-quenchers.

D. could result in inadequate heat absorption capability of the torus.

Proposed Answer:

D. could result in inadequate heat absorption capability of the torus.

Explanation (Optional):

Placing the RHR system in torus cooling is required to maintain torus water temperature below 105°F because continued HPCI operation ...

A. Incorrect: can result in mechanical damage to HPCI components. Per the precautions in OP-15, equipment damage will not occur to HPCI components unless torus water temperature is greater than 170°F.

B. Incorrect: can result in pump cavitation. Per the precautions in OP-15, equipment damage will not occur to HPCI components unless torus water temperature is greater than 190°F.

C. Incorrect: could result in damage to the torus T-quenchers. Plausible is the candidate determines that rising torus water temperature could lead to a rise in torus water level to one to a value that would threaten torus internal components.

However, this is not the basis for the requirement in the ST.

D. Correct: could result in inadequate heat absorption capability of the torus.

Continued operation of HPCI could cause heat addition to the torus and drive temperature to Tech Spec limits of 110 and 120 of respectively which are in place to ensure maintenance of heat removal capability of the suppression pool.

Technical Reference(s): OP-15, ST-4N Proposed references to be provided to applicants during examination: None Learning Objective: SDLP-23, EO 1.13.a Question Source: New Question History: New Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 10CFR55A1 (5)

Comments:

Question #14 A large break loss of coolant accident (LOCA) has occurred.

The operators in the Control Room are taking actions per EOP-2 and EOP-4.

Drywell pressure is 30 psig and rising.

Drywell temperature is 270'F and rising.

Per EOP-4 "Primary Containment Control", an NPO has been dispatched by the SNO to (1) the RBCLC cooling supply to the drywell coolers and drywell equipment sump cooler in order to (2)_________

ill m A. isolate maintain adequate cooling to other RBCLC cooled components B. isolate prevent piping rupture due to voiding and water hammer C. restore lower temperature in the drywell by providing a heat sink D. restore lower pressure in the drywell by condensing steam generated by the leak

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO/SRO Tier# 1 Group # 1 KIA # 295028 G 2.4.35 Importance Rating 3.8/4.0 Proposed Question:

A large break loss of coolant accident (LOCA) has occurred.

The operators in the Control Room are taking actions per EOP-2 and EOP-4.

Drywell pressure is 30 psig and rising.

Drywell temperature is 270'f and rising.

Per EOP-4 "Primary Containment Control", an NPO has been dispatched by the SNO to (1) the RBCLC cooling supply to the drywell coolers and drywell equipment sump cooler in order to (2)_ _ _ _ _ _ __

ill m A. isolate maintain adequate cooling to other RBCLC cooled components B. isolate prevent piping rupture due to voiding and water hammer C. restore lower temperature in the drywell by providing a heat sink D. restore lower pressure in the drywell by condensing steam generated by the leak Proposed Answer:

B. isolate prevent piping rupture due to voiding and water hammer

Explanation (Optional):

Based on the conditions provided in the question stem, the drywell temperature listed warrants executing the override from the Drywell Temperature leg of EOP-4 to isolate the cooling water supply to the drywell coolers, drywell equipment sump cooler, and RWR pump and motor coolers per EP-12 if drywell temperature exceeds 260'F. Isolating these cooling water supply lines helps to minimize the chance of losing primary containment integrity due to a RBCLC pipe rupture inside containment due to voiding and water hammer. Per precautions of OP-40 and AOP-11, when a LOCA condition exists, re-initiating RBC flow to drywell components is prohibited. Therefore, even though the reasons for restoring RBC flow to these components are plausible to help mitigate drywell environmental parameters (temperature and pressure), they are not allowed in the condition listed in the question stem.

Isolating RBC flow to the drywell components would ensure adequate RBC temperatures to the remainder of the RBC cooled components, but it is not the reason for isolation.

Technical Reference(s): EP-12, EOP-4, AOP-11, OP-40 Proposed references to be provided to applicants during examination: None Learning Objective: SDLP-15, EO 1.12.a, EO 1.15.a Question Source: New Question History: New Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 10CFR55.41 (1 0)

Comments:

Question #15 The plant is conducting a normal reactor shutdown with the 10600 bus out of service. During the shutdown, a loss of coolant accident (LOGA) occurs.

Plant conditions are as follows:

Torus Pressure is 3.0 psig.

Torus Level is 6.0 feet.

Torus Temperature is 190'F.

RPV water level is 58 inches and lowering slowly.

RPV pressure is 200 psig.

LPGI injection flow rate is 6500 gpm per loop.

Based on the conditions above, which one of the following is required with respect to RHR Pump NPSH limits while maintaining maximum LPGI injection flow rate?

A. No action is necessary.

B. Lower LPGI injection rate to 2000 gpm per loop.

C. Lower LPGI injection rate to 3000 gpm per loop.

D. Raise LPGI injection rate to 10,000 gpm per loop.

RHR PUMP NPSH LIMIT 250.00 1:- 1 pW; op;a6Og~ I 240.00 2 , ~_._, ..Opataling' L-..: PI/mpl ~_

230.00

_ 220.00 II.

! -Qp

~

2:10.00 200.00 - ""- ---- ~

ig OVerprenure'

~

190.00

: 110.00, UI

~

170.00

............... .... -~

.... ~p ~ OVerpre$IUre*

i'-,..i l- 160.00 150.00

" ""- ~J ..

P 5Ig OVerpressure" 140.00 i\. i 130.00 't o 1 2 3 .. 5 6 1 10 11 SINGLE lOOP RHR

FLOW PER PUMP (gpm X 1000)

"TorusOVerpreuun: " Torus Preuure + 0.4 (Torus WaterLllllel

1.92)

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO/SRO Tier# 1 Group # 1 KJA# 295030 EK 2.04 Importance Rating 3.7/3.8 Proposed Question:

The plant is conducting a normal reactor shutdown with the 10600 bus out of service. During the shutdown, an earthquake occurs which causes a loss of coolant accident (LOCA).

Plant conditions are as follows:

Torus Pressure is 3.0 psig.

Torus Level is 6.0 feet.

Torus Temperature is 190'F.

RPV water level is 58 inches and lowering slowly.

RPV pressure is 200 psig.

LPCI injection flow rate is 6500 gpm per loop.

Based on the conditions above, which one of the following is required with respect to RHR Pump NPSH limits while maintaining maximum LPCI injection flow rate?

A. No action is necessary.

B. Lower LPCI injection rate to 2000 gpm per loop.

C. Lower LPCI injection rate to 3000 gpm per loop.

D. Raise LPCI injection rate to 10,000 gpm per loop.

RHR PUMP NPSH LIMIT 250.00 240,00 1:-12 P~Op...a~

Pu!npe Operalilg' I

L-...: - ---..*. ,,0---

230.00 lI.

220.00

- r-- """ bia 1210.00

!!;( 200.00 - r-....... - ..... .......

~~

~

a:

~~

190.00 180.00, 170.00

--........ ...... r--.......

~~

~

fig Overpressure" g 1&0.00

"" !'-o..

1 150.00 140.00

&30.00

" ~ i~ ~

~

OVerpressure" o 1 2 3 . 5 II 10 SINGLE LOOP RHR

FLOW PER PUMP (gpm X 1OOQ)

"'orus 01/el1lre$llure = Toruli Pressure'" 0.4 (Torus Walter Level

1.92)

Proposed Answer:

C. Lower LPCI injection rate to 4000 gpm.

Explanation (Optional):

Based on the conditions listed in the question stem, at the current LPCI flow rate and torus temperature, the RHR pumps are being operated above their respective NPSH limits. Based on the curves provided from OP-13A, torus overpressure is approximately 4.6 psig. To determine the NPSH limit for a single loop of RHR, the candidate must use the 0 psig overpressure curves. Since the 10600 bus is out of service, there can only be one RHR pump operating per loop of RHR, therefore the 1 pump operating curve should be used. At 6500 gpm, with a torus temperature of 190'F, the RH R pump is operating at a pOint exceeding the NPSH limit, therefore LPCI flow rate should be lowered to 4000 gpm to be within the limits of the curve. If the candidates misinterpret the curves by either assuming 2 pumps are operating in a RHR loop or use the 5 psig overpressure curves, the other distractors become plausible.

Technical Reference(s): OP-13A, EOP-4 Proposed references to be provided to applicants during examination: None Learning Objective: SDLP-10, EO 1.10J Question Source: New Question History: New Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 10CFR55.41 (7)

Comments:

Question #16 The plant was operating at power when a series of events result in a scram. All control rods do NOT fully insert and reactor power is greater than 2.5%.

EOP-3 "Failure to Scram" has been entered. The Control Room Supervisor has directed the SNO to take action to achieve a RPV water level between -19 and +110 inches.

What is the Bases for a maximum RPV water level of +110 inches while experiencing the conditions listed above?

A. To provide more core inlet subcooling.

B. To raise the amount of natural circulation in the RPV.

C. To mitigate the consequences of any irregular neutron flux oscillations.

D. To lower water inventory thus raising the concentration of boron when added.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO/SRO Tier# 1 Group # 1 KIA # 295031 G 2.4.18 Importance Rating 3.3/4.0 Proposed Question:

The plant was operating at power when a series of events result in a scram. All control rods do NOT fully insert and reactor power is greater than 2.5%.

EOP-3 "Failure to Scram" has been entered. The Control Room Supervisor has directed the SNO to take action to achieve a RPV water level between -19 and +110 inches.

What is the bases for a maximum RPV water level of +110 inches while experiencing the conditions listed above?

A. To provide more core inlet subcooling.

B. To raise the amount of natural circulation in the RPV.

C. To mitigate the consequences of any irregular neutron flux oscillations.

D. To lower water inventory thus raising the concentration of boron when added.

Proposed Answer:

C. To mitigate the consequences of any irregular neutron flux oscillations.

Explanation (Optional):

Based on the conditions stated in the question stem, the SNO will terminate and prevent all injection into the RPV with the exception of RCIC, CRD, and SLC per the guidance in EP-5 in order to lower RPV water level to less than 110". This is done in order to prevent or mitigate the consequences of any large irregular neutron flux oscillations induced by neutronic/thermal hydraulic instabilities by lowering RPV water level below the elevation of the feedwater sparger nozzles. This places the feedwater spargers in the steam space providing effective heating of the relatively cold feedwater and eliminating the potential for high core inlet subcooling. Choice 'A' is incorrect since lowering RPV water level intentionally lowers the amount of core inlet subcooling per the explanation above. Choice

'B' is incorrect since lowering level will lower the amount of natural circulation potential in the RPV due to the lower thermal driving head available. Choice 'D', although a true statement in theory, is not the reason for intentionally lowering RPV water level.

Technical Reference(s): EOP-3, MIT-301.11D Proposed references to be provided to applicants during examination: None Learning Objective: MIT-301.11D, EO 1.07 Question Source: Bank - INPO Exam Bank 23478 Question History: Columbia 2003 ILO Exam Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 10CFR55.41 (1 0)

Comments:

Parent Question from INPO Exam Bank - Columbia 2003 ILO Exam The plant was operating at power when a series of events result in a scram. All control rods do not fully insert and reactor power is greater that 5%.

EOP 5.1.2 has been entered. The level leg directs that RPV level be maintained between 65" and -192".

What is the reason for lowering level to at least -65" in this condition?

a. To increase core inlet subcooling.

b. To increase the amount of natural circulation.

c. To mitigate the consequences of any irregular neutron flux oscillations.

d. To reduce water inventory thus increasing the concentration of boron when added.

Correct Answer:

c. To mitigate the consequences of any irregular neutron flux oscillations.

Question #17 Given the following conditions:

An automatic RPS scram has occurred

Multiple control rods remain fully withdrawn from the core

Reactor power is 7%

EOP-3 "Failure to Scram" is being executed

SLC System 'A' has been injecting into the RPV for 3 minutes

The Reactor Mode Selector Switch is in "SHUTDOWN"

RPV pressure is being controlled by the Main Turbine Bypass Valves Based on the conditions above, an inadvertent cooldown of the RPV will cause:

A. reactor power to rise or cause the reactor to return to criticality.

B. the MSIVs to close, thereby challenging primary containment integrity.

C. less accurate Fuel Zone RPV water level indication due to the instrument's calibration conditions.

D. less effective reactor power reduction by SLC due to precipitation of sodium penta borate at lower reactor coolant temperatures.

ES*401 Sample Written Examination Form ES*401*5 Question Worksheet Examination Outline Cross-

Reference:

Level RO/SRO Tier # 1 Group # 1 KIA # 295037 EK 1.06 Importance Rating 4.0/4.2 Proposed Question:

Given the following conditions:

An automatic RPS scram has occurred

Multiple control rods remain fully withdrawn from the core

Reactor power is 7%

EOP-3 "Failure to Scram" is being executed

SLC System 'A' has been injecting into the RPV for 3 minutes

The Reactor Mode Selector Switch is in "SHUTDOWN"

RPV pressure is being controlled by the Main Turbine Bypass Valves Based on the conditions above, an inadvertent cooldown of the RPV will cause:

A. reactor power to rise or cause the reactor to return to criticality.

B. the MSIVs to close, thereby challenging primary containment integrity.

C. less accurate Fuel Zone RPV water level indication due to the instrument's calibration conditions.

D. less effective reactor power reduction by SLC due to precipitation of sodium penta borate at lower reactor coolant temperatures.

Proposed Answer:

A. reactor power to rise or cause the reactor to return to criticality.

Explanation (Optional):

Based on the conditions given in the question stem, while the plant experiences an A TWS an inadvertent cooldown of the RPV will result in reactor power to rise or return to criticality due to the effects of the temperature reduction on the moderator coefficient of reactivity.

Since the reactor mode switch is in the shutdown position, even if reactor pressure lowers to less than 850 psig, the MSIVs will not close due to the mode switch out of "run" interlock. As reactor coolant temperature lowers, the fuel zone level indication of RPV water level will become more accurate since the instrument is calibrated at 212'F and 0 psig. Sodium penta borate will precipitate out of solution at low temperatures (per OP-17, SLC system temperature is maintained greater than 90'F in orde r to keep the poison in solution), but based on the conditions given, the reactor coolant temperatures in the RPV would not support SLC precipitation. Thus, choice 'A' is the correct response.

Technical Reference(s): EOP-3, OP-17, MIT-301.11D Proposed references to be provided to applicants during examination: None Learning Objective: MIT-301.11D, EO 1.03 and EO 1.07 Question Source: Bank-INPO Exam Bank #15329 Question History: Monticello - 1999 ILO Exam Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 10CFR55.41(8 through 10)

Comments:

Parent Question from INPO Exam Bank - Monticello 1999 ILO Exam Given the following conditions:

- ATWS Event in progress

- SBLC System 1 injecting

- Immediate SCRAM actions complete

- Reactor pressure being maintained by the Turbine Bypass Valves What effect could an inadvertent Reactor cool down during an ATWS event have regarding the plant?

A Cooldown of the Reactor will cause ...

a. Reactor power to increase or cause the Reactor to return to criticality.

b. SBLC effectiveness to increase due to more boron being dissolved in the Reactor coolant at lower temperatures.

c. the MSIVs to close at 840 psig thereby further challenging primary containment.

d. unreliable Safeguards level instrument readings due to the instruments being calibrated hot.

Correct Answer:

a. Reactor power to increase or cause the Reactor to return to criticality.

Question #18 The plant is operating at rated power when the following annunciator alarms in the Control Room due to exceeding the setpoint for the RBC RAD MONITOR (17RM-352):

09-3-2-30 "LIQUID PROCESS RAD MONITOR HI-HI" Which one of the choices below correctly states:

(1) where the SNO can monitor the actual rad level indication from 17RM-352 AND (2) the potential cause for the 17RM-352 alarm?

ill @

A. 09-3 Panel RBClC heat exchanger tube leak B. 09-3 Panel Fuel Pool Cooling heat exchanger tube leak C. 09-10 Panel RBClC heat exchanger tube leak D. 09-10 Panel Fuel Pool Cooling heat exchanger tube leak

ES*401 Sample Written Examination Form ES*401*5 Question Worksheet Examination Outline Cross-

Reference:

Level RO/SRO Tier # 1 Group # 1 KIA # 295038 EA 1.03 Importance Rating 3.7/3.9 Proposed Question:

The plant is operating at rated power when the following annunciator alarms in the Control Room due to exceeding the setpoint for the RBC RAD MONITOR (17RM-352):

09-3-2-30 "LIQUID PROCESS RAD MONITOR HI-HI" Which one of the choices below correctly states:

(1) where the SNO can monitor the actual rad level indication from 17RM-352 AND (2) a potential cause for the 17RM-352 alarm?

ill m A. 09-3 Panel RBCLC heat exchanger tube leak B. 09-3 Panel Fuel Pool Cooling heat exchanger tube leak C. 09-10 Panel RBCLC heat exchanger tube leak D. 09-10 Panel Fuel Pool Cooling heat exchanger tube leak Proposed Answer:

D. 09-10 Panel Fuel Pool Cooling heat exchanger tube leak

Explanation (Optional):

Based on the information provided in the question stem, Annunciator 09-3-2-30 "LIQUID PROCESS RAD MONITOR HI-HI" alarms due to exceeding the setpoint for RBC RAD MON 17RM-352. This annunciator, which is located at the 09-3 panel in the Control Room, is a common annunciator that will alarm is the setpoints are exceeded for the process rad monitors in either the RBC, Radwaste, or Service Water systems. The candidate is told in the question stem that the alarm is due to exceeding the setpoint for the RBClC rad monitor. Even though the alarm comes in at the 09-3 panel, the actual rad level has to be read at the 09-10 panel which is located behind the 09-3 panel in the Control Room. i7RM 352 will alarm from any of the following causes (per ARP 09-3-2-30): fuel pool cooling heat exchanger tube leak, RWCU non-regenerative heat exchanger tube leak, or high background radiation in the area surrounding the radiation element (which is on the 300' elevation of the Reactor Building). A RBClC heat exchanger tube leak would result in an alarm from the Service Water Rad Monitor 17RM-351 due to the pressure difference between the RBC and Service Water systems. Thus, choice 'D' is the correct response.

Technical Reference(s}: OP-3i , ARP-09-3-2-30 Proposed references to be provided to applicants during examination: None learning Objective: SDlP-17, EO 1.05.a.6, EO 1.ii.c.i, EO 1.14.d.9 Question Source: New Question History: New Question Cognitive level: Comprehension or Analysis 10 CFR Part 55 Content: 10CFR55.41 (7)

Comments:

Question #19 A fire has ignited in the RWR Motor Generator Set room rendering it inaccessible to members of the Fire Brigade.

This area is supplied with (1) for fire suppression. The fire suppression system in this area can be activated by (2) ill m A. CO2 heat sensed in the area only B. CO 2 heat sensed in the area or manual pushbutton on the FPP C. water sprinklers heat sensed in the area only D. water sprinklers heat sensed in the area or manual pushbutton on the FPP

ES*401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO/SRO Tier # 1 Group # 1 KIA # 600000 AK 2.01 Importance Rating 2.6/2.7 Proposed Question:

A fire has ignited in the RWR Motor Generator Set room rendering it inaccessible to members of the Fire Brigade.

This area is supplied with (1) for fire suppression. The fire suppression system in this area can be activated by _ _ _.. (2=..)<--_

ill W A. CO 2 heat sensed in the area only B. CO 2 heat sensed in the area .QLmanual pushbutton on the FPP C. water sprinklers heat sensed in the area only D. water sprinklers heat sensed in the area .QLmanual pushbutton on the FPP Proposed Answer:

D. water sprinklers heat sensed in the area and manual pushbutton on the FPP

Explanation (Optional):

Per OP-33, the RWR MG set room is equipped with water sprinklers for fire suppression.

Since the fire in the RWR MG set room has rendered the area inaccessible, manual hose stations and CO2 fire extinguishers are not able to be used to suppress the fire manually.

The water sprinkler system that supplies fire water to the RWR MG set room is a supervised preaction type assembly. This means that it takes two independent actions to occur prior to allowing fire water to flow into the area via the sprinkler head; the heat in the area must be high enough to open the fused sprinkler heads AND the flow control valve must be opened by one of the following methods: temperature switch activation, local breakglass station pushbutton, pushbutton on the FPP, or manual release lever at the flow control valve.

Choice'S' is plausible if the candidate does not recall that the Relay Room is the only area that requires manual initiation of CO 2 . Choice 'C' is plausible if the candidate does not recall that fire suppression system for the RWR MG set room is designed as a preaction unit.

Choice 'A' is plausible if the candidate does not recall that the area is protected by water sprinklers versus CO 2 .

Technical Reference(s): TRM Section 3.7, SDLP-76, OP-33 Proposed references to be provided to applicants during examination: None Learning Objective: SDLP-76: EO 1.05.a.8, EO 1.05.c.3.e, EO 1.06.e Question Source: New Question History: New Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 10CFR55.41(7,8)

Comments:

Question #20 The plant is conducting a reactor startup with the following conditions:

Reactor power is currently 7%.

Main Turbine shell warming is in progress

A heavy rain storm with lightning is occurring.

The following sequence of events occurs:

At T = 0 seconds: lightning severs NMP-FITZ 115 KV Line #4 in the swltchyard, rendering Line #4 out of service.

At T = 10 seconds: a Line fault on LHH-FITZ 115KV Line #3 causes the LHH-FITZ 115KV LINE 3 BKR 10022, to trip.

At T 70 seconds: the Line fault on LHH-FITZ 115KV Line #3 clears.

Based on the conditions above, which one of the following correctly states the status of the 4160V buses listed below?

10100 Bus 10200 Bus A. Energized Energized B. De-energized Energized C. Energized De-energized D. De-energized De-energized

ES-401 Sample Written Examination Form ES-401*5 Question Worksheet Examination Outline Cross-

Reference:

Level RO/SRO Tier# 1 Group # 1 KiA # 700000 AK 2.03 Importance Rating 3.0/3.1 AND Proposed Question:

The plant is conducting a reactor startup with the following conditions:

Reactor power is currently 7%.

Main Turbine shell warming is in progress.

A heavy rain storm with lightning is occurring.

The following sequence of events occurs:

At T =0 seconds: lightning severs NMP-FITZ 115 KV Line #4 in the switchyard, rendering Line #4 out of service.

=

At T 10 seconds: a Line fault on LHH-FITZ 115KV Line #3 causes the LHH FITZ 115KV LINE 3 BKR 10022, to trip.

At T =70 seconds: the Line fault on LHH-FITZ 115KV Line #3 clears.

Based on the conditions above, which of the following correctly states the status of the 4160V buses listed below?

10100 Bus 10200 Bus Energized Energized De-energized Energized Energized De-energized De-energized De-energized Proposed Answer:

D. De-energized De-energized

Explanation (Optional):

Based on the conditions provided in the question stem, the Main Generator is not yet supplying power to the 4160V electrical buses. Therefore, the 115KV system is supplying power to all 4160V buses. When the lightning strike causes line #4 to become inoperable, the only power supply left to the 4160V buses listed is from 115KV line #3. When the 10022 breaker trips on the line fault at T = 10 seconds, per ARP 09-8-6-1 and SDLP-71 D, an auto reclosure scheme activates. If the line fault clears within 10 seconds (by T =20 seconds) the breaker will reclose and power all of the buses listed. At T = 55 seconds the 10017 disconnect will open to sectionalize the 115KV supply lines in an effort to isolate the fault. If

=

the fault clears by T 60 seconds, the 10022 breaker will reclose and power the 10200 and 10400 buses only. Since the fault does not clear until T == 70 seconds, the 10022 breaker will not reclose, therefore all of the listed 4160V buses will be de-energized (the Main Generator is not available to carry the loads either). Choice 'C' is plausible if the candidate cannot recall which buses are supplied by Line #3 with the 10017 disconnect open.

Technical Reference{s): AOP-72, OP-65, ARP 09-8-6-1, SDLP-71D Proposed references to be provided to applicants during examination: None Learning Objective: SDLP-71 D, EO 1.05.b.1 Question Source: New Question History: New Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 10CFR55.41{4,5,7,10)

Comments:

Question #21 Refueling activities are being conducted in the Reactor Building. Fuel moves between the Spent Fuel Pool (SFP) and RPV Cavity are taking place.

The following occurs:

As a recently irradiated fuel bundle is being transferred to the SFP, the Refuel Bridge Grapple breaks and the bundle falls.

Numerous fuel pins break and a gaseous bubble is reported to have escaped the Cavity water.

Reactor Building Ventilation Radiation Monitors alarm and are reading 1x1 04 cpm.

Which of the following correctly lists the areas that are designed to be protected from exceeding dose limits during this event?

A. Drywell and Refuel Floor only B. Refuel Floor and Reactor Building only C. Reactor Building and Control Room only D. Control Room and Offsite only

ES-401 Sample Written Examination Form ES~401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO/SRO Tier # 1 Group # 2 KJA# 295034 2.1.28 Importance Rating 4.1/4.1 Proposed Question:

Refueling activities are being conducted in the Reactor Building. Fuel moves between the Spent Fuel Pool (SFP) and RPV Cavity are taking place.

The following occurs:

As a recently irradiated fuel bundle is being transferred to the SFP, the Refuel Bridge Grapple breaks and the bundle falls.

Numerous fuel pins break and a gaseous bubble is reported to have escaped the Cavity water.

Reactor Building Ventilation Radiation Monitors alarm and are reading 1x104 cpm.

Which of the following correctly lists the areas that are designed to be protected from exceeding dose limits during this event?

A. Drywell and Refuel Floor only B. Refuel Floor and Reactor Building only C. Reactor Building and Control Room only D. Control Room and Offsite only Proposed Answer:

D Control Room and Offsite only

Explanation: At 1x1 04 cpm, the RB Ventilation Rad Monitors will isolate RB Ventilation and start SBGT to filter radio-nuclides to prevent exceeding Control Room and Offsite dose limits.

This is the purpose \ design of the system.

Technical Reference(s): Tech Spec 3.3.6.2 Bases (RB Vent High Rad section)

Proposed references to be provided to applicants during examination: None Learning Objective: SDLP-01 B 1.02 SDLP-16A 1.02 Question Source: New Question History: New Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 10CFR55.41 (7)

Comments:

Question #22 The plant is conducting a reactor startup (RPV pressure is 550 psig).

The Rx has just been declared critical.

An inadvertent reactor scram occurs.

Concurrent with the reactor scram, the plant experiences a complete loss of UPS.

Following the reactor scram:

SRM meters indicate lowering reactor power.

AlIlRMs indicate downscale.

Based on the conditions listed above, in addition to AOP-21 "Loss of UPS," which of the following procedures should be executed?

A. AOP-1 only B. AOP-1 and EOP-2 only C. AOP-1 and EOP-3 only D. AOP-1, EOP-2, and EOP-3

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO/SRO Tier # 1 Group # 2 KIA # 295015 AK 2.03 Importance Rating 3.2/3.6 the Proposed Question:

The plant is conducting a reactor startup (RPV pressure is 550 psig).

The Rx has just been declared critical.

An inadvertent reactor scram occurs.

Concurrent with the reactor scram, the plant experiences a complete loss of UPS.

Following the reactor scram:

SRM meters indicate lowering reactor power.

All IRMs indicate downscale.

Based on the conditions listed above, in addition to AOP-21 "Loss of UPS," which of the following procedures should be executed?

A. AOP-1 only B. AOP-1 and EOP-2 only C. AOP-1 and EOP-3 only D. AOP-1, EOP-2, and EOP-3 Proposed Answer:

A. AOP-1 only

Explanation (Optional):

Based on the conditions provided in the question stem, since reactor power is low when the reactor scram occurs, an entry condition into EOP-2 on low RPV level will not occur.

Therefore, even though the full-core display (i.e. loss of RPIS full in indication for the control rods) is not available due to the loss of UPS, with the indications provided, the candidate can determine that reactor power is < 2.5%. Since EOP-2 is not entered, there is no reason to enter EOP-3 based on the conditions in the stem. Therefore, only AOP-1 needs to be executed. Refer to the Attachments in AOP-21 for the indications and actions to perform if a scram occurs concurrent with a complete loss of UPS.

Technical Reference(s): AOP-21, MIT-301.11C, MIT-301.11D, SDLP-3G Proposed references to be provided to applicahts during examination: None Learning Objective: SDLP-3G, EO 1.1 O.a; MIT -301.11 C, EO 1.02; MIT -301.11 D, EO 1.02 Question Source: New Question History: New Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 10CFR55.41 (7)

Comments:

Question #23 Given the following conditions:

The plant is in day 5 of a Refueling Outage.

The 'B' RHR loop is in a Shutdown Cooling Lineup removing heat from the RPV.

A Control Room operator notices that RPV pressure has risen to 110 psig.

Based on the above conditions, which of the following valves do you expect to be in the CLOSED position?

1. 1OMOV-15B - RHR Pump B Suct Shutdown Cooling Isol Valve

2. 10MOV RHR Shutdown Cooling Outbd Isol Valve

3. 10MOV RHR Shutdown Cooling Inbd Isol Valve

4. 10MOV-25B - RHR B LPCllnbd Inj Valve

5. 10MOV-27B - RHR B LPCI Outbd Inj Valve A. 1,2, and 3 only B. 2, 3, and 4 only C. 1,2,3, and 4 only D. 1, 2, 3, 4, and 5

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO/SRO Tier# 1 Group # 2 KIA # 295007 AK 2.06 Importance Rating 3.5 I 3.7 Proposed Question:

Given the following conditions:

The plant is in day 5 of a Refueling Outage.

The 'B' RHR loop is in a Shutdown Cooling Lineup removing heat from the RPV.

A Control Room operator notices that RPV pressure has risen to 110 psig.

Based on the above conditions, which of the following valves do you expect to be in the CLOSED position:

1. 1OMOV-15B - RHR Pump B Suct Shutdown Cooling Isol Valve

2. 10MOV RHR Shutdown Cooling Outbd Isol Valve

3. 10MOV RHR Shutdown Cooling Inbd Isol Valve

4. 1OMOV-25B - RHR B LPCI Inbd Inj Valve

5. 10MOV-27B - RHR B LPCI Outbd Inj Valve A. 1,2, and 3 onlv B. 2, 3, and 4 only C. 1,2,3, and 4 only D. 1, 2, 3, 4, and 5 Proposed Answer:

B. 2, 3, and 4 only

Explanation (Optional):

Per OP-13 and SDLP-10 for the Residual Heat Removal system, when the RHR system is in a shutdown cooling lineup, if RPV pressure rises to :=...109 psig the RHR system will isolate automatically. When the automatic isolation occurs, the following valves go to the CLOSED position: 10MOV-17, 10MOV-18, and 10MOV-25B (or 10MOV-25A if the 'A' RHR loop is the loop lined up for shutdown cooling mode). Even though 10MOV-15B and 10MOV-27B are part of the shutdown cooling flowpath for the RHR system, they do not auto isolate on the 109 psig high RPV pressure signal.

Technical Reference(s): OP-13, AOP-30, SDLp*10 Proposed references to be provided to applicants during examination: None Learning Objective: SDLP-16C, EO 1.09.i Question Source: New Question History: New Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 10CFR55.41 (7)

Comments:

Question #24 Given the following:

The plant is operating at rated power.

The Exosensors are currently out of service due to power supply failures.

A small break LOCA occurs and causes entry into EOP-2 "RPV Control" and EOP-4 "Primary Containment ControL" Per EOP-4, "if the hydrogen or oxygen monitoring system is unavailable, then the drywell and torus shall be sampled for hydrogen and oxygen".

Which of the following correctly describes the area of the plant where the necessary sample can be obtained to monitor the hydrogen and oxygen concentrations of both the drywell AND the torus?

A. Torus Room B. RWR MG Set Room C. 27CAD panel in the Relay Room D. Primary Sample Sink Panel 95SP-7

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO/SRO Tier # 1 Group # 2 KIA # 295010 AA 1.04 Importance Rating 3.1/3.0 Proposed Question:

Given the following:

The plant is operating at rated power.

The Exosensors are currently out of service due to power supply failures.

A small break LOCA occurs and causes entry into EOP-2 "RPV Control" and EOP-4 "Primary Containment ControL" Per EOP-4, "if the hydrogen or oxygen monitoring system is unavailable, then the drywell and torus shall be sampled for hydrogen and oxygen".

Which of the following correctly describes the area of the plant where the necessary sample can be obtained to monitor the hydrogen and oxygen concentrations of both the drywell AND the torus?

A. Torus Room B. RWR MG Set Room C. 27CAD panel in the Relay Room D. Primary Sample Sink Panel 95SP-7 Proposed Answer:

B. RWR MG Set Room

Explanation (Optional):

Based on the conditions provided in the question stem, with the Exosensors out of service and unavailable for restoration per EP-2 (as directed by EOP-4) the guidance contained in PSP-17 and/or CA-01.02 must be followed in order to obtain hydrogen and oxygen concentrations of the containment environs. In order to obtain a sample of both the drywell and torus hydrogen and oxygen levels, the only area in the plant that provides a sample of both containment areas is at the Post Accident Sampling Station in the RWR MG Set room per PSP-17. CA-01.02 allows sampling of both the drywell and the torus, but the samples come from either the Torus Room or at the Drywell Sampling Rack in the Reactor Building 300' elevation. The 27CAD panel in the Relay Room is the normal sample point with the Exosensors in service.

Technical Reference(s): EOP-4, PSP-17, CA-01.02, MIT-301.11E Proposed references to be provided to applicants during examination: None Learning Objective: MIT-301.11 E, EO-4.05 Question Source: New Question History: New Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 10CFR55.41 (7)

Comments:

Question #25 In accordance with EOP-5 "Secondary Containment Control," with a primary system discharging into the Reactor Building, why is EOP-2 (RPV Control) required to be entered before water level in either crescent area exceeds 18 inches?

Inserting a manual scram . ..

A provides mitigating action such that the condition does not pose an immediate threat to the health and safety of the public.

B. promptly reduces energy to decay heat levels and reduces the likelihood of requiring rapid depressurization of the RPV.

C. promptly reduces to decay heat levels the energy discharged into primary containment and reduces the driving head and flow of primary systems that are unisolated.

D. promptly places the primary system in its lowest possible energy state and reduces the driving head and flow of primary systems that are unisolated and discharging into the secondary containment.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO/SRO Tier # 1 Group # 2 KIA # 295036 EK 3.02 Importance Rating 2.8/2.8 Proposed Question:

In accordance with EOP-5 "Secondary Containment Control," with a primary system discharging into the Reactor Building, why is EOP-2 (RPV Control) required to be entered before water level in either crescent area exceeds 18 inches?

Inserting a manual scram...

A. provides mitigating action such that the condition does not pose an immediate threat to the health and safety of the public.

B. promptly reduces energy to decay heat levels and reduces the likelihood of requiring rapid depressurization of the RPV.

c. promptly reduces to decay heat levels the energy discharged into primary containment and reduces the driving head and flow of primary systems that are unisolated.

Q. promptly places the primary system in its lowest possible energy state and reduces the driving head and flow of primary systems that are unisolated and discharging into the secondary containment Proposed Answer:

B. promptly reduces energy to decay heat levels and reduces the likelihood of requiring rapid depressurization of the RPV.

Explanation (Optional):

A. Incorrect: provides mitigating action such that the condition does not pose an immediate threat to the health and safety of the public. Even though inserting a scram will mitigate the effects of a primary system discharging into the secondary containment, protecting the health and safety of the public is not basis for scramming.

B. Correct: promptly reduces energy to decay heat levels and reduces the likelihood of requiring rapid depressurization of the RPV.

C. Incorrect: promptly reduces to decay heat levels the energy discharged into primary containment and reduces the driving head and flow of primary systems that are unisolated. This is a basis for Primary Containment Control strategy, whereas this question is dealing with secondary containment.

D. Incorrect: promptly places the primary system in its lowest possible energy state and reduces the driving head and flow of primary systems that are unisolated and discharging into the secondary containment. This is the basis for performing an Emergency Depressurization of the RPV, not for scramming the reactor.

Technical Reference(s): MIT-301.11 F, EOP-5 Proposed references to be provided to applicants during examination: None Learning Objective: MIT-301.11F, EO 1.05 Question Source: Bank -INPO Exam Bank #27828 Question History: 2003 Cooper ILO Exam Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 10CFR55.41 (5)

Comments:

Parent Question From INPO Exam Bank - Cooper 20031LO Exam Why is a reactor scram required before exceeding a Maximum Safe Operating Water Level if a primary system is discharging into the Reactor building (EOP-SA step SC-12)?

Scramming the Reactor ...

a. provides mitigating action such that the condition does not pose an immediate threat to the health and safety of the public.

b. promptly reduces energy to decay heat levels and reduces the likelihood of requiring rapid depressurization of the RPV.

c. promptly reduces to decay heat levels the energy discharged into primary containment and reduces the driving head and flow of primary systems that are unisolated.

d. promptly places the primary system in its lowest possible energy state and reduces the driving head and flow of primary systems that are unisolated and discharging into the secondary containment.

Correct Answer:

b. promptly reduces energy to decay heat levels and reduces the likelihood of requiring rapid depressurization of the RPV.

Question #26 The plant is conducting a reactor startup with the following conditions:

RPV pressure is 950 psig.

'A' CRD pump is running.

The following annunciator is received:

09-5-1-49 HCRD PMP 3P-16A TRIP" Thirty (30) seconds after the annunciator is received:

CRD Charging Water Header pressure is 925 psig.

Accumulator alarm for fully withdrawn control rod 22-27 is received.

Based on the conditions above, which of the following choices correctly describes the required action?

A. Immediately insert a manual scram.

B. Fully insert and disarm control rod 22-27.

C. Take manual control and close CRD Flow Control (03FIC-301) then start the 'B' CRD pump.

D. Insert a manual scram IF an accumulator alarm on a different withdrawn control rod is received within 20 minutes.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO/SRO Tier # 1 Group # 2 KJA# 295022 G 2.1.23 Importance Rating 4.3/4.4 I I Proposed Question:

The plant is conducting a reactor startup with the following conditions:

RPV pressure is 950 psig.

'A' CRD pump is running.

The following annunciator is received:

09-5-1-49 "CRD PMP 3P-16A TRIP" Thirty (30) seconds after the annunciator is received:

CRD Charging Water Header pressure is 925 psig.

Accumulator alarm for fully withdrawn control rod 22-27 is received.

Based on the conditions above, which of the following choices correctly describes the required action?

A. Immediately insert a manual scram.

B. Fully insert and disarm control rod 22-27.

C. Take manual control and close CRD Flow Control (03FIC-301) then start the 'B' CRD pump.

D. Insert a manual scram !E an accumulator alarm on a different withdrawn control rod is received within 20 minutes.

Proposed Answer:

C. Take manual control and close CRD Flow Control (03FIC-301) and then start the

'B' CRD pump.

Explanation (Optional):

Based on the conditions provided in the question stem, the candidates must recognize that they are in an AOP-69 "CRD Pump Trouble" event. Per the overrides in AOP-69, "if RPV pressure is less than 900 psig and charging water header pressure is less than 940 psig and any accumulator alarm is received on a withdrawn control rod" a manual scram shall be inserted. Likewise, "if RPV pressure is greater than 900 psig and charging water header pressure is less than 940 psig and 20 minutes have elapsed since a second accumulator alarm is received on a withdrawn control rod" a manual scram shall also be inserted. In the conditions provided, RPV pressure is greater than 900 psig and charging water header pressure is 925 psig. There is only one accumulator alarm on a withdrawn control rod. Since a full scram condition does not exist, the immediate actions of AOP-69 require that the CRD Flow Controller be placed in Manual and closed prior to starting the 'B' CRD pump. B, candidate may incorrectly believe control rod 22-27 is inop and has to be valved out or service. Choice 'D' is incorrect, since a scram would only be required 20 minutes AFTER the second accumulator alarm is received.

Technical Reference(s): AOP-69 Proposed references to be provided to applicants during examination: None Learning Objective: LP-AOP, EO 1.03.a,b,c Question Source: New Question History: New Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 10CFR55.41 (1 0)

Comments:

Question #27 The plant is operating at power when the following occurs:

A small steam leak in the Orywell occurs

OW pressure: 1.8 psig and rising slowly

OW temp: 136'f and rising quickly EOP-4 directs "IF drywell temperature cannot be restored and maintained below 135'f THEN Operate all available drywell cooling."

This statement is attempting to preserve the integrity of which one of the following?

A. RBCLC isolation valves B. Primary Containment C. RPV level instrumentation D. Hydrogen \ Oxygen sampling capability

ES-401 Sample Written Examination Form ES~401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO/SRO Tier # 1 Group # 2 KIA # 295012 AK3.01 Importance Rating 3.5/3.6 Proposed Question:

The plant is operating at power when the following occurs:

A small steam leak in the Drywell occurs

DW pressure: 1.8 psig and rising slowly

DW temp: 136'f and rising quickly EOP-4 directs "IF drywell temperature cannot be restored and maintained below 135'f THEN Operate all available drywell cooling."

This statement is attempting to preserve the integrity of which one of the following?

A. RBCLC isolation valves B. Primary Containment C. RPV level instrumentation D. Hydrogen \ Oxygen sampling capability Proposed Answer:

B Primary Containment

Explanation: TS Bases 3.6.1.5 states: Analyses assume an initial average drywell air temp of 135F. This limitation ensures that the safety analysis remains valid by maintaining the expected initial conditions and ensures that the peak LOCA drywell temp and pressure do not exceed the drywell design pressure of 56 psig coincident with a design temp or 309F.

Exceeding these design limitations may result in the degradation of the primary containment structure under accident loads. Equipment inside primary containment required to mitigate the effects of a DBA is designed to operate and be capable of operating under environmental conditions expected for the spectrum of break sizes.

Technical Reference(s): Tech Spec 3.6.1.5 + Bases Proposed references to be provided to applicants during examination: None Learning Objective: SDLP-16A 1.07 Question Source: New Question History: New Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 10CFR55.41(5)

Comments:

Question #28 The plant has experienced a Design Bases Loss of Coolant Accident on the "An Recirc Loop.

The following conditions exist:

OW pressure is 20 psig and rising

RPV pressure is 200 psig and lowering

"C" RHR Pump is the only ECCS pump running

Annunciator 09-3-1-34 (RWR INJ VLV PERM) is in alarm

02MOV-53A (RWR PMP A DISCH) is open

02MOV-53B (RWR PMP B DISCH) is closed Based on the above conditions, which one of the following correctly states the status of LPCI and the required Operator action?

LPCI status Operator action A. Injecting into the RPV Verify Open 02MOV-53A B. NOT injecting into the RPV Open 02MOV-538 C. Injecting into the RPV Verify Closed 02MOV-538 D. NOT injecting into the RPV Close 02MOV-53A

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO/SRO Tier # 2 Group # 1 KIA # 203000 A2.1 0 Importance Rating 3.3/3.5 Proposed Question:

The plant has experienced a Design Bases Loss of Coolant Accident on the "A" Recirc Loop.

The following conditions exist:

DW pressure is 20 psig and rising

RPV pressure is 200 psig and lowering

"c" RHR Pump is the only ECCS pump running

Annunciator 09-3-1-34 (RWR II\JJ VLV PERM) is in alarm

02MOV-53A (RWR PMP A DISCH) is open

02MOV-53B (RWR PMP B DISCH) is closed Based on the above conditions, which one of the following correctly states the status of LPCI and the required Operator action?

LPCI status Operator action A. Injecting into the RPV Verify Open 02MOV-53A B. NOT injecting into the RPV Open 02MOV-53B C. Injecting into the RPV Verify Closed 02MOV-53B D. NOT injecting into the RPV Close 02MOV-53A Proposed Answer:

D. NOT injecting into the RPV Close 02MOV-53A

Explanation: 2.7 psig or <59.5" is a LPCI initiation. When RPV pressure is < 345#, RWR discharge valve 02MOV-53A should auto close to isolate a break in the Recirc loop and allow RHR to recover RPV level. With Annunciator 09-3-1-34 (RWR INJ VLV PERM) in alarm, means that the pressure transmitter \ close initiation signal to 02MOV-53A is active and the valve should be closed. 80th the ARP and OP-*13A direct the Operator to verify \ ensure the Recirc discharge valves are closed. This is required to insure LPCI does not inject out the break. The "c" RHR pump injection valve has automatically opened with the LPCI initiation and low reactor pressure. It is injecting into the loop that has the break. With the RWR 53A valve open, LPCI is being diverted into the break and the lowest pressure region (Le. the 20 psig drywell rather than the 200 psig reactor).

A. Incorrect - Chosen if the candidate believes LPCI injects downstream of the valve and associates the RWR isolation signal with the RWR suction valve rather than the discharge valve ..

8. incorrect - This would be chosen for the same reasons as choice "An with the exception n

that the candidate believes "C RHR injects into the "8" RWR Loop.

C. incorrect - This would be correct if "c" RHR injected into the "8" RWR loop.

D. Correct see above.

Technical Reference(s): OP-13A, ARP 09-3-1-34 Proposed references to be provided to applicants during examination: None Learning Objective: SDLP-10 1.10,1.14 SDLP-02H 1.10,1.14 Question Source: New Question History: New Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 10CFR55.41 (5)

Comments:

Question #29 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months ago, the Plant entered into a Forced Outage.

The following conditions exist:

RHR Loop 'B' is in a Shutdown Cooling lineup

Recirc Pump 'A' is running

RPV level is 217 inches The following occurs:

10300 and 10500 Busses de-energize due to a lockout condition Which one of the following choices is correct concerning Reactor Coolant temperature indication and the reason why?

Coolant temperature indication Reason why A. Could become invalid Decay heat is insufficient B. Could become invalid RPV level is insufficient C. Will remain valid RWR flow is sufficient O. Will remain valid SOC flow is sufficient

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO/SRO Tier # 2 Group # 1 KJA# 205000 K3.03 Importance Rating 3.8/3.9 Proposed Question:

36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months ago, the Plant entered into a Forced Outage.

The following conditions exist:

RHR Loop 'B' is in a Shutdown Cooling lineup

Recirc Pump 'A' is running

RPV level is 217 inches The following occurs:

10300 and 10500 Busses de-energize due to a lockout condition Which one of the following choices is correct concerning Reactor Coolant temperature indication and the reason why?

Coolant temperature indication Reason why A. Could become invalid Decay heat is insufficient B. Could become invalid RPV level is insufficient C. Will remain valid RWR flow is sufficient D. Will remain valid SOC flow is sufficient Proposed Answer:

B Could become invalid, RPV level is insufficient

Explanation: 10300 de-energizing results in RWR 'A' tripping. Therefore no RWR pumps running. 10500 de-energizing results in a loss of RPS 'A' which results in a loss of SDC. With no forced circulation in RPV, OP-13D states: CAUTION If RPV water level is less than 234.5 inches with no forced core recirculation, reactor coolant temperature indications could be invalid due to insufficient natural circulation. AOP-30 states: RPV WATER LEVEL CONTROL Attempt to maintain RPV water level as follows:

A. No RWR pump running: BETWEEN 234.5 and 270 inches.

B. RWR pump running: BETWEEN 200 and 270 inches.

Candidate may incorrectly believe a Rx shutdown only 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months ago would have enough decay heat to promote natural circulation.

Technical Reference( s): OP-13D, AOP-30 Proposed references to be provided to applicants during examination: None Learning Objective: SDLP-10 1.09, 1.15 Question Source: New Question History: New Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 10CFR55.41 (7)

Comments:

Question #30 A loss of coolant accident (LOCA) has occurred concurrent with a complete loss of off-site power.

The reactor has successfully scrammed and the following conditions exist:

Operators are performing actions per EOP-2 "RPV Control".

Operators are performing actions per AOP-72 "115KV Grid Loss, Instability, or Degradation" .

RPV Water Level is 57 inches and rising slowly.

RPV Pressure is 490 psig.

Which of the following systems should be used to restore and maintain RPV water level between 177 and 222.5 inches?

A. Condensate B. Feedwater C. LPCI D. HPCI

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO/SRO Tier# 2 Group # 1 KJA# 206000 G 2.4.6 Importance Rating 3.7/4.7 ofEOP /4fi Proposed Question:

A loss of coolant accident (LOCA) has occurred concurrent with a complete loss of off-site power.

The reactor has successfully scrammed and thH following conditions exist:

Operators are performing actions per EOP-2 "RPV Control".

Operators are performing actions per AOP-72 "115KV Grid Loss, Instability, or Degradation".

RPV Water Level is 57 inches and rising slowly.

RPV Pressure is 490 psig.

Which of the following systems should be used to restore and maintain RPV water level between 177 and 222.5 inches?

A. Condensate B. Feedwater C. LPCI D. HPCI Proposed Answer:

D. HPCI

Explanation (Optional):

Based on the conditions provided in the question stem, since there has been a complete loss of off-site power, the 10300 and 10400 buses wi" be de-energized. Since the reactor has scrammed, the Main Turbine will have automatically tripped, de-energizing the 10700 bus.

This means that there is no power available to the condensate pumps. RPV pressure is 490 psig, therefore the MSIVs should be shut (both by the 500 psig benchmark and per the AOP-1 actions for no running Circulating Water pumps). If the MSIVs are shut, the Feedwater system is not available for injection. Additionally, RPV pressure is still greater than the LPCI injection permissive pressure of 450 psig. Therefore, the only system listed that is available for injection to restore level is HPCI.

Technical Reference(s): EOP-2, MIT-301.11C Proposed references to be provided to applicants during examination: None Learning Objective: MIT-301.11C, EO 1.07 Question Source: New Question History: New Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 10CFR55.41 Comments:

Question #31 Given the following:

The plant has experienced an Anticipated Transient Without Scram (ATWS).

The ATC has just commenced injection of Standby Liquid Control with the 'A' SLC pump.

RPV water level is 100 inches and steady.

RPV pressure is 1000 psig and slowly rising.

Which of the following valves would you expect to be in the CLOSED position?

1. CLN UP INBD SUCT ISOL VALVE (12MOV-15)

2. CLN UP OUTBD SUCT ISOL VALVE (12MOV-18)

3. CLN UP RETURN ISOL VALVE (12MOV-69)

4. DRYWELL EQUIPMENT DRAIN (20MOV-94)

A. 1 and 3 only B. 2 and 3 only C. 1, 2, and 3 only D. 1,2,3,and4

ES-401 Sample Written Examination Form ES*401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO/SRO Tier # 2 Group # 1 KIA # 211000 K 1.05 Importance Rating 3.4/3.6 Proposed Question:

Given the following:

The plant has experienced an Anticipated Transient Without Scram (ATWS).

The ATC has just commenced injection of Standby Liquid Control with the 'A' SLC pump.

RPV water level is 100 inches and steady.

RPV pressure is 1000 psig and slowly rising.

Which of the following valves would you expect to be in the CLOSED position?

1. CLN UP INBD SUCT ISOL VALVE (12MOV-15)

2. CLN UP OUTBD SUCT !SOL VALVE (12MOV-18)

3. CLN UP RETURN ISOL VALVE (12MOV-69)

4. DRYWELL EQUIPMENT DRAIN (20MOV-94)

A. 1 and 3 only B. 2 and 3 only C. 1,2, and 3 only D. 1, 2, 3, and 4 Proposed Answer:

D. 1, 2, 3, and 4

Explanation (Optional):

During an ATWS, operators are expected to carry out the actions of EOP-3. The power control leg of EOP-3 mandates RPV injection using SLC. There is an interlock between SLC and the RWCU system that causes 12MOV-18 (RWCU suction isolation valve) and 12MOV-69 (RWCU return isolation valve) to auto close when an SLC pump is started from the 09-3 panel. Auto closure of these valves prevents the sodium penta borate from being removed from the RCS water makeup by the RWCU system filters. Based on the conditions provided in the question stem, the candidates must also realize that a Group 2 isolation would have also occurred based on RPV water level such that 12MOV-15 (RWCU inboard suction isolation valve) and 20MOV-94 (Drywell Equipment Isolation valve) would also be in the CLOSED position.

Technical Reference(s): OP-17, AOP-15, SDLP-11 Proposed references to be provided to applicants during examination: None Learning Objective: SDLP-11, EO-1.09.a Question Source: New Question History: New Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 10CFR55.41 (2 through 9)

Comments:

Question #32 The unit was operating at near rated power when a transient occurred resulting in the following timeline:

Time: 00:00:00, "C" EOG is tagged out for maintenance

Time: 00:05:00, Orywell pressure increased to 3.5 psig

Time: 00:06:00, "8" & "0" EOGs are not running and can not be started

Time: 00: 10:00, the feed breaker between 8uses 10300 and 10500 inadvertently trips At time 00:12:00, which of the following choices indicates which Core Spray pumps(s) (if any) will be operating:

A None B. "A" Pump ONLY C. "8" Pump ONLY O. 80TH "A" and "8" Pumps

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level ROISRO Tier # 2 Group # 1 KIA # 209001 A 4.01 Importance Rating 3.8/3.6 Iii Proposed Question: #32 The unit was operating at near rated power when a transient occurred resulting in the following timeline:

Time: 00:00:00, "C" EDG is tagged out for maintenance

Time: 00:05:00, Drywell pressure increased to 3.5 psig

Time: 00:06:00, "B" & "D" EDGs are not running and can not be started

Time: 00: 10:00, the feed breaker between Buses 10300 and 10500 inadvertently trips At time 00:12:00, which of the following choices indicates which Core Spray pumps(s) (if any) will be operating:

A. None B. "A" Pump ONLY C. "B" Pump ONLY D. BOTH "/\' and "B" Pumps Proposed Answer:

D. BOTH "A" and "B" Pumps

Explanation (Optional):

Both Core Spray pumps started on a LOCA signal with Drywell pressure at 2.7 psig. The loss of the feeder breaker to 10500 caused an undervoltage trip and load shed of the bus. The EDGs start the ECCS pumps and sequence onto the bus. The "C" EDG powers the 10500 bus as does the "A" EDG. The loss of "C" EDG does not prevent the "A" EDG from powering the bus. The loads sequence on as before with the exception that one RHR pump will no longer sequence on. The time for the diesel to start (10 secs) and for the CS pump to sequence on (-12 secs) is within the band of the 2 minutes between the breaker trip and the CS pump starting. The other CS pump has been steadily running from time 00:05:00.

A. Incorrect: This would be selected if the candidate thought the "B" and "D" EDGs powered the 10500 and the "C" EDG would not power CS (as opposed to RHR).

B. Incorrect: This would be selected if the candidate thought the "B" and "D" EDGs powered the 10500 and that "1\' CS was powered from 10600.

C. Incorrect: This would be selected if the candidate thought the "B" and "D" EDGs powered the 10500 and correctly remember "B" CS is powered from 10600.

D. Correct Technical Reference(s): AOP-18, Proposed references to be provided to applicants during examination:

Learning Objective: SDLP-93 1.04a, 1.06c Question Source: Dresden 2010 NRC Exam Question History:

Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 10CFR55.41 (7)

Comments:

Question #33 The plant is conducting a reactor startup per OP-65 with the following conditions:

The Mode switch is in Startup.

Reactor power is 6%.

The following events occur:

The 'A' Main Steam Line unexpectedly isolates (Outboard MSIV goes shut).

Rx Pressure indicator 06PI-90A on the 09-5 Panel indicates upscale

02-3PT-55A (Reactor Vessel Press Xmitter-Scram and SOC Logic) has failed high.

Based on these conditions, which one of the following correctly states the status of the reactor?

A. No full or half scram has occurred.

B. Y2 scram RPS 'N only.

C. Y2 scram RPS 'B' only.

O. Full scram has occurred.

ES*401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO/SRO Tier # 2 Group # 1 KJA# 212000 K 6.02 Importance Rating 3.7/3.9 of Proposed Question:

The plant is conducting a reactor startup per OP-65 with the following conditions:

The Mode switch is in Startup.

Reactor power is 6%.

The following events occur:

The 'A' Main Steam Line unexpectedly isolates (Outboard MSIV goes shut).

Rx Pressure indicator 06PI-90A on the 09-5 Panel indicates upscale

02-3PT-55A (Reactor Vessel Press Xmitter-Scram and SDC Logic) has failed high.

Based on these conditions, which of the following correctly states the status of the reactor?

A. No full or half scram has occurred.

B. Y2 scram RPS 'A' only.

C. Y2 scram RPS 'B' only.

D. Full scram has occurred.

Proposed Answer:

B. Y2 scram RPS 'A' only.

Explanation (Optional):

Based on the conditions in the question stem, with the Mode Selector Switch NOT in the "RUN" position. The isolation of the 'A' Main Steam Line will have no effect on RPS System B since the contacts for the applicable MSIVs not full open in the RPS 'B' logic will be bypassed.

Therefore, with the upscale failure of 02-3PT -55A, only a scram of 'A' RPS will occur.

Technical Reference(s): SDLP-05, GE Drawings 1.60-17, 1.67-97 through 100 Proposed references to be provided to applicants during examination: None Learning Objective: SDLP-05, EO 1.10.c Question Source: New Question History: New Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 10CFR55.41 (7)

Comments:

Question #34 Given the following:

A reactor startup is in progress.

IRM 'A' is bypassed.

The reactor has been declared critical and the ATC has established a 150 second period.

All IRMs are on range 5.

The following alarm is received in the Control Room:

Annunciator 09-5-2-52 "IRM TRIP SYSTEM A INOP OR UPSCALE" It has been determined that the cause of the annunciator was a failure of the power supply to IRM'E'.

Based on these conditions, which of the following correctly describes the status:

(1) of RPS?

(2) of Reactor Manual Control System?

ill A. No scram No Control Rod Block B. No scram Control Rod Block C. % scram RPS 'A' No Control Rod Block D. % scram RPS 'A' Control Rod Block

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO/SRO Tier # 2 Group # 1 KIA # 215003 K 3.04 Importance Rating 3.6/3.6 Proposed Question:

Given the following:

A reactor startup is in progress.

IRM 'A' is bypassed.

The reactor has been declared critical and the ATC has established a 150 second period.

All IRMs are on range 5.

The following alarm is received in the Control Room:

Annunciator 09-5-2-52 "IRM TRIP SYSTEM A INOP OR UPSCALE" It has been determined that the cause of the annunciator was a failure of the power supply to IRM'E'.

Based on these conditions, which of the following correctly describes the status:

(1) of RPS?

(2) of Reactor Manual Control System?

ill ru A. No scram No Control Rod Block B. No scram Control Rod Block C. Y2 scram RPS 'A' No Control Rod Block D. !h scram RPS 'A' Control Rod Block

Proposed Answer:

D. 'A' RPS scram Control Rod Block Explanation (Optional):

Per OP-16, IRM trips are dependent upon reactor mode switch position. When the reactor mode switch is in a position other than RUN (as in the question stem), an IRM upscale (High) or inoperative trip actuates a Neutron Monitoring System trip of the associated RPS trip channel (in this case, RPS 'A' will scram). A downscale indication on an IRM indicates that the instrument has failed or is not sensitive enough. In either case, the instrument will not respond to changes in control rod motion and control rod withdrawal is prevented.

Technical Reference(s): OP-16, ARP 09-5**2-2, ARP 09-5-2-42, ARP 09-5-2-52 Proposed references to be provided to applicants during examination: None Learning Objective: SDLP-07B, EO 1.09.a,b Question Source: New Question History: New Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 10CFR55.41 (7)

Comments:

Question #35 The plant is conducting a plant startup per OP-65.

The following conditions exist:

The reactor has just been declared critical.

The ATC has just completed obtaining critical data and is withdrawing control rods to establish conditions for RPV heatup.

While establishing an allowable heatup rate, L 16 de-energizes.

With respect to SRMs, which of the following choices below correctly describes the effect of the loss of L 16?

A. Loss of SRM 'B' and '0' detector drive capability B. Loss of SRM 'B' and '0' trip auxiliary units C. Loss of indication from the SRM recorder at the 09-5 panel O. Loss of indication from the SRM count and period meters

ES*401 Sample Written Examination Question Worksheet Examination Outline Cross-

Reference:

Level RO/SRO Tier # 2 Group # 1 KIA # 215004 K 2.01 Importance Rating 2.6/2.8 Proposed Question:

The plant is conducting a plant startup per OP-65.

The following conditions exist:

The reactor has just been declared critical.

The ATC has just completed obtaining critical data and is withdrawing control rods to establish conditions for RPV heatup.

While establishing an allowable heatup rate, L 16 de-energizes.

With respect to SRMs, which of the following choices below correctly describes the effect of the loss of L 167 A. Loss of SRM 'B' and '0' detector drive capability B. Loss of SRM 'B' and '0' trip auxiliary units C. Loss of indication from the SRM recorder at the 09-5 panel O. Loss of indication from the SRM count and period meters Proposed Answer:

A. Loss of SRM detector drive capability

Explanation (Optional):

Based on the conditions provided in the question stem, the loss of L 16 results in a loss of 71ACNMS-B will cause the detector drive motors for SRMs Band D to lose power. The other distractors are powered by AC power supplies either from 24DC or from the Uninterruptible Power Supply (UPS). Refer to OP-46B for the loads supplied.

Technical Reference(s): OP-16, OP-46B Proposed references to be provided to applicants during examination: None Learning Objective: SDLP-07B, EO 1.04.a, b, c, d Question Source: New Question History: New Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 10CFR55.41(7)

Comments:

Question #36 During a reactor shutdown, OP-65 requires (1) to be performed.

Additionally, OP-65 requires EN-OP-116, Infrequently Performed Tests or Evolutions to apply to (2) ill A SRM-IRM Overlap Verification Only Rapid Power Reductions lAW RAP ST-5G 7.3.16 B. SRM-IRM Overlap Verification All reactor startups and planned ST-5G shutdowns.

C. IRM-APRM Instrument Range Overlap Only Rapid Power Reductions lAW RAP Check, ST-5C 7.3.16 D. IRM-APRM Instrument Range Overlap All reactor startups and planned Check, ST-5C shutdowns.

ES-401 Sample Written Examination Form ES-401*5 Question Worksheet Examination Outline Cross-

Reference:

Level RO/SRO Tier # 2 Group # 1 KIA # 215005 K 1.02 Importance Rating 3.7/3.7 Proposed Question #36:

During a reactor shutdown, OP-65 requires (1) to be performed.

Additionally, OP-65 requires EN-OP-116, Infrequently Performed Tests or Evolutions to apply to (2) ill A. SRM-IRM Overlap Verification Only Rapid Power Reductions lAW RAP ST-5G 7.3.16 B. SRM-IRM Overlap Verification All reactor startups and planned ST-5G shutdowns.

C. IRM-APRM Instrument Range Overlap Only Rapid Power Reductions lAW RAP Check, ST -5C 7.3.16 D. IRM-APRM Instrument Range Overlap All reactor startups and planned Check, ST-5C shutdowns.

Proposed Answer:

D. fRM-APRM Instrument Range Overlap All reactor startups and planned Check, ST-5C shutdowns.

Explanation (Optional):

A. INCORRECT: SRM-IRM overlap is performed on startups, not shutdowns. OP-65 references RAP 7.3.16 and would be correct if it was being used in conjunction with a "planned" shutdown (and RAP 7.3.16 is used for planned shutdowns). But a Rapid Power reduction is not planned and thus doesn't apply.

B. INCORRECT: SRM-IRM overlap is performed on startups, not shutdowns C. INCORRECT: OP-65 references RAP 7.3.16 and would be correct if it was being used in conjunction with a "planned" shutdown (and RAP 7.3.16 is used for planned shutdowns).

But a Rapid Power reduction is not planned and thus doesn't apply.

D. CORRECT:

Technical Reference(s): OP-65, OP-16 Proposed references to be provided to applicants during examination: None Learning Objective: SDLP-07B, EO 1.13.b, c; LPOP-65B EO 1.06 Question Source: New Question History: New Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 10CFR55.41 (2,6,7)

Comments:

Question #37 The plant is currently carrying out actions per EOP-2 "RPV Control" and EOP-3 "Failure to Scram."

The following conditions exist:

RCIC is injecting into the RPV at 400 gpm.

SLC 'A' is injecting into the RPV.

Both CRD pumps are running.

CST level is 57 inches.

Torus level is 14.10 feet.

Torus temperature is 195'F and rising slowly.

RCIC flow indication at the 09-4 panel begins to oscillate unexpectedly.

A field operator reports from the RCIC room that the RCIC pump is extremely noisy and that it "sounds like the pump is pumping rocks."

Which of the following choices correctly describes the reason for the current status of the RCIC pump?

A. Pump vortexing due to the current Torus water level B. Pump vortexing due to the current CST water level C. Pump cavitation due to the current Torus temperature D. Pump cavitation due to the current CST water level

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO/SRO Tier# 2 Group # 1 KIA # 217000 K 5.01 Importance Rating 2.6/2.6 K 5AJ1 indicaUons Proposed Question:

The plant is currently carrying out actions per EOP-2 "RPV Control" and EOP-3 "Failure to Scram."

The following conditions exist:

RCIC is injecting into the RPV at 400 gpm.

SLC 'A' is injecting into the RPV.

Both CRD pumps are running.

Torus level is 14.10 feet.

Torus temperature is 195'F and rising slowly.

RCIC flow indication at the 09-4 panel begins to oscillate unexpectedly.

A field operator reports from the RCIC room that the RCIC pump is extremely noisy and that it "sounds like the pump is pumping rocks."

Which of the following choices correctly describes the reason for the current status of the RCIC pump?

A. Pump vortexing due to the current Torus water level B. Pump vortexing due to the current CST water level C. Pump cavitation due to the current Torus temperature D. Pump cavitation due to the current CST water level Proposed Answer:

C. Pump cavitation due to the current Torus temperature

Explanation (Optional):

Based on the conditions provided in the question stem, CST level is below the low level setpoint which causes RCIC suction to automatically swap to the torus. The flow oscillations and field observations of the RCIC pump are indications of pump cavitation. Per the precautions of OP-19 for RCIC operation, operation of RCIC with suction from the torus and torus water temperature> 190'F could cause pump ca vitation. Choices Band D are plausible if the candidate fails to recognize that RCIC suction is aligned to the torus. Pump vortexing is only a concern at low torus water levels of less than 5.7 feet. Thus, the only response that is valid based on the given conditions is "C".

Technical Reference(s): OP-19 Proposed references to be provided to applicants during examination: None Learning Objective: SDLP-13, EO 1.0'7.e, 1.1 OJ Question Source: New Question History: New Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 10CFR55.41 (5)

Comments:

Question #38 The Plant was operating at power with RHR Pumps 'B' and 'D' tagged out of service when a coolant leak in the Drywell developed.

The following conditions exist:

RPV level 59 inches and lowering slowly

Core Spray Pump 'A' failed to start

10600 Bus is de-energized

RHR Pump 'A' is running with a discharge pressure of 195 psig

No EOP actions have been taken yet 90 seconds later, RHR Pump 'A' shows signs of cavitation.

RHR Pump 'A' discharge pressure is oscillating between 85-115 psig Based on the above conditions, which one of the following correctly states the response of the Automatic Depressurization System (ADS) Timer and the ADS valves?

ADS Timer ADS Valves A. Continues to time out Will not open at time 120 seconds B. Resets to zero Will open at time 120 seconds C. Resets to zero Will not open at time 120 seconds D. Continues to time out Will open at time 120 seconds

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Level RO/SRO Tier # 2 Group # 1 VJA# 218000 K6.01 Importance Rating 3.9/4.1 Proposed Question: 38 The Plant was operating at power with RHR Pumps 'B' and '0' tagged out of service when a coolant leak in the Drywell developed.

The following conditions exist:

RPV level 59 inches and lowering slowly

Core Spray Pump 'A' failed to start

10600 Bus is de-energized

RHR Pump 'A' is running with a discharge pressure of 195 psig

No EOP actions have been taken yet 90 seconds later, RHR Pump 'A' shows signs of cavitation.

RHR Pump 'A' discharge pressure is oscillating between 85-115 psig Based on the above conditions, which one of the following correctly states the response of the Automatic Depressurization System (ADS) Timer and the ADS valves?

ADS Timer ADS Valves A. Continues to time out Will !!.Q.! open at time 120 seconds B. Resets to zero Will open at time 120 seconds C. Resets to zero Will not open at time 120 seconds D. Continues to time out Will open at time 120 seconds Proposed Answer: A, Continues to time out Will not open at time 120 seconds

Explanation: Adjustable timer (S134 seconds) a) Timer starts with signal from low-low-low water level and the low level confirmation. b) Will reset on: (1) Loss of power, (2) Placing associated normal/override switch in "override". (3) Level increasing to;:: 59.5" or ;::177" within 134 seconds.

Low pressure ECCS pump 1) ADS logic requires either an RHR or Core Spray Pump running with a discharge pressure of ;::125 psig or ;::100 psig respectively for initiation permissive.

Assures that the relief valves will not open unless there is a method of putting water back into the reactor vessel. Stopping all pumps will allow ADS valves to shut with an initiation signal in. It does not reset the timer, and if a pump is started, the valves will open provided RPV level is

< 59.5".

A. correct-see above explanation B. incorrect- see above.

C. incorrect- see above.

D. incorrect- see above.

Technical Reference(s): OP-68 Proposed references to be provided to applicants during examination: none Learning Objective: SDLP-02J 1.10 Question Source: Bank #

Modified Bank # _ _ _ (Note changes or attach parent)

New X Question History: Last NRC Exam (Optional: Questions validated at the facility since 10195 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 ..ill 55.43 Comments:

Question #39 The Plant is operating at power when a medium sized LOCA occurs in the Drywell.

The Reactor is manually scrammed and the following conditions exist:

Control Rods: All not full in

RPV pressure: 705 psig lowering slowly

RPV level: Zero inches lowering quickly

HPCI: Running injecting 4250 gpm

RCIC: Running injecting 400 gpm

RHR: Running

Core Spray: Running Which one of the following correctly states which Plant System is designed to assist in protecting the fuel cladding from overheating?

A. Control Rods B. RCIC C. ADS D. SLC

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO/SRO Tier # 2 Group # 1 KJA# 218000 2.1.27 Importance Rating 3.9/4.0 ADS:

Proposed Question: 39 The Plant is operating at power when a medium sized LOCA occurs in the Drywell.

The Reactor is manually scrammed and the following conditions exist:

Control Rods: All not full in

RPV pressure: 705 psig lowering slowly

RPV level: Zero inches lowering quickly

HPCI: Running injecting 4250 gpm

RCIC: Running injecting 400 gpm

RHR: Running

Core Spray: Running Which one of the following correctly states which Plant System is designed to assist in protecting the fuel cladding from overheating?

A. Control Rods B. RCIC C. ADS D. SLC Proposed Answer: C, ADS

Explanation:

Purpose:

The Automatic Depressurization System (ADS) is provided to reduce reactor coolant system pressure during a small or intermediate break accident when the HPCI System either fails to operate or cannot provide water fast enough to maintain reactor water level. It is necessary in these situations to reduce reactor pressure so that the LPCI mode of RHR and/or Core Spray can restore reactor water level to protect the fuel cladding barrier from failure due to overheating.

A. incorrect- see above. Control Rods will lower the heat output of a fuel rod but not prevent overheating if a low RPV level exists B. incorrect- see above. Even if RPV pressure was lower that the malfunctioning RCIC discharge pressure, RCIC is not credited\desjgned for Adequate Core Cooling C. correct-see above explanation D. incorrect- see above. SLC will shutdown the Rx but cannot prevent fuel overheating with low level Technical Reference(s): OP-68 Proposed references to be provided to applicants during examination: none Learning Objective: _SDLP-02J 1.01 Question Source: Bank #

Modified Bank # _--,--_ (Note changes or attach parent)

New X Question History: Last NRC Exam (Optional: Questions validated at the facility since 10195 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 Jl1 55.43 Comments:

Question #40 The Plant has just completed a normal Reactor shutdown with the following conditions present:

A cooldown via the Turbine Bypass Valves (TBVs) is in progress.

The Bypass Opening Jack has a 20% demand signal.

RPV pressure is 560 psig

RPV level is 180 inches being maintained by Condensate.

The following occurs:

A spurious invalid Group I isolation "comes in" and clears.

Which one of the following correctly states the response of RPV pressure and the required manipulation of PCIS VLV RESET switches 16A-S32 (33) to correct the above condition?

Reactor pressure wilL .. PCIS VLV RESET: 16A-S32 and S33 A. continue to lower thru the TBVs. Place either 16A-S32 or 16A-S33 in Reset B. rise until the MSIVs are re-opened. Place either 16A-S32 ill 16A-S33 in Reset C. continue to lower thru the TBVs. Place both 16A-S32 and 16A-S33 in Reset D. rise until the MSIVs are re-opened. Place both 16A-S32 and 16A-S33 in Reset

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO/SRO Tier # 2 Group # 1 K/A# 223002 K3.07 Importance Rating 3.7/3.8 Proposed Question: 40 The Plant has just completed a normal Reactor shutdown with the following conditions present:

A cooldown via the Turbine Bypass Valves (TBVs) is in progress.

The Bypass Opening Jack has a 20% demand signal.

RPV pressure is 560 psig.

RPV level is 180 inches being maintained by Condensate.

The following occurs:

A spurious invalid Group I isolation "comes in" and clears.

Which one of the following correctly states the response of RPV pressure and the required manipulation of PCIS VLV RESET switches 16A-S32 (33) to correct the above condition?

Reactor pressure wilL .. PCIS VLV RESET: 16A-S32 and S33 A. continue to lower thru the TBVs. Place either 16A-S32 or 16A-S33 in Reset B. rise until the MSIVs are re-opened. Place either 16A-S32 or 16A-S33 in Reset C. continue to lower thru the TBVs. Place both 16A-S32 and 16A-S33 in Reset D. rise until the MSIVs are re-opened. Place both 16A-S32 and 16A-S33 in Reset Proposed Answer: DJ rise until the MSIVs are re-opened. Place both 16A-S32 and 16A-S33 in Reset

Explanation: A Group I isolation, even momentary \ spurious, will lock in and must be reset by an Operator. To reset the isolation, both S32 and 33 must be simultaneously placed in Reset.

When a GP I isolation is processed, MSIVs close, therefore Rx pressure will rise.

A. incorrect- see above. MSIVs shut, therElfore no steam to go thru TBVs B. incorrect- see above. Both switches must be manipulated C. incorrect- see above. MSIVs shut, therefore no steam to go thru TBVs D. correct-see above explanation Technical Reference(s): AOP-15 Proposed references to be provided to applicants during examination: none Learning Objective: _SDLP-16C 1.09, 14, 15 Question Source: Bank#

Modified Bank # _ _ _ (Note changes or attach parent)

New Question History: Last NRC Exam (Optional: Questions validated at the facility since 10195 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis 10 CFR Part 55 Content: 55.41 m 55.43 Comments:

Question #41 The Plant is operating at 100% power when the following occurred:

The "8' 125 VDC distribution system de-energizes (complete loss)

To control Rx pressure (1 ) ADS/Safety Relief Valves could be operated from local SRV Panel 02ADS-071 and (2) ADS/Safety Relief Valves could be operated from Control Room Panel 09-4.

(1 ) (2)

A. NO NO B. NO All C. All NO D. All All

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO/SRO Tier # 2 Group # 1 KJA# 239002 K3.01 Importance Rating 3.9/4.0 Proposed Question: 41 The Plant is operating at 100% power when Ule following occurred:

The'S' 125 VDC distribution system de-energizes (complete loss)

To control Rx pressure (1 } ADS/Safety Relief Valves could be operated from local SRV Panel 02ADS-071 and (2) ADS/Safety Relief Valves could be operated from Control Room Panel 09-4.

(1) (2)

A. NO NO B. NO All C. All NO D. All All Proposed Answer: 8: NO, All

Explanation: The solenoids associated with Control Room SRVs operation are normally energized by 125 VDC Bus A and have 12ei VDC Bus B. The solenoids associated with local operation are normally power by 125 VDC Bus B and have no backup. The loss of 125 VDC Bus B does not affect control room operation but will completely disable local operation.

A. correct-see above explanation B. incorrect- see above.

C. incorrect- see above.

D. incorrect- see above.

Technical Reference(s): OP-68 Proposed references to be provided to applicants during examination: none Learning Objective: _SDLP-02J 1.04, 10 Question Source: Bank #

Modified Bank # _ _ _ (Note changes or attach parent)

New X Question History: Last NRC Exam (Optional: Questions validated at the facility since 10195 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 m 55.43 Comments:

Question #42 The Plant has been operating at 100% power for 124 days when the following occurs:

09-4-2-6 SRV SONIC MON ALARM HI alarms

09-4-1-16 SRV LEAKING alarms Which one of the following sets correctly describes the Plant response and the correct Operator action to mitigate the event?

Plant response Operator action A. Steam flow lowers Take manual control and lower Feedwater flow B. Main Generator load lowers Trip the Main Turbine C. Reactor power rises Lower Recirculation flow D. Reactor pressure fluctuates Place backup EHC pressure regulator in service

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level ROISRO Tier # 2 Group # 1 KIA # 239002 A2.03 Importance Rating 4.1/4.2 Proposed Question: 42 The Plant has been operating at 100% power for 124 days when the following occurs:

09-4-2-6 SRV SONIC MON ALARM HI alarms

09-4-1-16 SRV LEAKING alarms Which one of the following sets correctly describes the Plant response and the correct Operator action to mitigate the event?

Plant response Operator action A. Steam flow lowers Take manual control and lower Feedwater flow B. Main Generator load lowers Trip the Main Turbine C. Reactor power rises Lower Recirculation flow D. Reactor pressure fluctuates Place backup EHC pressure regulator in service Proposed Answer: C, Reactor power rises, Lower Recirculation flow

Explanation: Alarms listed are symptoms of a stuck open SRV. Rx power will rise due to loss of Feedwater heating. Immediate actions per AOP-62 (Loss FW Heating) direct lowering power

~ 100% using RWR. AOP-36 Subsequent Actions direct lowering Rx power only if SRV cannot be closed.

A. incorrect- sensed steam flow lowers and FW responds by lowering FW flow. This lowers vessel level and AOP-42, Feewater Control Malfunction, maybe entered on low level.

However manual control of feedwater is only allowed if the feedwater control system malfunctions. In this case, the controller is responding as it should.

B. incorrect- the candidate may correctly assume a lower main steam line flow also means lower generator load. The candidate may select the Turbine Trip if it is assumed that the negative reactivity (caused by depressurization and voiding) will cause power to decrease to the point where the generator would trip on reverse power (when normally the turbine is manually tripped before this happens).

C. correct-see above explanation D. incorrect- reactor pressure oscillations is a symptom of an EHC regulator malfunction (and SRV lifting). However, the backup regulator will not mitigate this event. Any oscillations will continue since the primary regulator was controlling as designed.

Technical Reference(s): AOP-36,62 Proposed references to be provided to applicants during examination: none Learning Objective: _SDLP-29 1.09 LP-AOP 1.03 Question Source: Bank#

Modified Bank # _ _ _ (Note changes or attach parent)

New x Question History: Last NRC Exam (Optional: Questions validated at the facility since 10195 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis 10 CFR Part 55 Content: 55.41 1§l 55.43 Comments:

Question #43 The Plant is operating at 35% power with the Main Generator "synchronized" to the Grid.

A large load near Rack 25-5 is dropped in the Reactor Building causing an RPV Level Instrument perturbation.

The following conditions exist for the Feedwater and Main Turbine Level Instruments:

06LT-52A (Narrow Range 'A') reads: 224.5 inches

06LT-52B (Narrow Range 'B') reads: 201 inches

06LT -52C (Narrow Range 'C') reads: 164.5 inches Which one of the following sets correctly lists the status of the Main Turbine and Rx?

Main Turbine A. Not tripped Not scrammed B. Tripped Scrammed C. Tripped Not scrammed D. Not tripped Scrammed

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO/SRO Tier# 2 Group # 1 KIA # 259002 K1.14 Importance Rating 2.9/3.0 Proposed Question: 43 Question #43 The Plant is operating at 35% power with the Main Generator "synchronized" to the Grid.

A large load near Rack 25-5 is dropped in the Reactor Building causing an RPV Level Instrument perturbation.

The following conditions exist for the Feedwater and Main Turbine Level Instruments:

06LT-52A (Narrow Range 'A') reads: 224.5 inches

06LT-52B (Narrow Range 'B') reads: 201 inches

06LT-52C (Narrow Range 'C') reads: 164.5 inches Which one of the following sets correctly lists the status of the Main Turbine and Rx?

Main Turbine A. Not tripped Not scrammed B. Tripped Scrammed C. Tripped Not scrammed D. Not tripped Scrammed Proposed Answer: A, Not tripped, Not scrammed

Explanation: The Main Turbine trip logic is 2 out of 3 high inputs required. One high and one low (below) the scram setpoint will not scram the Rx.

A. correct-see above explanation B. incorrect- see above. Student may mis-believe that a Level < 177 inches is a Scram\TT C. incorrect- see above. Student may mis-believe that only one high level is a TT D. incorrect- see above. Student may mis-believe that a Level < 177 inches is a Scram Technical Reference(s}: OP-27A Proposed references to be provided to applicants during examination: none Learning Objective: _SDLP-02B 1, .09 Question Source: Bank #

Modified Bank # _ _ _ (Note changes or attach parent)

New X Question History: Last NRC Exam (Optional: Questions validated at the facility since 10195 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 (2-9) 55.43 Comments:

Question #44 The Plant is operating at 75% with elevated Offgas Radiation levels.

The only equipment out of service is: Standby Gas Treatment train 'B'.

The following events take place:

The Rx scrams

A steam leak in the Reactor Building occurs

Reactor Building Ventilation Radiation monitors are upscale

A loss of Offsite power occurs

Emergency Diesel Generator 'A' fails to start With no Operator action, which one of the following correctly states the expected response of radiation levels in the Reactor Building?

Radiation levels ...

A. in the Reactor Building will lower due to RB Ventilation filtration B. in the Reactor Building will lower due to Standby Gas filtration C. inside the Reactor Building will rise due to no filtration D. outside the Reactor Building will rise due to positive pressure in RB

ES-401 Sample Written Examination Form ES-401*5 Question Worksheet Examination Outline Cross-

Reference:

Level RO/SRO Tier# 2 Group # 1

~JA# 261000 K6.03 Importance Rating 3.0/3.1 Proposed Question: 44 The Plant is operating at 75% with elevated Offgas Radiation levels.

The only equipment out of service is: Standby Gas Treatment train 'B'.

The following events take place:

The Rx scrams

A steam leak in the Reactor Building occurs

Reactor Building Ventilation Radiation monitors are upscale

A loss of Offsite power occurs

Emergency Diesel Generator 'A' fails to start With no Operator action, which one of the following correctly states the expected response of radiation levels in the Reactor Building?

Radiation levels ...

A. in the Reactor Building will lower due to RB Ventilation filtration B. in the Reactor Building will lower due to Standby Gas filtration C. inside the Reactor Building will rise due to no filtration D. outside the Reactor Building will rise due to positive pressure in RB Proposed Answer: B, in the Reactor Building will lower due to Standby Gas filtration

Explanation: A Rx scram followed by the loss of Offsite power will result in emergency busses 10500 and 10600 being energized by the Emergency Diesel Generators (EDGs).

RB Vent Rad monitors upscale will cause normal RB ventilation to isolate and both trains of Standby Gas Treatment to start and begin filtering and maintaining a negative pressure in the RB. However SBGT train 'B' was out of service. SBGT Fan 'A' is powered from MCC 151. which gets its power from the 10500 bus ('A' and 'C' EDGs). Therefore Fan 'A' will still be running because EDG 'C' will keep the 10500 bus energized only at a lower load capability.

A. incorrect-see above.

B. correct-see above explanation C. incorrect- see above.

D. incorrect- see above.

Technical Reference(s): OP-20,22 Proposed references to be provided to applicants during examination: none Learning Objective: _SDLP-01B 1.04, 1.10 Question Source: Bank#

Modified Bank # ___(Note changes or attach parent)

New X Question History: Last NRC Exam (Optional: Questions validated at the facility since 10195 wiff generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of eve/)' question.)

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis 10 CFR Part 55 Content: 55.41 ..ill 55.43 Comments:

Question #45 The Plant is operating in Mode 5 with the Spent Fuel Pool (SFP) gates removed.

The Control Room receives a report from the Refuel Floor that the RPV cavity level is lowering.

AOP-53 (Loss of Spent Fuel Storage Pool, Reactor Head Cavity Well, or Dryer Separator Storage Pit Water Level) is entered.

The Immediate Actions of AOP-53 state: "Ensure Reactor Building ventilation is isolated per Section G of OP-51A".

Which one of the following states the (1) correct manner to isolate RB ventilation and W the required differential pressure (dP) to be maintained in the Reactor Building?

correct manner required dP A. Depress the RB VENT ISOL A pushbutton then start HO.1 to (-)0.25 inches water train 'A' of Standby Gas Treatment B. Depress the RB VENT ISOL B pushbutton then start > (-)0.25 inches water train 'B' of Standby Gas Treatment C. Start Standby Gas Treatment train 'A' then depress > (-)0.25 inches water pushbutton RB VENT ISOL A and B D. Start Standby Gas Treatment train 'A' and 'B' then (-)0.1 to (-)0.25 inches water depress pushbutton RB VENT ISOL A and B

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO/SRO Tier# 2 Group # 1 KJA# 261000 A4.06 Importance Rating 3.3/3.6 Proposed Question: 45 The Plant is operating in Mode 5 with the Spent Fuel Pool (SFP) gates removed.

The Control Room receives a report from the Refuel Floor that the RPV cavity level is lowering.

AOP-53 (loss of Spent Fuel Storage Pool, Reactor Head Cavity Well, or Dryer Separator Storage Pit Water level) is entered.

The Immediate Actions of AOP-53 state: "Ensure Reactor Building ventilation is isolated per Section G of OP-51A".

Which one of the following states the (1) correct manner to isolate RB ventilation and m the required differential pressure (dP) to be maintained in the Reactor Buildinq?

correct manner required dP A. Depress the RB VENT ISOl A pushbutton then start (-)0.1 to (-)0.25 inches water train 'A' of Standby Gas Treatment B. Depress the RB VENT ISOl B pushbutton then start > (-)0.25 inches water train 'B' of Standby Gas Treatment C. Start Standby Gas Treatment train 'A' then depress > (-)0.25 inches water pushbutton RB VENT ISOl A and B D. Start Standby Gas Treatment train 'A' and 'B' then (-)0.1 to (-)0.25 inches water depress pushbutton RB VENT ISOL A and B Proposed Answer: C, Start Standby Gas Treatment train 'A' then depress pushbutton RB VENT ISOL A and B, > (-)0.25 inches water

Explanation: OP-53 requires RB vent to be isolated per OP-51A. First step of OP-51A requires SBGT running per OP-20 before securing RB ventilation. Therefore, start one train of SBGT, ensure> negative .25 inches water then isolate RB vent by depressing both ISOL push buttons A. incorrect-see above. This would physically work but is not correct B. incorrect- see above. This would physically work but is not correct C. correct-see above explanation D. incorrect- see above. This would physically work but is not correct Technical Reference(s): AOP-53, OP-51A, 20 Proposed references to be provided to applicants during examination: none Learning Objective: _SDLP-01B 1.02,1.15 Question Source: Bank #

Modified Bank # _ _ _ (Note changes or attach parent)

New Question History: Last NRC Exam (Optional: Questions validated at the facility since 10195 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge _X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 m 55.43 Comments:

Question #46 The Plant is operating at 100% power with the 10100 bus being supplied via the 10102 breaker (Normal Supply Breaker to the 10100 Bus) when ...

A loss of DC Control Power to Breaker 10102 occurs.

Which one of the following correctly states the automatic response of the 10100 bus to this condition?

The 10100 bus ...

A. immediately de-energizes and remains de-energized B. continuously remains energized via the 10102 breaker C. continuously remains energized via the 10112 breaker D. de-energizes then re-energizes via the 10112 breaker

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross~

Reference:

Level RO/SRO Tier# 2 Group # 1 KIA # 262001 K6.01 Importance Rating 3.1/3.4 the Proposed Question: 46 The Plant is operating at 100% power with the 10100 bus being supplied via the 10102 breaker (Normal Supply Breaker to the 10100 Bus) when ...

A loss of DC Control Power to Breaker 10102 occurs.

Which one of the following correctly states the automatic response of the 10100 bus to this condition?

The 10100 bus ...

A. immediately de-energizes and remains de-energized B. continuously remains energized via the 10102 breaker C. continuously remains energized via the 10112 breaker D. de-energizes then re-energizes via the 10112 breaker Proposed Answer: B, continuously remains energized via the 10102 breaker

Explanation: A loss of DC control power to the 4160 VAC breaker 10102 will not cause it to automatically open due to not being able to energize the Trip Coil. The breaker can only be opened manually \ locally. Therefore, the 10100 bus continuously remains energized.

A. incorrect-see above B. correct-see above explanation C. incorrect- see above D. incorrect- see above Technical Reference(s): OP-46A Proposed references to be provided to applicants during examination: none Learning Objective: - SDLP-71E- 1.11 Question Source: Bank#

Modified Bank # _ _ _ (Note changes or attach parent)

New x Question History: Last NRC Exam (Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge _X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 m 55.43 Comments:

Question #47 The Plant is operating at power when the following occurs:

12600 Bus de-energizes

The UPS Inverter section trips Which one of the following correctly states the status of the 71 ACUPS Distribution bus?

The 71ACUPS Distribution bus has had (1) ___ and is being powered from (2) _ _

(1 ) (2)

A. a momentary interruption Battery 71 BCB-2A B. a complete interruption no power source C. no interruption Battery 71 BCB-2A D. no interruption MCC-252 (Alternate Feed)

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

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l.evel RO/SRO Tier# 2 Group # 1 KIA # 262002 A3.01 Importance Rating 2.8/3.1 Proposed Question: 47 The Plant is operating at power when the following occurs:

12600 Bus de-energizes

The UPS Inverter section trips Which one of the following correctly states the status of the 71ACUPS Distribution bus?

The 71ACUPS Distribution bus has had (1) ___ and is being powered from (2) _ _

(1 ) (2)

A. a momentary interruption Battery 71 BCB-2A B. a complete interruption no power source C. no interruption Battery 71 BCB-2A D. no interruption MCC-2S2 (Alternate Feed)

Proposed Answer: D, no interruption MCC-2S2 (Alternate Feed)

Explanation: 12600 bus feeds MCC-262 which is the normal feed to the UPS. If it is unavailable, the UPS transfers to battery supply and uses the Inverter section. With a short in the Inverter, the alternate feed (MCC-252) will supply the UPS bus. The UPS has a static electronic switch that seamlessly transfers all power sources to the 71ACUPS Distribution bus.

A. incorrect-see above B. incorrect- see above C. incorrect- see above D. correct-see above explanation Technical Reference(s): OP-46B Proposed references to be provided to applicants during examination: none Learning Objective: _SDLP-71F 1.05_

Question Source: Bank#

Modified Bank # _ _ _ (Note changes or attach parent)

New x Question History: Last NRC Exam (Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 ill 55.43 Comments:

Question #48 The plant is at full power operations

An NPO reports that LPCI MOV Bus "Au Main Battery Breaker 1CB2 has tripped.

Should a DBA LOCA occur, complete the following statement. ..

Without the battery, the battery charger, (1) provide sufficient power to the LPCI bus and/but the Alternate (Maintenance) Feed (2) provide sufficient power the LPCI bus.

(1) (2)

A. could NOT could NOT B. could NOT could C. could could NOT D. could could

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

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Level RO/SRO Tier # 2 Group # 1 KJA# 263000 2.1.20 Importance Rating 4.6/4.6 Proposed Question: 48 The plant is at full power operations

An NPO reports that LPCI MOV Bus "A" Main Battery Breaker 1CB2 has tripped.

Should a DBA LOCA occur, complete the following statement. ..

Without the battery, the battery charger, (1) provide sufficient power to the LPCI bus and/but the Alternate (Maintenance) Feed (2) provide sufficient power the LPCI bus.

(1) (2)

A. could NOT could NOT B. could NOT could C. could could NOT D. could could Proposed Answer: B, Could NOT, could

Explanation: C.2.5 Precaution in OP-43C. A LPCI MOV independent power supply inverter shall not be lined up to power its associated LPCI MOV bus with the Main Battery Breaker 1CB2 open. The charger is not able to provide adequate current to the inverter to allow stroking any of the MOVs. During a LPCI initiation, there are procedural steps that allow the Alternate Feed to supply the bus during a LPCI event.

A. incorrect-see above B. correct-see above C. incorrect- see above D. incorrect-see above Technical Reference(s): OP-13B, AOP-43A1B, Proposed references to be provided to applicants during examination:

Learning Objective: _SDLP-71B 1.05,1.08,1.13_

Question Source: Bank#

Modified Bank # _ _ _ (Note changes or attach parent)

New Question History: Last NRC Exam (Optional: Questions validated at the facility since 10195 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detai/ed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis _ X_

10 CFR Part 55 Content: 55.41 11Q) 55.43 Comments:

Question #49 The Plant just completed a normal Reactor shutdown.

A leak in Containment occurs with the following timeline conditions:

T=O sec Orywell pressure is 2.9 psig and rising slowly

T=1 sec Emergency Diesel Generator 'C' fails to start

T=5sec RPV level is 52.5 inches and lowering fast

T=10 sec Transformer T3 de-energizes Which one of the following correctly states the response of the RHR system at time T=60 sec?

A. Only RHR pump 'A' auto started B. Only RHR pumps 'A' and 'C' auto started C. Only RHR pumps 'A', 'C' and '0' auto started O. All four RHR pumps auto started

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO/SRO Tier# 2 Group # 1 KJA# 264000 A3.05 Importance Rating 3.4/3.5 Proposed Question: 49 The Plant just completed a normal Reactor shutdown.

A leak in Containment occurs with the following timeline conditions:

T=O sec Orywell pressure is 2.9 psig and rising slowly

T=1 sec Emergency Diesel Generator 'C' fails to start

T=5sec RPV level is 52.5 inches and lowering fast

T=10 sec Transformer T3 de-energizes Which one of the following correctly states the response of the RHR system at time T=60 sec?

A. Only RHR pump 'A' auto started B. Only RHR pumps 'A' and 'C' auto started C. Only RHR pumps 'A', 'C' and '0' auto started O. All four RHR pumps auto started Proposed Answer: C, Only RHR pumps 'A', 'C' and '0' auto start

Explanation: 2.7 psig and < 59.5 inches is an auto start signal to all four RHR pumps. However, the EOG load shedding \ sequencing will not allow the second RHR pump to auto start if paired EOG output breaker is not shut (i.e. both EOGs powering its respective bus). Therefore only 3 RHR pumps will \ can auto start.

A. incorrect-see above B. incorrect-see above C. correct-see above explanation O. incorrect-see above Technical Reference(s): OP-22 Proposed references to be provided to applicants during examination: none Learning Objective: _SDLP-93 1.05, 1.09_

Question Source: Bank#

Modified Bank # _ _ _ (Note changes or attach parent)

New X Question History: Last NRC Exam (Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 m 55.43 Comments:

Question #50 The Plant is operating at 100% power with MOD 10017 tagged open due to Maintenance, when the following transient occurs:

An NPO reports lightning strike in the 115KV switchyard

Annunciators received:

~ 09-8-6-15 NMP-FITZ 115KV LINE 4 BKR 10012 TRIP

~ 09-5-1-2 MSIVs NOT FULL OPEN TRIP

The SNO reports all MSIVs have both red and green lights illuminated

~ No Operator actions are taken Ten (10) minutes after the transient begins, how many instrument air compressors, if any, would be available?

A. 0 B. 1 C. 2 D. 3

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO/SRO Tier # 2 Group # 1 KIA # 300000 K2.01 Importance Rating 2.8/2.8 Proposed Question: 50 The Plant is operating at 100% power when the following transient occurs:

An NPO reports lightning strike in the 'I 15KV switchyard

Annunciators received:

)i> 09-8-6-15 NMP-FITZ 115KV LINE 4 BKR 10012 TRIP

)i> 09-5-1-2 MSIVs NOT FULL OPEN TRIP

The SNO reports all MSIVs have both red and green lights illuminated

)i> No Operator actions are taken Ten (10) minutes after the transient begins, how many instrument air compressors, if any, would be available?

A. 0 B. 1 C. 2 D. 3 Proposed Answer: B, 1

Explanation: Power supplies to SAC's are: 'A': L23 (10300) 'B': L24 (10400) 'C' L33 (10300).

With the plant at power, the plant gets internal power from the T4 transformer. The trip of 10012 caused a loss of the T3 transformer which powers the 10300 bus with the Unit off line. At 100%

power, the mode switch is in RUN, and the closure of the MSIVs will cause a reactor scram. The scram will cause a generator/turbine trip (reverse power as steam flow decays down). The plant will normally fast transfer from the T4 transformer to the T2 and T3 transformers. With only the T2 transformer available, only "B" air compressor will be available.

A. incorrect-see above B. correct-see above C. incorrect-see above D. incorrect-see above Technical Reference(s): OP-39 Proposed references to be provided to applicants during examination: none Learning Objective: _SDLP-39 1.03, 1.05_

Question Source: Bank #

Modified Bank # _ _ _ (Note changes or attach parent)

New Question History: Last NRC Exam (Optional: Questions validated at the facility since 10195 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 ..ill 55.43 Comments:

Question #51 The Plant is operating at rated power with the following Service Air Compressor (SAC) lineup:

39AC-2A: Lead and running loaded

39AC-2C: 1st Standby

39AC-2B: 2nd Standby The following EPIC alarm is received in the Control Room:

EPIC-A-1539 39PT-10l l02psig NPO reports from the field:

Air Compressor 39AC-2A Trouble Alarm is in.

39AC-2A inlet air filter is showing a hi~Jh dP and is clogged with debris.

39AC-2A has an Inlet Restriction reading of 5 psig steady Based on this report \ conditions, which one of the following choices correctly states the current Service Air Compressor 39AC-2A, B, C alignment?

A ~ C A. Running Running Running B. Running Running Standby C. Tripped Standby Standby D. Tripped Running Running

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO/SRO Tier # 2 Group # 1 KJA# 300000 K5.13 Importance Rating 2.9/2.9 Proposed Question: 51 The Plant is operating at rated power with the following Service Air Compressor (SAC) lineup:

39AC-2A: Lead and running loaded

39AC-2C: 1st Standby

39AC-2B: 2nd Standby The following EPIC alarm is received in the Control Room:

EPIC-A-1539 39PT-10l l02psig NPO reports from the field:

Air Compressor 39AC-2A Trouble Alarm is in.

39AC-2A inlet air filter is showing a high dP and is clogged with debris.

39AC-2A has an Inlet Restriction reading of 5 psig steady Based on this report \ conditions, which one of the following choices correctly states the current Servic Air Compressor 39AC-2A, B, C alignment?

A § C A. Running Running Running B. Running Running Standby C. Tripped Standby Standby D. Tripped Running Running Proposed Answer: D, Tripped Running Running

Explanation: With an Instrument Air pressure value of 102 psig, all three SAC should be running however an Inlet Restriction value of 5 psig is a compressor trip. Therefore 'A' is tripped.

A. incorrect-see above B. incorrect-see above C. incorrect-see above D. correct-see above explanation Technical Reference(s): OP-39 Proposed references to be provided to applicants during examination: none Learning Objective: _SDLP-39 1.08, 14_

Question Source: Bank#

Modified Bank # _ _ _ (Note changes or attach parent)

New Question History: Last NRC Exam (Optional: Questions validated at the facility since 10195 will generally undergo Jess rigorous review by the NRC; failure to provide the information will necessitate a detaifed review of eveljl question.)

Question Cognitive Level: Memory or Fundamental Knowledge x Comprehension or Analysis 10 CFR Part 55 Content: 55.41 J.§l 55.43 Comments:

Question #52 The Plant is operating at rated power during the month of January with the following component lineup:

Service Water Pump (SWP) 'c' is running

Service Water Header pressure is 95 psig

TBCLC pumps 'A' and 'c' are running

TBCLC Header pressure is 87 psig

Circulating Water inlet temperature is 39 'F The following occurs:

The 10700 Bus de-energizes With no Operator action, which one of the following choices correctly describes the component alignment of the Service Water and TBCLC systems after 10 minutes?

Service Water TBClC A. No pumps running 'A' pump only running B. 'A' and 'B' pumps running 'A' and 'B' pumps running C. 'B' pump only running 'A' pump only running D. 'A' pump only running 'A' and 'B' pumps running

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

level ROISRO Tier # 2 Group # 1 KIA # 400000 K6.05 Importance Rating 2.8/2.9 a Of Proposed Question: 52 The Plant is operating at rated power during the month of January with the following component lineup:

Service Water Pump (SWP) 'C' is running

Service Water Header pressure is 95 psig

TBClC pumps 'A' and 'C' are running

TBClC Header pressure is 87 psig

Circulating Water inlet temperature is 39 'F The following occurs:

The 10700 Bus de-energizes With no Operator action, which one of the following choices correctly describes the component alignment of the Service Water and TBClC systems after 10 minutes?

Service Water Teele A. No pumps running 'A' pump only running B. 'A' and 'B' pumps running 'A' and 'B' pumps running C. 'B' pump only running 'A' pump only running D. 'A' pump only running 'A' and 'B' pumps running Proposed Answer: B, 'A' and 'B' pumps running, 'A' and 'B' pumps running

Explanation: Loss of 10700 Bus results in 'C' SWP and TBCLC pumps tripping. When the

'C' SWP electrically trips, both SWP 'A' and 'B' receive an auto start signal in their respective 4160 Vac breaker start ckt. When "C' TBCLC pump trips, TBCLC header pressure will lower. When 85 psig is sensed, the standby 'B' TBCLC auto starts.

A. incorrect-see above B. correct-see above explanation C. incorrect-see above D. incorrect-see above Technical Reference(s): OP-13C and OP-41 Proposed references to be provided to applicants during examination: none Learning Objective: _SDLP-37_1.05, 10 and SDLP-46A 1.05,10 Question Source: Bank#

Modified Bank # _ _ _ (Note changes or attach parent)

New Question History: Last NRC Exam (Optional: Questions validated at the facility since 10195 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 Jll 55.43 Comments:

Question #53 The Plant is operating at power with the following conditions:

'A' & 'B' TBCLC pumps in service

'C' TBCLC pump has been started for post maintenance testing following pump rebuild.

'B' TBCLC pump is stopped 5 minutes later the following occurs:

o Annunciator 09-6-2-25 (TBC MAKEUP TK LVL HI OR LO) alarms o TBCLC Header Pressure is observed to be 110 psig The System Engineer has determined the 'A' TBCLC Heat Exchanger has a tube leak.

The cause of the makeup tank alarm is from a (1) level which the system ---'........._

designed to automatically compensate for.

ill m A. Low is B. Low is NOT C. High is D. High is NOT

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

level RO/SRO Tier# 2 Group # 1 KIA # 400000 A 1.04 Importance Rating 2.8/2.8 Proposed Question: 53 The Plant is operating at power with the following conditions:

'A' & 'B' TBClC pumps in service

'C' TBClC pump has been started for post maintenance testing following pump rebuild

'B' TBClC pump is stopped 5 minutes later the following occurs:

o Annunciator 09-6-2-25 (TBC MAKEUP TK l Vl HI OR lO) alarms o TBClC Header Pressure is observed to be 110 psig The System Engineer has determined the 'A' TBClC Heat Exchanger has a tube leak.

The cause of the makeup tank alarm is from a (1) level which the system -->.:.l-_

designed to automatically compensate for.

ill g}

A. Low is B. low is NOT C. High is D. High is NOT Proposed Answer: A - low, is

Explanation: Normal TBCLC header pressure is 95 to 135 psig. The normal range for SW pressure is 85 to 100 psig. The start of the TBCLC pump caused a heat exchanger leak to develop. With TBCLC pressure higher than the normal SW pressure range, the leak is into SW.

Makeup tank level decreases. However, the system has an automatic makeup valve which compensates for the lowering level.

A. correct-see above explanation B. incorrect- selected if the candidate remembers there is a manual fill valve that is used (and proceduralized) but doesn't remember there is an auto makeup valve.

C. incorrect- selected if candidate confuses the SW and TBCLC normal pressures but remembers that there is an overfill line from the tank vent that goes to floor drains.

D. incorrect- selected if candidate confuses the SW and TBCLC normal pressures and does not know/remembers that there is an overfill line from the tank vent that goes to floor drains.

Technical Reference(s): OP-41 Proposed references to be provided to applicants during examination: none Learning Objective: _SDLP-37 1.05,1.12_

Question Source: Bank #

Modified Bank # _--,-:-_ (Note changes or attach parent)

New Question History: Last NRC Exam (Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis 10 CFR Part 55 Content: 55.41 JQl 55.43 Comments:

Question #54 The Plant has just completed a normal shutdown \ cooldown in preparation to isolate the 'A' Recirculation Pump (RWR).

RPV pressure is 210 psig, lowering slowly

RWR pump 'A' is tripped Which one of the following choices states the correct manipulation of RWR Seal Purge and the reason why?

A. Maintain seal purge in order to preserve the integrity of the RWR pump casing B. Lower seal purge rate due to a reduced heat load on the CRD system C. Raise seal purge rate to maintain differential pressure across the seals D. Secure seal purge to prevent over pressurizing the RWR seal casing

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO/SRO Tier# 2 Group # ~

k'JA # 201001 K1.03 Importance Rating 3.1/3.1 of the "in;,1<::'" cormections and/or hetween CONTt'¥OL ROD HYDRAUUC S'(STEM <utd the Rocin:::ufation pumps Proposed Question: 54 The Plant has just completed a normal shutdown \ cooldown in preparation to isolate the 'A' Recirculation Pump (RWR).

RPV pressure is 210 psig, lowering slowly

RWR pump 'A' is tripped Which one of the following choices states the correct manipulation of RWR Seal Purge and the reason why?

A. Maintain seal purge in order to preserve the integrity of the RWR pump casing B. Lower seal purge rate due to a reduced heat load on the CRD system C. Raise seal purge rate to maintain differential pressure across the seals D. Secure seal purge to prevent over pressurizing the RWR seal casing Proposed Answer: D, Secure seal purge to prevent over pressurizing the RWR seal casing

Explanation: Each RWR pump provided with 2-4 gpm mini-purge flow from CRD to the RWR seals. Mini-purge must be valved-out (with a manual valve) when the pump is isolated from the Reactor to prevent excessive cooldown of and/or over-pressurization of the pump and seal casing.

A. incorrect-see above B. incorrect-see above C. incorrect-see above D. correct-see above explanation Technical Reference(s): OP-25 and OP-27 Proposed references to be provided to applicants during examination: none Learning Objective: _SDLP-02H 1.. 05_

Question Source: Bank#

Modified Bank # _ _ _ (Note changes or attach parent)

New Question History: Last NRC Exam (Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 fl:.ID 55.43 Comments:

Question #55 The Plant is conducting a Reactor startup. Control Rod 10-19 is being withdrawn continuously from position 00 to 48 per OP-16.

As Control Rod 10-19 is withdrawing past position 32, the following occurs:

The Notch Override Switch is released and spring returns to Off.

The Rod Movement Control switch remains in Notch Out.

Which one of the following choices correctly lists the response of Control Rod 10-19?

Control Rod 10-19 ... Settles A. continues to 48 No B. continues to 48 Yes C. stops at 34 No D. stops at 34 Yes

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO/SRO Tier# 2

(;roup # ~

KJA# 201002 K4.05 Importance Rating 3.3/3.3 nO:WfE!CO'G of r<EACTOR MANUAL CONTROL SYSTEM fot lhe fDHnWlfHl "Notch Proposed Question: 55 The Plant is conducting a Reactor startup. Control Rod 10-19 is being withdrawn continuously from position 00 to 48 per OP-16.

As Control Rod 10-19 is withdrawing past position 32, the following occurs:

The Notch Override Switch is released and spring returns to Off.

The Rod Movement Control switch remains in Notch Out.

Which one of the following choices correctly lists the response of Control Rod 10-197 Control Rod 10-19 ... Settles A. continues to 48 No B. continues to 48 Yes C. stops at 34 No D. stops at 34 Yes Proposed Answer: D, stops at 34 Yes

Explanation: To continuously withdraw a Control Rod, two switches must be manipulated simultaneously; Notch Override and Notch Out. As soon as either control switch is released, the RMCS timer will complete its cycle and the rod will settle into an even notch position.

A. incorrect-see above B. incorrect-see above C. incorrect-see above D. correct-see above explanation Technical Reference(s): OP-26 Proposed references to be provided to applicants during examination: none Learning Objective: _SDLP-03F_1.05 Question Source: Bank#

Modified Bank # _~_ (Note changes or attach parent)

New Question History: Last NRC Exam (Optional: Questions validated at the facility since 10195 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis 10 CFR Part 55 Content: 55.41 JD 55.43 Comments:

Question #56 The Plant is operating at 75% power with the 'A' Recirculation Pump (RWR) Scoop Tube "locked".

The following conditions exist:

'A' RWR Scoop Tube Auto Unlock: On

'B' Reactor Feedwater Pump Turbine (RFT) Bearing Oil pressure: 3 psig and steady.

RPV level: 192 inches and steady Which one of the following choices correctly lists the status of the 'A' RWR MG set?

A. "locked" B. 44%

C.30%

D. tripped

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

L.evel RO/SRO Tier# 2 Group # Z

~(jA # 202001 A1.01 Importance Rating 3.6/3.5 Proposed Question: 56 The Plant is operating at 75% power with the 'A' Recirculation Pump (RWR) Scoop Tube "locked".

The following conditions exist:

'A' RWR Scoop Tube Auto Unlock: On

'B' Reactor Feedwater Pump Turbine (RFT) Bearing Oil pressure: 3 psig and steady.

RPV level: 192 inches and steady Which one of the following choices correctly lists the status of the 'A' RWR MG set?

A. "locked" B.44%

C.30%

D. tripped Proposed Answer: B, 44%

Explanation: RFT Bearing Oil pressure of 3 psig is a Feed pump trip. That trip and an RPV level of

< 196.5 is a Runback of the RWR MG set to 44%. With the Scoop Tube locked, an MG set won't runback unless the auto unlock feature is active (on) then it will automatically unlock and runback.

A. incorrect-see above B. correct-see above explanation C. incorrect-see above D. incorrect-see above Technical Reference(s): OP-27 and OP-3 Proposed references to be provided to applicants during examination: none Learning Objective: _SDLP-021_1.05 Question Source: Bank#

Modified Bank # _--,--:-_ (Note changes or attach parent)

New Question History: Last NRC Exam (Optional: Questions validated at the facility since 10195 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of eve!}' question.}

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 1§1 55.43 Comments:

Question #57 The Plant is operating at power when the following Annunciator alarms:

09-5-1-60 CRD TEMP HI Which one of the following choices correctly lists (1) where this temperature is being sensed and (2) what the significance of the alarm is?

(1) sensed location (2) significance A. Stub tube Stub tube leakage B. Indicator tube Seal degradation C. Index tube Rod drift D. Piston tube Rod speed

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO/SRO Tier # 2 Group # ~

KJA# 214000 K4.02 Importance Rating 2.5/2.5 Proposed Question: 57 The Plant is operating at power when the following Annunciator alarms:

09-5-1-60 CRD TEMP HI Which one of the following choices correctly lists (1) where this temperature is being sensed and (2) what the significance of the alarm is?

(1) sensed location (2) significance A. Stub tube Stub tube leakage B. Indicator tube Seal degradation C. Index tube Rod drift D. Piston tube Rod speed Proposed Answer: B, Indicator tube Graphitar seal degradation

Explanation: The Indicator tube houses the Position Indicator Probe (PIP). A thermocouple is placed at the top of the PIP to monitor for high temps in the CRD Graphitar seals. The seals become brittle at high temperatures, and seal breakdown can result in undesirable drive speeds and scram times.

A. incorrect-see above B. correct-see above explanation C. incorrect-see above D. incorrect-see above Technical Reference(s): FSAR Proposed references to be provided to applicants during examination: none Learning Objective: _SDLP-3A_1.05_

Question Source: Bank#

Modified Bank # _ _ _ (Note changes or attach parent)

New Question History: Last NRC Exam (Optional: Questions validated at the facility since 10195 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of evelY question)

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis 10 CFR Part 55 Content: 55.41 JD 55.43 Comments:

Question #58 The Plant is operating at power performing a Traversing In-core Probe (TIP) calibration of LPRMs.

When the TIP detector is almost full in-core, the following occurs:

The Reactor scrams on high Drywell Pressure.

71ACNMS (power supply to TIP Ball Valves) de-energizes.

Which one of the following correctly states the response of the TIP system?

The TIP system ...

A. completes its normal trace run then returns to the lead shield.

B. immediately retracts to the lead shield then the ball valve closes.

C. does not retract to the lead shield and the ball valve remains open.

D. shear valves fire and isolates the guide tube penetration.

ES-401 Sample Written Examination Form 1:5-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO/SRO Tier # 2 Group # g KIA # 215001 K1.05 Importance Rating 3.3/3.4 Proposed Question: 58 The Plant is operating at power performing a Traversing In-core Probe (TIP) calibration of LPRMs.

When the TIP detector is almost full in-core, the following occurs:

The Reactor scrams on high Drywell Pressure.

71ACNMS (power supply to TIP Ball Valves) de-energizes.

Which one of the following correctly states the response of the TIP system?

The TIP system ...

A. completes its normal trace run then returns to the lead shield.

B. immediately retracts to the lead shield then the ball valve closes.

C. does not retract to the lead shield and the ball valve remains open.

D. shear valves fire and isolates the guide tube penetration.

Proposed Answer: C, does not retract to the lead shield and the ball valve remains open.

Explanation: High Drywell pressure scram setpoint is 2.7# and is also an auto retract signal for the TIP system. The Ball valves fail closed on a loss of power to them however, with the cable inserted in the guide tube, the ball valve cannot go closed. Loss of 71ACNMS also is a loss of retract power.

A. incorrect-see above B. incorrect-see above C. correct-see above explanation D. incorrect-see above Technical Reference(s): RAP-7.3.14 Proposed references to be provided to applicants during examination: none Learning Objective: _SDLP-07F_1.04 and 1.10 Question Source: Bank #

Modified Bank # _ _ _ (Note changes or attach parent)

New x Question History: Last NRC Exam (Optional: Questions validated at the facility since 10195 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis _X_

10 CFR Part 55 Content: 55.41 (2 to 9) 55.43 Comments:

Question #59 The Plant is operating at power when following isolation of Reference Leg Backfill to Level transmitter 02-3LT-72A, the following occurs:

A leak on the Reference leg of level transmitter 02-3LT-72A occurs, 02-3LT-72A is an input to the RHR LPCI auto initiation circuit

The leak results in OW pressure reaching 2.7 psig and rising slowly.

Concerning only the status of LPCI, which one of the following selections correctly lists (1) the condition of 02-3LT-72A and (2) the status of the LPCI Mode of RHR?

{1} 02-3LT-72A {2} status of the LPCI Mode of RHR A. Downscale Running B. Upscale Running C. Downscale Not running D. Upscale Not running

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO/SRO Tier# 2 Group # ~

KIA # 216000 K1.05 Importance Rating 3.7/3.8 Instrumentation: of

/::JlJtween NUCLEAR BO!LEPr INSTRUMENTATION and heat removal.

Proposed Question: 59 The Plant is operating at power when following isolation of Reference Leg Backfill to level transmitter 02-3LT -72A the following occurs:

A leak on the Reference leg of level transmitter 02-3LT-72A occurs, 02-3LT -72A is an input to the RHR LPCI auto initiation circuit

The leak results in DW pressure reaching 2.7 psig and rising slowly.

Concerning only the status of LPCI, which one of the following selections correctly lists ill..!.b&.

condition of 02-3LT-72A and (2) the status of the LPCI Mode of RHR?

(1} 02-3LT-72A (2} status of the LPCI Mode of RHR A. Downscale Running B. Upscale Running C. Downscale Not running D. Upscale Not running Proposed Answer: B, Upscale Running

Explanation: A leak in the Reference leg of transmitter 02-3LT-72A causes minimum dP being felt across the transmitter diaphragm. Minimum clP results in an upscale indication. LPCI initiation is 2.7# or 59.5 inches. The logic arrangement of the LPCI ckt is designed so that a single detector failure will not result in an inadvertent initiation or prevention of initiation.

A. incorrect-see above B. correct-see above explanation C. incorrect-see above D. incorrect-see above Technical Reference(s): OP-13 and OP-27A Proposed references to be provided to applicants during examination: none Learning Objective: - SDLP-10 1.10-Question Source: Bank #

Modified Bank # _ _ _ (Note changes or attach parent)

New x Question History: Last NRC Exam (Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis 10 CFR Part 55 Content: 55.41 (2 to 9) 55.43 Comments:

Question #60 The Plant has just placed RHR Pump 'B' in the Fuel Pool Cooling Assist (FPCA) mode of operation when the following events occur:

RPV level lowers to 58 inches and is continuing to lower slowly

The 10500 bus receives a "lock-out" condition Which one of the following correctly states the response of the RHR system to the above conditions?

A. RHR pump '0' aligns to the FPCA mode of operation B. RHR pump 'B' remains in the FPCA Mode of operation C. RHR pump 'A' aligns to the LPCI Mode of operation D. RHR Pump 'e' aligns to the LPCI Mode of operation

ES-401 Sample Written Examination Form ES*401*5 Question Worksheet Examination Outline Cross-

Reference:

Level RO/SRO Tier # 2 Group # f.

KIA # 233000 K 2.02 Importance Rating 2.8/2.9 Proposed Question: 60 The Plant has just placed RHR Pump '8' in the Fuel Pool Cooling Assist (FPCA) mode of operation when the following events occur:

RPV level lowers to 58 inches and is continuing to lower slowly

The 10500 bus receives a "lock-out" condition Which one of the following correctly states the response of the RHR system to the above conditions?

A. RHR pump '0' aligns to the FPCA mode of operation B. RHR pump '8' remains in the FPCA Mode of operation C. RHR pump 'A' aligns to the LPCI Mode of operation D. RHR Pump 'C' aligns to the LPCI Mode of operation Proposed Answer: D. Pump 'C' aligns to the LPCI Mode of operation

Explanation: RHR pumps A \ C are in Loop A and pumps B \ 0 are in Loop B. A Lock-out of the 10500 bus results in loss of power to RHR pumps A and B. RPV level ~ 59.5 inches is a LPCI initiation signal. Loop B will not re-align to the FPCA mode due to jumpers being installed to prevent the LPCI mode of operation from occurring. Since RHR pump C has power to it, it is the only component that can align to the LPCI mode of operation.

A. incorrect-see above B. incorrect-see above C. incorrect-see above D. correct-see above explanation Technical Reference(s): OP-13 and 13F Proposed references to be provided to applicants during examination:

Learning Objective: _SDLP-1 0 1.10 Question Source: Bank#

Modified Bank # _ _ _ (Note changes or attach parent)

New Question History: Last NRC Exam (Optional: Questions validated at the facility since 10195 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or AnalysiS _X_

10 CFR Part 55 Content: 55.41 ill 55.43 Comments:

Question #61 The Plant is in the process of Refueling when the following SEQUENCE of events occurs:

1. All Control Rods are full in

2. A Fuel Bundle is latched (grappled) in the Spent Fuel Pool

3. A Control Rod is withdrawn two notches by a Control Room Operator

4. The Fuel Bundle is raised to the full up position

5. The Refuel Bridge moves over the core

6. The Fuel Bundle is lowered into the core Which one of the following describes which Refuel Interlock first failed?

The first Refuel Interlock that failed was when ...

A. the Refuel Bridge moved over the core.

B. the Fuel Bundle was lowered into the core.

C. a Control Rod was selected.

D. a Control Rod was moved two notches.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level ROISRO Tier # 2 Group # 6 KIA # 234000 A 3.02 Importance Rating 3.1/3.7 Proposed Question: 61 The Plant is in the process of Refueling when the following SEQUENCE of events occur:

1. All Control Rods are full in

2. A Fuel Bundle is latched (grappled) in the Spent Fuel Pool

3. A Control Rod is withdrawn two notches by a Control Room Operator

4. The Fuel Bundle is raised to the full up position

5. The Refuel Bridge moves over the core

6. The Fuel Bundle is lowered into the core Which one of the following describes which Refuel Interlock first failed?

The first Refuel Interlock that failed was when ...

A. the Refuel Bridge moved over the core.

B. the Fuel Bundle was lowered into the core.

C. a Control Rod was selected.

D. a Control Rod was moved two notches.

Proposed Answer: A, Refuel Bridge moved over the core.

Explanation: The first violated interlock was the Refuel Bridge Reverse stop, which should have stopped the Bridge as it first moved over the core since a control rod is withdrawn, the mode switch is in REFUEL, and the main grapple is loaded. A second interlock was violated when the grapple lowered with a control rod withdrawn and the main grapple loaded (Fuel Hoist Interlock).

The reason for these interlocks is to prevent criticality while refueling.

A. correct-see above explanation B. incorrect-see above C. incorrect-see above D. incorrect-see above Technical Reference(s): OP-66 Proposed references to be provided to applicants during examination: none Learning Objective: _SDLP-8A 1.09__

Question Source: Bank #

Modified Bank # _ _ _ (Note changes or attach parent)

New x Question History: Last NRC Exam (Optional: Questions validated at the facility since 1019,5 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis 10 CFR Part 55 Content: 55.41 m 55.43 Comments:

Question #62 The plant is at 50% power when the followin!~ alarm comes in:

09-7-3-10 MOIST SEP DRN TK 2A LVL HI Level in the drain tank continues to rise until it is 3" from the bottom of the MSR and picks up a second level switch.

A few minutes after the switch has picked up, the operator checks the MSRs First Stage Extraction Steam Supply.

The operator should expect to find that the steam supply _ _ __

A. In service, full steam flow B. Is isolated, no steam flow C. Throttled with the Hi Load Controller in service D. Throttled with the Lo Load Controller in service

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO/SRO Tier # 2 Group # 6.

KIA # 239001 A 3.03 Importance Rating 2.8/2.8 Proposed Question: 62 The plant is at 50% power when the following alarm comes in:

09-7-3-10 MOIST SEP DRN TK 2A LVL HI Level in the drain tank continues to rise until it is 3" from the bottom of the MSR and picks up a second level switch.

A few minutes after the switch has picked up, the operator checks the MSRs First Stage Extraction Steam Supply.

The operator should expect to find that the steam supply _ _ __

A. In service, full steam flow B. Is isolated, no steam flow C. Throttled with the Hi Load Controller in service D. Throttled with the Lo Load Controller in service Proposed Answer: B, Is isolated, no steam flow Explanation: The second level switch on the MSR drain tank causes a Turbine Trip. On a Turbine Trip, the Non-Return Valves get a close signal. The Steam Stop to the First Stage Reheater also gets a isolation signal. The steam supply to the 2nd stage reheater does not isolate on a turbine trip but does have hi and 10 load controllers (which the 15t stage does not).

A. incorrect- this would be selected if the candidate did not recognize that a turbine trip occurred (and the resultant isolation).

B. correct-see above explanation C. incorrect- this would be selected if the candidate recognized the turbine trip but confused the first stage steam supply with the second stage steam supply D. incorrect- this would be selected if the candidate did NOT recognize the turbine trip and confused the first stage steam supply with the second stage steam supply.

Technical Reference(s): OP-10 Proposed references to be provided to applicants during examination: none Learning Objective: _SDLP-94A 1.05_

Question Source: Bank#

Modified Bank # _-:-:-_ (Note changes or attach parent)

New X Question History: Last NRC Exam (Optional: Questions validated at the facility since 10195 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis _X_

10 CFR Part 55 Content: 55.41 m 55.43 Comments:

Question #63 The Plant is operating at full power with the 'A' and 'c' Emergency Diesel Generators danger tagged out of service when the following occurs:

An electric plant transient results in a loss of the 10700 and 10300 Busses.

RPV Low Level alarm is received and is trending down rapidly Which one of the following correctly states the required action for the above condition?

A. Maintain Rx power steady post transient B. Perform a Rapid power reduction C. Commence a normal Rx shutdown D. Immediately Scram the Rx

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO/SRO Tier # 2 Group # 2 KfA# 256000 K 6.04 Importance Rating 2.8/2.8 Proposed Question: 63 The Plant is operating at full power with the 'A' and 'C' Emergency Diesel Generators danger tagged out of service when the following occurs:

An electric plant transient results in a loss of the 10700 and 10300 Busses.

RPV Low Level alarm is received and is trending down rapidly Which one of the following correctly states the required action for the above condition?

A. Maintain Rx power steady post transient B. Perform a Rapid power reduction C. Commence a normal Rx shutdown D. Immediately Scram the Rx Proposed Answer: D, Immediately Scram the Rx

Explanation: Condensate pumps 'C' and 'A' are powered from the 10700 and 10300 Busses. Loss of these Busses results in only one Condensate\Booster Pump. RPV level lowering rapidly is given in stem. Immediate Actions of AOP-41 {Feedwater Malfunction} state: If any of the following exist:

RPV level is at or approaching designated low level benchmark then insert a manual scram.

A. incorrect-see above.

B. incorrect-see above.

C. incorrect-see above.

D. correct-see above explanation Technical Reference(s): AOP-41 Proposed references to be provided to applicants during examination:

Learning Objective: SDLP-33 1.1 Question Source: Bank #

Modified Bank # _ _ _ (Note changes or attach parent)

New Question History: Last NRC Exam (Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 m 55.43 Comments:

Question #64 The Plant is operating at 65% power with the following conditions:

Hydrogen Injection System has just been returned to service.

Annunciator 09-6-4-35 H2 ADDITION SYS TROUBLE alarms

Annunciator 09-6-1-7 OFF GAS LINE PRESS HI alarms

The Rx is manually scrammed

AOP-1 actions are complete With no further Operator actions, which one of the following correctly states the response of the Plant?

A. After 15 minutes, 01-107AOV-1 00 (Off-gas Outlet Isolation Valve) closes B. The MSIVs go closed at 850 psig C. The Turbine Bypass Valves close on low Condenser Vacuum D. 38AOV-115 (Off-gas Air Purge Valve) opens after 90 minutes

ES-401 Sample Written Examination Form ES-401*5 Question Worksheet Examination Outline Cross-

Reference:

Level RO/SRO Tier # 2 Group # ~

KfA# 271000 A 1.01 Importance Rating 3.3/3.2 Proposed Question: 64 The Plant is operating at 65% power with the following conditions:

Hydrogen Injection System has just been returned to service.

Annunciator 09-6-4-35 H2 ADDITION SYS TROUBLE alarms

Annunciator 09-6-1-7 OFF GAS LINE PRESS HI alarms

The Rx is manually scrammed

AOP-1 actions are complete With no further Operator actions, which one of the following correctly states the response of the Plant?

A. After 15 minutes, 01-107AOV-100 (Off-gas Outlet Isolation Valve) closes B. The MSIVs go closed at 850 psig C. The Turbine Bypass Valves close on low Condenser Vacuum D. 38AOV-115 (Off-gas Air Purge Valve) opens after 90 minutes Proposed Answer: C, The Turbine Bypass Valves close on low Condenser Vacuum

Explanation: Annunciators listed are symptoms of an explosion in the Air Ejector Discharge Piping which is entry to AOP-4. Vacuum SJAE Off-gas Trip system valves (38AOV-113A\B) (Le. inter stage blocking valves) auto close on either 20 psig or 300F. When they close the SJAEs are isolated from the Main Condenser and Condenser vacuum degrades rather quickly. At 8 inches Hg the TBVs auto close.

A. incorrect-see above. Also, 01-1 07AOV-1 00 closes on high radiation not pressure or temp B. incorrect-see above. Also, AOP-1 Immediate Actions are to place the Mode Switch in Shutdown therefore the 850 psig Closure is bypassed C. correct-see above explanation D. incorrect-see above. Also, 38AOV-115 closes 90 min after the SJAEs are placed in service Technical Reference(s): AOP-4 Proposed references to be provided to applicants during examination: none Learning Objective: _SDLP-38 1.09 Question Source: Bank#

Modified Bank # _ _ _ (Note changes or attach parent)

New X Question History: Last NRC Exam (Optional: Questions validated at the facility since 1019.5 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed rel/iew of evelY question.)

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis -X --

10 CFR Part 55 Content: 55.41 J§l 55.43 Comments:

Question #65 The plant is at 100% power with the following conditions:

An NPO is doing a surveillance test run on East Diesel Fire Pump 76P-4.

The NPO reports a local alarm bell and the 76P-4 "Water Temperature Signal Lighf' is ON

A plant fire occurs activating the sprinkler system

Fire Main Header pressure drops to a low of 100 psig

Among other Annunciators, "09-8-3-32 BUS 10400 NORM SUPP BKR 10402 TRIP" alarms Including the Makeup pump, how many Fire Pumps should the NPO expect to find running?

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

ES-401 Sample Written Examination Form ES*401*5 Question Worksheet Examination Outline Cross-

Reference:

LI9vei RO/SRO Tier # 2 Group # ~

KIA # 286000 K3.03 Importance Rating 3.6/3.8 Proposed Question: 65 The plant is at 100% power with the following conditions:

An NPO is doing a surveillance test run on East Diesel Fire Pump 76P-4.

The NPO reports a local alarm bell and the 76P-4 "Water Temperature Signal Light" is ON

A plant fire occurs activating the sprinkler system

Fire Main Header pressure drops to a low of 100 psig

Among other Annunciators, "09-8-3-32 BUS 10400 NORM SUPP BKR 10402 TRIP" alarms Including the Makeup pump, how many Fire Pumps should the NPO expect to find running?

A. 1 B. 2 C. 3 D. 4 Proposed Answer: B, 2

Explanation: The normal lineup is the makeup pump is always running and maintains pressure between 140-160 psig. The AC pump 76P-2 starts at 109 psig. The Diesel pump 76P-1 starts at 101 psig. The second Diesel pump 76P-4 starts at 92 psig. The makeup pump and AC pump received power from 600 VAC bus L-44 (14400). This bus is powered from 104000. The annunciator indicates a loss of the 104000 bus (since at 100% power it is powered from the Normal Supply breaker and not the Reserve). This loss removes the makeup pump and AC pump from service. Pressure dropped low enough to start the 76P-1 and the 76P-4 is already running. The alarm on 76P-4 indicates a Jacket Water Temperature of 210 deg F. However, this is not a trip, just an alarm.

A. incorrect- This would be picked if the loss of the two AC pumps IS recognized but it is thought the 101 pSig start is for the 76P-4 pump OR the P-4 pump has tripped and the P-1 pump has started.

B. correct-see above explanation C. incorrect- This would be picked if the candidate does NOT recognize the loss of AC but it is thought the 101 psig start is for the 76P-4 pump OR the P-4 pump has tripped and the P-1 pump has started.

D. incorrect- This would be picked if the candidate does NOT recognize the loss of AC and the candidate confuses the pressure settings of the system. i.e. the makeup pump is running, the second pump starts at the lower end of normal pressure of 140 pSig, the third pump starts at 109 psig and the fourth pump starts at 101 psig.

NOTE: There are other plausible combinations for picking wrong answers that combine the loss of AC with how many (if any) pumps, the possible trip of the Diesel and the start pressures of the pumps.

Technical Reference(s): OP-36 Proposed references to be provided to applicants during examination: none Learning Objective: _SDLP-76 1.03_1.05 Question Source: Bank#

Modified Bank # _ _ _ (Note changes or attach parent)

New Question History: Last NRC Exam (Optional: Questions validated at the facility since 10195 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detai/ed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis - x- -

10 CFR Part 55 Content: 55.41 m 55.43 Comments:

Question #66 The Plant has just experienced a loss of coolant accident compounded with electric plant complications.

The Control Room Supervisor (CRS) is executing numerous AOPs and EOPs.

The Shift Manager is in route to the Control Room to provide oversight.

The Operations Manager is present in the Control Room.

Which one of the following actions would be in violation of EN-OP-115 (Conduct of Operations) with regards to a Control Room Operator?

A. Coordinate NPO actions in the field B. Perform Immediate Actions without direction C. Execute direct orders from the Operations Manager D. Challenge a perceived incorrect order from the CRS

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO/SRO Tier #

Group #

KIA # 2.1.2 Importance Rating 4.1/4.4 Proposed Question: 66 The Plant has just experienced a loss of coolant accident compounded with electric plant complications.

The Control Room Supervisor (CRS) is executing numerous AOPs and EOPs.

The Shift Manager is in route to the Control Room to provide oversight.

The Operations Manager is present in the Control Room Which one of the following actions would be in violation of EN-OP-115 (Conduct of Operations) with regards to a Control Room Operator?

A. Coordinate NPO actions in the field B. Perform Immediate Actions without direction C. Execute direct orders from the Operations Manager D. Challenge a perceived incorrect order from the CRS Proposed Answer: C, Execute direct orders from the Operations Manager

Explanation: The only action not allowable per EN-OP-115 is taking direction for control manipulations from a person not having Control Room Command and Control. The CRS maintains Command and Control during a transient unless relieved from the duty. Stem did not state C&C was turned over to the Ops Manager therefore; an RO should not execute orders given by the OM.

A. incorrect-see above B. incorrect-see above C. correct-see above explanation D. incorrect-see above Technical Reference(s): EN-OP-115 Proposed references to be provided to applicants during examination: none Learning Objective: _LP-AP 46.04_

Question Source: Bank#

Modified Bank # _ _ _ (Note changes or attach parent)

New Question History: Last NRC Exam (Optional: Questions validated at the facility since 10195 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detai/ed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge _x_

Comprehension or Analysis 10 CFR Part 55 Content: 55.41 l1Ql 55.43 Comments:

Question #67 The Plant is operating at 72% performing a Control Rod Sequence Exchange.

In accordance with EN-OP-115 (Conduct of Operations). which one of the following choices correctly lists an authorized individual who can verify Control Rod selection and movement?

A. Inactive licensed FSS B. License candidate RO C. Shift Technical Advisor D. Qualified Rx Engineer

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO/SRO Tier # -L Group # .JL KJA# 2.1.1 Importance Rating 3.8/4.2 Proposed Question: 67 The Plant is operating at 72% performing a Control Rod Sequence Exchange.

In accordance with EN-OP-115 (Conduct of Operations), which one of the following choices correctly lists an authorized individual who can verify Control Rod selection and movement?

A. Inactive licensed FSS B. License candidate RO C. Shift Technical Advisor D. Qualified Rx Engineer Proposed Answer: C, STA

Explanation: EN-OP-115: A second individual (licensed operator or Shift Technical Advisor) with no concurrent activities should be present during manual rod insertion or withdrawal. The second individual should:

Verify operator selected correct control rod prior to rod movement

Verify proper control rod movement by observing diverse indications A. incorrect-see above B. incorrect-see above C. correct-see above explanation D. incorrect-see above Technical Reference( s): EN-OP-115 Proposed references to be provided to applicants during examination: none Learning Objective: - LP-AP 46.04-Question Source: Bank #

Modified Bank # _ _ _ (Note changes or attach parent)

New X Question History: Last NRC Exam (Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis 10 CFR Part 55 Content: 55.41 J.1Q}

55.43 Comments:

Question #68 The Plant is operating at 65% power with EPIC out of service due to planned maintenance.

HPCI automatically starts and begins to inject into the RPV Which one of the following correctly states the minimum indications to check prior to tripping HPCI?

A. Verify two level indicators on the 09-5 panel read greater than 126.5 inches B. Verify two Drywell pressure indicators on the 09-3 panel read less than 2.7 psig C. Verify one level indicator on the 09-5 panel reads greater than 126.5 inches AND verify one Drywell pressure indicator on the 09-3 panel reads less than 2.7 psig D. Verify two level indicators on the 09-5 panel read greater than 126.5 inches AND verify two Drywell pressure indicators on the 09-3 panel read less than 2.7 psig

ES-401 Sample Written Examination Form ES*401*5 Question Worksheet Examination Outline Cross-

Reference:

Lever RO/SRO Tier # ~

Group # .lL KIA # 2.1.45 Importance Rating 4.3/4.3 Proposed Question: 68 The Plant is operating at 65% power with EPIC out of service due to planned maintenance.

HPCI automatically starts and begins to inject into the RPV Which one of the following correctly states the minimum indications to check prior to tripping HPCI?

A. Verify two level indicators on the 09-5 panel read greater than 126.5 inches B. Verify two Drywell pressure indicators on the 09-3 panel read less than 2.7 psig C. Verify one level indicator on the 09-5 panel reads greater than 126.5 inches AND verify one Drywell pressure indicator on the 09-3 panel reads less than 2.7 psig D. Verify two level indicators on the 09-5 panel read greater than 126.5 inches AND verify two Drywell pressure indicators on the 09-3 panel read less than 2.7 psig Proposed Answer: D, Verify two level indicators on the 09-5 panel read greater than 126.5 inches AND verify two Drywell pressure indicators on the 09-3 panel read less than 2.7 psig then trip HPCI

Explanation: Overriding an automatic initiation of a safety function shall not be performed unless one of the following conditions exists:

Adequate core cooling is assured by at least two independent indications.

Misoperation in automatic mode is confirmed by at least two independent indications The auto initiation logic scheme for HPCI comprises of an arrangement of RPV level and DW pressure contacts. In order to be assured that the initiation signal is not valid, two independent checks of the same parameter should be validated.

A. incorrect-see above B. incorrect-see above C. incorrect-see above D. correct-see above explanation Technical Reference(s): EN-OP-115, EP-1 Proposed references to be provided to applicants during examination: none Learning Objective: __ ~ILP-LOI-TBDI03 1.02 Question Source: Bank#

Modified Bank # _ _ _ (Note changes or attach parent)

New x Question History: Last NRC Exam (Optional: Questions validated at the facility since 10195 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis 10 CFR Part 55 Content: 55.41 m 55.43 Comments:

Question #69 The Plant is in Mode 4 preparing to startup Reactor Recirculation Pump (RWR) pump 'A'.

While executing Section D of OP-27 (Recirculation System), the Operator recognizes the following two discrepancies with the procedure:

Ensure open 02-RWP-38B (RWR pump a seal purge dnstr isol valve)

The Component ID should read 02-RWR-38A.

Which one of the following correctly states most simple allowable method to fix the Component ID discrepancy?

Initiate a (an) ...

A. Editorial Correction Form B. Temporary Operating Procedure C. Limited Revision to OP-27 D. Full Revision to OP-27

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO/SRO Tier # 3 Group # .JL KJA#

Importance Rating the Tor Proposed Question: 69 The Plant is in Mode 4 preparing to startup Reactor Recirculation Pump (RWR) pump 'A'.

While executing Section D of OP-27 (Recirculation System), the Operator recognizes the following two discrepancies with the procedure:

Ensure open 02-RWP-38B (RWR pump a seal purge dnstr isol valve)

The Component ID should read 02-RWR-38A Which one of the following correctly states most simple allowable method to fix the Component ID discrepancy?

Initiate a (an) ...

A. Editorial Correction Form B. Temporary Operating Procedure C. Limited Revision to OP-27 D. Full Revision to OP-27 Proposed Answer: A, Editorial Correction Form

Explanation: AP-02.04 (Control of Procedures) Attachment 4 (Editorial Correction Form) is the appropriate method to correct Equipment number or titles. Its purpose is to correct typographical errors or add clarification to procedures without changing the technical content or intent of the procedure.

A. correct-see above explanation B. incorrect-TOP would be used if no procedure exists to address a situation or condition C. incorrect-Limited Rev is outside of its intended use D. incorrect-Full Rev is outside of its intended use Technical Reference(s): AP-02.04 Proposed references to be provided to applicants during examination: none Learning Objective: __LP-AP 4.02 Question Source: Bank #

Modified Bank # _ _ _ (Note changes or attach parent)

New x Question History: Last NRC Exam (Optional: Questions validated at the facility since 10195 will generally undergo less rigorous review by the NRC; faifure to provide the information will necessitate a detaifed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge _x_

Comprehension or Analysis 10 CFR Part 55 Content: 55.41 J.1Q) 55.43 Comments:

Question #70 Given the following information:

The Mode Switch is in Refuel

One Reactor Vessel Head Closure bolt has just been de-tensioned.

RPV temperature is 212F Which one of the following correctly states the Mode the Reactor is now in?

A. 2 (Startup)

B. 3 (Hot Shutdown)

C. 4 (Cold Shutdown)

D. 5 (Refuel)

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO/SRO Tier #

Group #

KIA #

Importance Rating Proposed Question: 70 Given the following information:

The Mode Switch is in Refuel

One Reactor Vessel Head Closure bolt has just been de-tensioned.

RPV temperature is 212F Which one of the following correctly states the Mode the Reactor is now in?

A. 2 (Startup)

B. 3 (Hot Shutdown)

C. 4 (Cold Shutdown)

D. 5 (Refuel)

Proposed Answer: D, 5 Refuel

Explanation: Tech Spec Definitions Table 1.1-1 states the Rx is in Refuel Mode (5) when one or more RPV head closure bolts are less than fully tensioned with the Mode Switch is in Refuel no matter what RPV temperature is.

A. incorrect-plausible should candidate not understand the definitions of Modes B. incorrect-plausible should candidate not understand the definitions of Modes C. incorrect-plausible should candidate not understand the definitions of Modes D. correct-see above explanation Technical Reference(s): TS 1.1 Proposed references to be provided to applicants during examination: none Learning Objective: JLP-OPS-ITS02 1.03 Question Source: Bank#

Modified Bank # _ _ _ (Note changes or attach parent)

New X Question History: Last NRC Exam (Optional: Questions validated at the facility since 10195 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 m 55.43 Comments:

Question #71 The Plant has experienced a Loss of Coolant Accident.

The Shift Manager has ordered the Primary Containment to be vented per EP-6, (Post Accident Containment Venting and Gas Control).

Which one of the following correctly states Qreferred vent path and the reason why?

preferred path reason why A. Torus maintain nitrogen in Drywell B. Drywell maintain nitrogen in Torus C. Drywell lower moisture content prior to reaching SBGT D. Torus "scrub" radio-nuclides prior to reaching SBGT

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO/SRO Tier #

Group #

KJA# 2.3.11 Importance Rating 3.8/4.3 Control:

Proposed Question: 71 The Plant has experienced a Loss of Coolant Accident.

The Shift Manager has ordered the Primary Containment to be vented per EP-6, (Post Accident Containment Venting and Gas Control).

Which one of the following correctly states prE~ferred vent path and the reason why?

preferred path reason why A. Torus maintain nitrogen in Drywell B. Drywell maintain nitrogen in Torus C. Drywell lower moisture content prior to reaching SBGT D. Torus "scrub" radio-nuclides prior to reaching SBGT Proposed Answer: D, Torus "scrub" radio-nuclides prior to reaching SBGT

Explanation: EP-6 states: The preferred vent/purge flow path is to vent from the torus and purge through the drywell. This flow path has the following advantages:

Maintains pressure suppression function

Scrubs gases before release

Ensures effective vent and purge of both drywell and torus.

A. incorrect-plausible should candidate believe nitrogen should be kept in Containment to minimize H2\02 production B. incorrect- plausible should candidate believe nitrogen should be kept in Containment to minimize H2\02 production C. incorrect- plausible should candidate believe drier air would be in DW and to protect charcoal in SBGT from getting wet D. correct-see above explanation Technical Reference(s): EP-6 Proposed references to be provided to applicants during examination: none Learning Objective: MIT-301.lIE 4.05_ _

Question Source: Bank #

Modified Bank # _ _ _ (Note changes or attach parent)

New Question History: Last NRC Exam (Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis 10 CFR Part 55 Content: 55.41 J..11j 55.43 Comments:

Question #72 The Plant is operating at full power with the ISFSI Fuel Cask being loaded.

The Bridge Operator informs the Control Room that a fuel bundle has just dropped.

The Refuel Floor ARM and RB Ventilation Rad Monitors show !!.Q change in value.

Which one of the following correctly states the action(s) to be taken?

A. Re-grapple the fuel bundle and return it to the Spent Fuel Pool rack B. Re-grapple the fuel bundle and complete the fuel move to the Cask C. Stop further fuel moves and evacuate the Refuel Bridge only D. Stop further fuel moves and evacuate the Refuel Floor and Reactor Bldg

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO/SRO Tier# ~

Group #

KIA # 2.3.13 Importance Rating 3.4/3.8 Proposed Question: 72 The Plant is operating at full power with the ISFSI Fuel Cask being loaded.

The Bridge Operator informs the Control Room that a fuel bundle has just dropped.

The Refuel Floor ARM and RB Ventilation Rad Monitors show!!Q change in value.

Which one of the following correctly states the action(s) to be taken?

A. Re-grapple the fuel bundle and return it to the Spent Fuel Pool rack B. Re-grapple the fuel bundle and complete the fuel move to the Cask C. Stop further fuel moves and evacuate the Refuel Bridge only D. Stop further fuel moves and evacuate the Refuel Floor and Reactor Bldg Proposed Answer: D, Stop further fuel moves and evacuate the Refuel Floor and Reactor Bldg

Explanation: Dropped fuel assembly in the SFP is entry to AOP-44 (Dropped Fuel Assembly)

Immediate operator actions, regardless of radiological conditions, require stopping further fuel moves and evacuation of the Refuel Floor and Reactor Bldg A. incorrect-plausible should candidate believe it is acceptable to pick-up and move a dropped fuel assembly.

B. incorrect- plausible should candidate believe it is acceptable to pick-up and move a dropped fuel assembly.

C. incorrect- plausible should candidate believe it is acceptable to pick-up and move a dropped fuel assembly.

D. correct-see above explanation Technical Reference(s): AOP-44 Proposed references to be provided to applicants during examination: none Learning Objective: SDLP-08A 1.14 Question Source: Bank#

Modified Bank # _ _ _ (Note changes or attach parent)

New X Question History: Last NRC Exam (Optional: Questions validated at the facility since 10/96 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of eve!}' question.)

Question Cognitive Level: Memory or Fundamental Knowledge _x_

Comprehension or Analysis 10 CFR Part 55 Content: 55.41 Jjl) 55.43 Comments:

Question #73 The Plant is operating at full power when a serious fire occurs in the RB West Crescent area.

The Fire Protection Panel indicates ionization and actuation of fire suppression systems.

Numerous unexplained EPIC alarms and Annunciators are received in the Control Room.

Which one of the following correctly states the action(s) to be taken?

A. Maintain Rx power steady, dispatch Fire Brigade B. Perform a rapid power reduction to 55% core flow C. Scram the Rx, trip both Recirc Pumps D. Commence a normal Rx shutdown

ES*401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level ROISRO Tier # ...L Group # .JL KIA # 2.4.27 Importance Rating 3.4/3.9 Proposed Question: 73 The Plant is operating at full power when a serious fire occurs in the RB West Crescent area.

The Fire Protection Panel indicates ionization and actuation of fire suppression systems.

Numerous unexplained EPIC alarms and Annunciators are received in the Control Room.

Which one of the following correctly states the action{s) to be taken?

A. Maintain Rx power steady, dispatch Fire Brigade B. Perform a rapid power reduction to 55% core flow C. Scram the Rx, trip both Recirc Pumps D. Commence a normal Rx shutdown Proposed Answer: C, Scram the Rx, trip both Recirc Pumps

Explanation: Utilizing AOP-28 (Operation During Plant Fires) Flowchart 1, the given information directs entry into AOP-28 Section E. Section E directs performing Section 1 of applicable attachment. Attachment 2 (RB West Crescent) immediate actions direct the Rx to be manually scrammed and RWR pump tripped.

A. incorrect-plausible should candidate believe a plant transient should not be introduced during a fire due to one Control Room Operator has to be dispatched to fight the fire.

B. incorrect-plausible should candidate believe Rx power should be lowered before inserting a scram.

C. correct-see above explanation D. incorrect-plausible should candidate believe conditions warrant the Rx be shutdown normally vice scrammed Technical Reference(s): AOP-28 (Operation During Plant Fires)

Proposed references to be provided to applicants during examination: none Learning Objective: SDLP-76 '1.15 Question Source: Bank#

Modified Bank # _ _ _ (Note changes or attach parent)

New X Question History: Last NRC Exam (Optional: Questions validated at the facility since 10195 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge _X_

Comprehension or Analysis 10 CFR Part 55 Content: 55.41 l1Q) 55.43 Comments:

Question #74 The Plant is operating at full power when the following occurs:

09-3-2-40 "RX BLDG VENT RAD MON HI HI" alarms.

RP confirms the alarm is valid and value is 2x10E4 cpm.

Which one of the following correctly states applicable procedures to enter and execute?

A. ARP 09-3-2-40 only B. ARP 09-3-2-40 and OP-20 (Standby Gas Treatment System) only C. ARP 09-3-2-40, OP-20 (Standby Gas Treatment System) and AOP-15 (Isolation Verification and Recovery) only D. ARP 09-3-2-40, OP-20 (Standby Gas Treatment System). AOP-15 (Isolation Verification and Recovery) and EOP-5 (Secondary Containment Control)

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO/SRO Tier # -L Group # .JL KJA# 2.4.5 Importance Rating 3.7/4.3 Proposed Question: 74 The Plant is operating at full power when the following occurs:

09-3-2-40 "RX BLDG VENT RAD MON HI HI" alarms.

RP confirms the alarm is valid and value is 2x10E4 cpm.

Which one of the following correctly states applicable procedures to enter and execute?

A. ARP 09-3-2-40 only B. ARP 09-3-2-40 and OP-20 (Standby Gas Treatment System) only C. ARP 09-3-2-40 and AOP-15 (Isolation Verification and Recovery) only D. ARP 09-3-2-40, OP-20 (Standby Gas Treatment System), AOP-15 (Isolation Verification and Recovery) and EOP-5 (Secondary Containment Control)

Proposed Answer: D, ARP 09-3-2-40, OP-20 (Standby Gas Treatment System), AOP-15 (Isolation Verification and Recovery) and EOP-5 (Secondary Containment Control)

Explanation (Optional):

RB Ventilation value of 1x1 OE4 cpm results in the following:

Annunciator

RB Vent isolation

SBGT auto start

EOP-5 entry condition A. incorrect-not the only procedure to enter B. incorrect-OP-20 will be entered due to SBGT auto starting but distractor is not complete C. incorrect-AOP-15 will be entered due to RB Vent isolation but distractor is not complete D. correct-all four procedures should be entered and executed Technical Reference(s): EP-1 Proposed references to be provided to applicants during examination: none Learning Objective: SDLP-66A 1.14_ __

Question Source: Bank #

Modified Bank # _ _ _ (Note changes or attach parent)

New x Question History: Last NRC Exam (Optional: Questions validated at the facility since 10/9.5 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of eveIY question.)

Question Cognitive Level: Memory or Fundamental Knowledge _x_

Comprehension or Analysis 10 CFR Part 55 Content: 55.41 l1.Q.)

55.43 Comments:

Question #75 The Plant was operating at 92% power when a loss of Feedwater heating occurs.

As actions were being taken to mitigate the event, RWR Pump 'B' tripped.

The following conditions now exist:

APRM average power = 60% (varying between 54-66%)

Core Flow = 30Mlb\hr

Annunciator 09-5-2-44 APRM UPSCALE is intermittently in alarm Based on the above, which one of the following correctly states the required Operator action?

A. Raise RWR Flow B. Insert Control Rods C. Select SLO setpoints D. Manually scram the Rx

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO/SRO Tier #

Group #

KIA # 2.4.49 Importance Rating 4.6 I 4.4 Proposed Question: 75 The Plant was operating at 92% power when a loss of Feedwater heating occurs.

As actions were being taken to mitigate the event, RWR Pump 'B' tripped.

The following conditions now exist:

APRM average power = 60% (varying between 54-66%)

Core Flow = 30Mlb\hr

Annunciator 09-5-2-44 APRM UPSCALE is intermittently in alarm Based on the above, which one of the following correctly states the required Operator action?

A. Raise RWR Flow B. Insert Control Rods C. Select SLO setpoints O. Manually scram the Rx Proposed Answer: 0, Manually scram the Rx

Explanation (Optional):

Given indications are symptoms of THI. Scram required.

A. incorrect-B. incorrect-C. incorrect-D. correct-Technical Reference(s): _ _OP-16, '-'A.=.O-'-P.....;:-S"--__

Proposed references to be provided to applicants during examination:

Learning Objective: SDLP- (As available)

Question Source: Bank#

Modified Bank # _ _ _ (Note changes or attach parent)

New x Question History: Last NRC Exam (Optional: Questions validated at the facility since 10195 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis 10 CFR Part 55 Content: 55.41 .J.1Ql 55.43 Comments:

Question #76 A LOCA and reactor scram has occurred with the following:

One control rod is stuck at position 40

All other control rods are fully inserted

RPV pressure is 750 psig

All high pressure injection capability has been lost

RPV level is +10" and lowering slowly

Off-Site power has been lost The CRS should be implementing:

A. EOP-3, Failure to Scram.

B. The Alternate RPV Level Control leg of EOP-2.

C. The Emergency RPV Depressurization leg of EOP-2.

D. EOP-7, RPV Flooding.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level SRO Tier# _1_

Group # _1_

KIA # 295031 EA2.04 Importance Rating 4.8 Proposed Question: 76 A LOCA and reactor scram has occurred with the following:

One control rod is stuck at position 40

All other control rods are fully inserted

RPV pressure is 750 psig

All high pressure injection capability has been lost

RPV level is +10" and lowering slowly

Off-Site power has been lost The CRS should be implementing:

A. EOP-3, Failure to Scram B. The Alternate RPV Level Control leg of EOP-2.

C. The Emergency RPV Depressurization leg of EOP-2 D. EOP-7, RPV Flooding.

Answer: B Explanation:

A. This answer is incorrect because the Override criteria to enter EOP-3 are not met.

With only one stuck rod, no ATWS condition exists. There is one rod that is withdrawn past the indicated position (I.e not all rods not at 02) but it CAN be determined that the reactor will remain shutdown in all cases without boron. Thus EOP-3 is not entered.

B. This answer is correct because with RPV level cannot be maintained above 0" and there is an injection source available. The Alternate Level Control leg is required.

C. This answer is incorrect because additional steps need to be taken with Alternate Level Control (EOP-2) before entering Emergency Depressurization. Additional means of restoring level need to be attempted prior to RPV level decreasing below -19".

D. This answer is incorrect because the RPV level is known and none of the conditions provided call the validity of level indication into question. (Caution #1 in EOP-2)

A. incorrect- see above B. correct-C. incorrect- see above D. incorrect- see above Technical Reference(s): -

EOP -

Proposed references to be provided to applicants during examination: none Learning Objective: _----!.!M~IT.!..-..:::3~01.!..:...!..11.!..:C~1c.:..:.O~3~1..!..1.:,.:::O..:::,5_1.:...:,.0;::,27'--_ _ _ _ (As available)

Question Source: Bank#

Modified Bank # _ _ _ (Note changes or attach parent)

New Question History: Last NRC Exam Grand Gulf 2009-:-_ _

(Optional: Questions validated at the facility since 10195 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of evety question.)

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 55.43 _(5L Comments:

Question #77 The Plant is in Mode 4 with the following conditions:

RHR Pump 'A' has just been placed in Shutdown Cooling.

RHR Loop 'B' is unavailable and inoperable

"A" RWR pump is running

RPV temperature is 200 'F The following occurs:

A pipe crack develops, upstream of 10MOV-15A, RHR SOC Suction Isolation Valve

"A" RHR pump has been manually tripped

West Crescent Area water level is 19 inches and rising slowly

The leak has not been isolated

Rx Level: +175" and steady Which one of the below completes the following statement:

EOP-2 entry _~'--_ _ and the "A" RWR pump should be ~~=-_ _

ill ill A. Is required Left in service lAW AOP-30 B. Is NOT required Tripped so the leak may be isolated C. Is required Tripped so the leak may be isolated O. Is NOT required Left in service lAW AOP-30

ES*401 Sample Written Examination Form ES*401*5 Question Worksheet Examination Outline Cross-

Reference:

Level SRO Tier# _1_

Group # _1_

KIA # 295021 AA2.07 Importance Rating 3.1 Proposed Question: 77 The Plant is in Mode 4 with the following conditions:

RHR Pump 'A' has just been placed in Shutdown Cooling.

RHR Loop 'B' is unavailable and inoperable

"An RWR pump is running

RPV temperature is 200 'F The following occurs:

A pipe crack develops, upstream of 10MOV-15A, RHR SDC Suction Isolation Valve

"A" RHR pump has been manually tripped

West Crescent Area water level is 19 inches and rising slowly

The leak has not been isolated

Rx Level: +175" and steady Which one of the below completes the following statement:

EOP-2 entry _ .......(...:,.1)<--_ _ and the "A" RWR pump should be _-->.:(2::..)'--__

ill W A. Is required Left in service lAW AOP-30 B. Is NOT required Tripped so the leak may be isolated C. Is required Tripped so the leak may be isolated D. Is NOT required Left in service lAW AOP-30

Proposed Answer: D, Is NOT required; Left in service lAW AOP-30 Explanation: Loss of SDC is entry to AOP-30. Crescent Area level> 0 inches is an EOP-5 entry.

With a primary system discharging into the area and crescent level> 18", EOP-2 entry is required (a branching step within EOP-5). However, the candidate may incorrectly think EOP-2 entry is required <177 inches but:

EP-1 states: This procedure applies during all plant operating modes, except when reactor coolant temperature is less than 212 deg F and a reactor startup or shutdown is not in progress. The candidate must recognize stem parameters of Mode 4 and 200 deg F.

AOP-30 states: IF shutdown cooling cannot be restored within one hour, OR reactor coolant temperature EXCEEDS 1501=, THEN ensure one or both RWR pumps are running per Section D of OP-27. It also has requirements that, if a RWR pump is running, to monitor inlet loop temperatures A. incorrect- see above B. incorrect- see above C. incorrect- see above D. correct- see above Technical Reference(s):

Proposed references to be provided to applicants during examination: none Learning Objective: _--"L:.!...P....!-A...!.:O~P!...-...!1.,:.: :.0:.: :;3_ _ _ _ _ _ (As available)

Question Source: Bank#

Modified Bank # _ _ _ (Note changes or attach parent)

New Question History: Last NRC Exam (Optional: Questions validated at the facility since 10195 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 (5)

Comments:

Question #78 The Plant is operating at full power when the following events occur:

Annunciator 09-5-1-22 (RPS HI RX PRESS TRIP) alarms

RPS A & B SCRAM GROUPS white li~lhts are illuminated

ATC reports:

o Pressure Indicator 06PI-90A reads 1080 psig and steady o Pressure Indicator 06PI-90B reads 1080 psig and steady o Pressure Indicator 06PI-90C reads 1040 psig and steady o Pressure Indicator 06PI-90D reads 1080 psig and steady Which one of the following correctly lists the required Operator action, Emergency Action Level (EAL) and NRC maximum notification time limit for the above conditions?

(Note: It may assumed that the below Operator actions can be completed successfully.)

Action EAL Notification time A. Rapid power reduction Alert Immediate B. Manual scram Site Area Immediate C. Manual scram Alert 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months D. Rapid power reduction Site Area 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months

Sample Written Examination Question Worksheet Examination Outline Cross-

Reference:

Level SRO Tier #

Group #

KJA# 295025 G 2.1.7 Importance Rating 4.7 Proposed Question: 78 The Plant is operating at full power when the following events occur:

Annunciator 09-5-1-22 (RPS HI RX PRESS TRIP) alarms

RPS A & B SCRAM GROUPS white lights are illuminated

A TC reports:

o Pressure Indicator 06PI-90A reads 1080 psig and steady o Pressure Indicator 06PI-90B reads 1080 psig and steady o Pressure Indicator 06PI-90C reads 1040 psig and steady o Pressure Indicator 06PI-90D reads 1080 psig and steady Which one of the following correctly lists the required Operator action, Emergency Action Level (EAL) and NRC maximum notification time limit for the above conditions?

Note: It may assumed that the below Operator actions can be completed successfully.

Action Reportability time A. Rapid power reduction Alert Immediate B. Manual scram Site Area Immediate C. Manual scram Alert 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months D. Rapid power reduction Site Area 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months

Proposed Answer: C, Manual scram, Alert, 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Explanation: Annunciator 09-5-1-22 (RPS HI RX PRESS TRIP) is an automatic Rx scram signal (even with Pressure Indicator 06PI-90C reading 1040 psig and steady, one detector failure should not prevent the auto scram signal). White RPS lights illuminated is indicative of a failure to scram.

A rapid power reduction would lower power however a manual scram is required. An automatic failure of RPS is an EAL call of Alert and the NRC notification within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months is required.

A. incorrect - rapid power reduction is plausible because AOP-32 (Unanticipated Reactivity Change) and AOP-36 (Stuck Open Relief Valves) call for reducing power per Rap 7.3.16 which includes a Rapid Power Reduction. An positive reactivity change occurred due to the pressure increase and ali SRVs would be open at 1045 psig.

B. incorrect- see above C. correct - see above D. incorrect- see above Technical Reference(s): _ARP-09-5-1-22 Proposed references to be provided to applicants during examination: IAP-2 Learning Objective: TS 3.3.1.1 SDLP-02D 1.07,1.18 SDLP-051.07.1.14 JLP-OPS-ITS02 1.01,1.05 (As available)

Question Source: Bank #

Modified Bank # _ _ _ (Note changes or attach parent)

New Question History: Last NRC Exam (Optional: Questions validated at the facility since 10195 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed rel/iew of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 (6)

Comments:

Question #79 The Plant has experienced a seismic event with a Primary leak outside of Containment.

The following conditions are present:

HPCI and RCIC are the only injection sources currently injecting into the RPV

All RHR and Core Spray pumps are running and available for injection

RPV level (-)5 inches and lowering

RPV pressure is 700 psig and lowerinq

Torus level 9.5 feet and lowering

No Operator actions have been taken Which one of the following correctly states the action(s) that are required to be performed?

A. HPCI and RCIC are required to bl3 tripped due to loss of turbine steam discharge condensation ability B. Only HPCI is required to be tripped due to loss of turbine steam discharge condensation ability C. HPCI and RCIC are required to be left running and injecting due to lowering RPV water level D. Only HPCI is required to be left running and injecting due to lowering RPV water level

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level SRO Tier#

Group #

KIA # 295030 EA2.03 Importance Rating 3.9 Proposed Question: 79 The Plant has experienced a seismic event with a Primary leak outside of Containment.

The following conditions are present:

HPCI and RCIC are the only injection sources currently injecting into the RPV

All RHR and Core Spray pumps are running and available for injection

RPV level (-)5 inches and lowering

RPV pressure is 700 psig and lowering

Torus level 9.5 feet and lowering

No Operator actions have been taken Which one of the following correctly states the action(s) that are required to be performed?

A. HPCI and RCIC are required to be tripped due to loss of turbine steam discharge condensation ability B. Only HPCI is required to be tripped due to loss of turbine steam discharge condensation ability C. HPCI and RCIC are required to be left running and injecting due to lowering RPV water level D. Only HPCI is required to be left running and injecting due to lowering RPV water level Proposed Answer: B, Only HPCI is required to be tripped due to loss of turbine steam discharge condensation ability

Explanation: This question tests the Candidates knowledge of Primary Containment Control and RPV level strategy prioritization. EOP-4 bases states that HPCI should be tripped, irrespective of whether the core will be adequately cooled, if Torus level cannot be maintained> 10.75 feet due to the HPCI steam exhausting directly into the Torus air space will pressurize the Torus and lead to Containment failure. RCIC does not have this requirement due to the size \ volume of RCIC steam discharge. An RPV level on < 0 inches means that adequate core cooling does not exist so, although HPCI is adding water to the core, protection of Containment takes precedence. At 700 psig, RPV pressure is too high for RHR\CS to assist in Level control.

A. incorrect- see above B. correct- see above C. incorrect- see above D. incorrect- see above Technical Reference(s): _EOP-4_ __

Proposed references to be provided to applicants during examination: none Learning Objective: _--!.!M:!!.IT.!....-..:::3~01.:.:.... :. 1'~IE=-4...!.:.~0:::..5_ _ _ _ _ (As available)

Question Source: Bank #

Modified Bank #: _-,-:-_ (Note changes or attach parent)

New X Question History: Last NRC Exam (Optional; Questions validated at the facility since 10196 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detai/ed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 (5)

Comments:

Question #80 The Plant has experienced a Seismic Event resulting in damage to the Reactor Building.

The following conditions exist:

RPV status

Four Control Rods are stuck at position 02

RPV level is (-)10 inches and steady

RCIC is the only available injection source

RPV pressure is 650 psig and lowering

RCIC supply steam is escaping thru a Reactor Building breach Containment status

Drywell pressure is 0 psig

Drywell dose rates are 300 R\hr

Dose projections at the Site Boundary are at the Site Area Emergency level and rising Based on the above conditions, which one of the following correctly lists the action(s) and governing procedure that is required to be implemented?

A. Isolate RCIC in accordance with EOP-6.

B. Perform an Emergency Depressurization as directed by EOP-6.

C. Perform an Emergency Depressurization as directed by EOP-3a.

D. Flood the Primary Containment in accordance with the SAOGs.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level SRO Tier # 1 Group # _1_

KJA# 295038 G2.4.4 Importance Rating 4.7 Proposed Question: 80 The Plant has experienced a Seismic Event rElsulting in damage to the Reactor Building.

The following conditions exist:

RPV status

Four Control Rods are stuck at position 02

RPV level is (-)10 inches and steady

RCIC is the only available injection source

RPV pressure is 650 psig and lowering

RCIC supply steam is escaping thru a Reactor Building breach Containment status

Drywell pressure is 0 psig

Drywell dose rates are 300 R\hr

Dose projections at the Site Boundary are at the Site Area Emergency level and rising Based on the above conditions, which one of the following correctly lists the action(s) and governing procedure that is required to be implemented?

A. Isolate RCIC in accordance with EOP-6.

B. Perform an Emergency Depressurization as directed by EOP-6.

C. Perform an Emergency Depressurization as directed by EOP-3a.

D. Flood the Primary Containment in accordance with the SAOGs.

Proposed Answer: B, Perform an Emergency Depressurization as directed by EOP-6.

Explanation: This question evaluates a Candidates knowledge of EOP-6 (Radioactivity Release Control) entry and mitigating actions to protect the Core and\or the Public. Conditions given are the E-Plan Alert level therefore EOP-6 is entered. EOP-6 contains two statements that concern a Primary system leaking outside of Containment: (1): Isolate all primary systems that are discharging into areas outside the primary and secondary containments except systems required to be operated by the EOPs and (2): IF a primary system is discharging into an area outside of the primary and secondary containments THEN ED is required. Enter EOP-2 and execute concurrently. RCIC leak is a Pri system leaking outside secondary Containment however; RCIC is the only injection source of water and is required per EOP-2. ED per EOP-3a is plausible should Candidate believe position 02 warrants entry to EOP-3 (it does not). Conditions given also indicate minor Core Damage has \ is occurring but not at 3000R\hr criteria. Containment Flooding is not yet warranted with given conditions although plausible should Candidate believe RCS integrity cannot be relied upon to protect fuel from further damage.

A. incorrecl- see above B. correct- see above C. incorrect- see above D. incorrect- see above Technical Reference(s):

Proposed references to be provided to applicants during examination: none Learning Objective: MIT-301.11G 6.03_ _ (As available)

Question Source: Bank#

Modified Bank # _ _ _ (Note changes or attach parent)

New Question History: Last NRC Exam (Optional: Questions validated at the facility since 10195 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge ComprehenSion or Analysis X 10 CFR Part 55 Content: 55.41 55.43 (2)

Comments:

Question #81 The Plant is operating at full power with EDGs 'A' and 'B' inoperable due to a common cause failure.

The following events take place:

Line 3 and Line 4 ( 115KV line) have become inoperable due to severe weather.

Unidentified Drywell leakage has risen to 7 gpm.

Drywell pressure is 1.9 psig and rising very slowly.

Which one of the following correctly states the required actions to be taken for the above conditions?

A. Place a vent on the Torus and continue Plant operation for a maximum of 7 days.

B. Place HPCIIRCIC in Pressure Control Mode, enter LCO 3.0.3 and lower Rx coolant temperature to ~ 212'F within 37 hours4.282407e-4 days 0.0103 hours 6.117725e-5 weeks 1.40785e-5 months .

C. Secure all non-essential electrical loads then lower and maintain Rx power to 50-70 MWe until EDG 'A' and 'C' are returned to service.

D. Start \ load EDGs 'C' and 'D', place RHR in Torus cooling mode and maintain continuous EDG loading ~ 3050 KW.

ES-401 Sample Written Examination Form ES*401*5 Question Worksheet Examination Outline Cross-

Reference:

Level SRO Tier #

Group # _1_

KlA# 295003 AA2.04 Importance Rating 3.7 Proposed Question: 81 The Plant is operating at full power with EDGs 'A' and 'B' inoperable due to a common cause failure.

The following events take place:

Line 3 and Line 4 (115KV line) have become inoperable due to severe weather.

Unidentified Dryweilleakage has risen to 7 gpm.

Drywell pressure is 1.9 pSig and rising very slowly.

Which one of the following correctly states the required actions to be taken for the above conditions?

A. Place a vent on the Torus and continue Plant operation for a maximum of 7 days.

B. Place HPCI/RCIC in Pressure Control Mode, enter LCO 3.0.3 and lower Rx coolant temperature to ~ 212'F within 37 hours4.282407e-4 days 0.0103 hours 6.117725e-5 weeks 1.40785e-5 months .

C. Secure all non-essential electrical loads then lower and maintain Rx power to 50-70 MWe until EDG 'A' and 'C' are returned to service.

D. Start \ load EDGs 'C' and 'D', place RHR in Torus cooling mode and maintain continuous EDG loading ~ 3050 KW.

Proposed Answer: B, Place HPCIIRCIC in Pressure Control Mode, enter LCO 3.0.3 and lower Rx coolant temperature to ~ 212'F within 37 hours4.282407e-4 days 0.0103 hours 6.117725e-5 weeks 1.40785e-5 months

Explanation: This question evaluates a Candidates knowledge on what to do should a Plant shutdown be required with reduced electrical power sources. Both EDGs 'A' and 'B' (one subsystem) inoperable is a 14 day LCO 3.8.1.B. Two offsite 115kv circuit inoperable now results in 4 AC power sources inoperable. Which, from TS 3.8.1.G directs LCO 3.0.3 immediately (a 7 day LCO exists if just the two off~site power sources were inoperable). LCO 3.0.3 requires Mode 4 (.:::

212F) within 37 hours4.282407e-4 days 0.0103 hours 6.117725e-5 weeks 1.40785e-5 months . Additionally, a leak in the DW requires the Rx to be shutdown. A shutdown without Offsite power is a unique situation addressed in AOP-72 (115kv loss). AOP-72 provides guidance on how to shutdown the Rx w\o offsite power. In addition to lowering power to 50~75 MWe, HPCI and RCIC are to be placed in Pressure Control Mode, then scram the Rx and close the MSIVs. Not only will the Rx have to be shutdown but TS 3.0.3 has to be met. Plant cooldown with HPCI and RCIC in Pressure Control Mode will have to be utilized to meet the LCO.

A. incorrect~ see above B. correct- see above C. incorrect- see above D. incorrect- see above Technical Reference(s): __TS 2i.8.1 and TS 3.0.3 and AOP-72 and SOP~39~

Proposed references to be provided to applicants during examination: TS 3.8.1 Learning Objective: (As available)

Question Source: Bank#

Modified Bank # _ _ _ (Note changes or attach parent)

New X Question History: Last NRC Exam (Optional: Questions validated at the facifity since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis _X_

10 CFR Part 55 Content: 55.41 55.43 (2)

Comments:

Question #82 A steam leak in the Drywell has resulted in the following:

Drywell average temperature is 310'F and rising 1 'F/min Drywell pressure is 15 psig and rising 'I psig/min No Operator actions have been taken.

Which one of the following describes the next action REQUIRED to be taken?

_ _ _ _ _H _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ~ , ____________________ ~

If THEN Lk~"Wdl tOllI'K'r:t1Uf\.' ctnU(jt h*t: Hpev;nc:J'U "1.V~Ltl ..tik d-f)- Wt*U t'ooling n::'>i~ort.,:"t ;~Ili' u'1J.tnt.lil)"il Ix:-!.. },-"," !,;'f BEFORE

~

!>fY'",l1l<'IIII'CTJllln, n:;lI:hb

I()'}'! IIH~wdI5I'r~))

< DRYWEU.SI>RAY IS REQt,ltRllD.

If THEN n'r!o \\'dl n mp('r;HHn:~ (';,l:U].;J! h(: EMERGENCY SlPV IT~ht~'d and m.utlt;l.tlli:d hdil\\' _'i~ff DEPRESSURIZATION IS REQUlllED, ",lin fOP,'!, "Rl'V Con*n)'."" .md e:\'L\.Uh",* it <:..-}n\.ufrt.*ntl~

wifh th~~ pn"'lI.-\:'dufi:.

A. Spray the Drywell.

8. Start all eight Drywell cooling fans.

C. Emergency Depressurize the Rx.

D. Anticipate Emergency Depressurization.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level SRO Tier# _1_

Group # _1_

KIA # 295024 EA2.02 Importance Rating 4.0 to determine and/or rhe DRYlIVELL PRESSUf<:E:

Proposed Question: 82 A steam leak in the Drywell has resulted in the following:

Drywell average temperature is 310'F and rising 1 'F/min Drywell pressure is 15 psig and rising 1 psig/min No Operator actions have been taken.

Which one of the following describes the next action REQUIRED to be taken?

r-"----.--------------.----~'--------------------~

If THEN t )r\"\vdl tnUp.*rAtHrt I,.. ;HHh~! t~,

n ...tnrvi ;ind Hl~~i:n.L1UU.'.J hdo>\-' \J~~ I BEFORE

~

I)fy"dl h:/lll'"TJIl.n" reJt:h~' J(I')'1'

- (ll') wdl 'I'r;.~')

< DRYWEI..I.. SPRAY IS REQUIRf.D.

[.--------------,-------------.,.1 IF THEN Drr..",d~ H tn;Wr.HHH' \".HlfH;ll 1)(' EMERG£NO RPV

(,*...t(JrI,d :md nlllmt~t~nn.l tx:lu'l;ts, )o9~r DEPR£SSURIZATION IS REQUI!tED.~tlwr EO!'-..!. "Ill'\'

Cllrttrol,'" :UH~ (.:Xt;'(,,'Ht4" tt "~)It1.;Ufn.*nl1l

~~ __ ~ ______________-L__________________- '\vnh thh prot'If,,'I,Jun:,'. -/

A. Spray the Drywell.

B. Start all eight Drywell cooling fans.

C. Emergency Depressurize the Rx.

D. Anticipate Emergency Depressurization.

Proposed Answer: A, Spray the Drywell

Explanation: EOP-4 requires that Emergency depressurization would be appropriate if temperature could not be restored and maintained less that 309F. With no operator actions taken yet (as given in stem conditions) then OW sprays would have to be attempted before that determination can be made. If an Emergency Depressurization is expected, then it can be anticipated and a rapid depressurization could be performed, however, once again, sprays would have to be attempted before it could be anticipated. Starting a fourth OW fans is only allowed when shifting fans, not for continued use. Operate All available DW cooling means 3 of 4; not 4 of 4 per Cooling Unit.

A. correct-see above B. incorrect-see above C. incorrect-see above O. incorrect-see above Technical Reference(s): EOP-4 EOP-2 Proposed references to be provided to applicants during examination: EOP-11 Learning Objective: MIT-301.11E 4.05_ _ _ (As available)

Question Source: Bank#

Modified Bank # _ _ _ (Note changes or attach parent)

New -X- -

Question History:

(Optional: Questions validated at the facility since 10195 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of evety question.)

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 (5)

Comments

Question #83 The Plant has experienced a Reactor scram from full power.

The following report is received from the At the Controls (ATC) Operator:

The Mode switch is in Shutdown

All APRMs are downscale

IRMs are mid-scale on Range 5

Rod position for two Control Rod cannot be determined by any indication

EPIC "PLANT" screen is displaying: RODS OUT

An un-isolable steam leak has occurred on RCIC

One Reactor Building area temperature is above Max Safe and rising

Another Reactor Building area temperature is above Max Normal and rising Which one of the following correctly lists the procedure entered and strategy for the above conditions?

A. AOP-1 Insert Control Rods per AOP-1 B. AOP-1 Inject Standby Liquid Control C. EOP-3a Emergency Depressurization D. EOP-3 Commence a cooldown < 100'F/hr

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level SRO Tier #

Group # L KIA # 295015 AA2.02 Importance Rating Proposed Question: 83 The Plant has experienced a Reactor scram from full power.

The following report is received from the At the Controls (ATC) Operator:

The Mode switch is in Shutdown

All APRMs are downscale

IRMs are mid-scale on Range 5

Rod position for two Control Rod cannot be determined by any indication

EPIC "PLANT" screen is displaying: RODS OUT

An un-isolable steam leak has occurred on RCIC

One Reactor Building area temperature is above Max Safe and rising

Another Reactor Building area temperature is above Max Normal and rising Which one of the following correctly lists the Procedure to enter and strategy for the above conditions?

A. AOP-1 Insert Control Rods per AOP-1 B. AOP-1 Inject Standby Liquid Control C. EOP-3a Emergency Depressurization D. EOP-3 Commence a cooldown < 100'F Proposed Answer: D. EOP-3. Commence a cooldown < 100'F

Explanation: This question evaluates Candidates knowledge of "will the reactor remain shutdown under all conditions without boron" with more than one Control Rod at or beyond position 02 and Emergency Depressurization requirements.

EOP-5 is entered based on an Area Temp greater than Max Normal. EOP-5 directs entry into EOP-2 with a primary system (RCIC) discharging and one temp above Max Normal. EOP-2 directs entry into EOP-3 because with more than one Rod position unknown, the answer to "will the reactor remain shutdown under all conditions without boron" is NO. EOP-3 directs, in the pressure leg, to cooldown < 100 deg F/hr if the Rx is shutdown without boron (shutdown in this box is defined IRMs <R6). For the distractors, AOP-1 will also be entered on any scram however the second part of the question makes the AOP-1 distractors plausible but wrong for the conditions given (SLC not directed in the AOP and rods are directed if not in the EOPs). For distractor EOP-3a "Emergency Depressurization ", would be warranted if there were 2 temperatures above Max safe.

A. incorrect- see above B. incorrect- see above C. incorrect- see above D. correct- see above Technical Reference(s): _AOP-"l, EOP-3a. EOP-3, EP-1 _ _

Proposed references to be provided to applicants during examination:

Learning Objective: MIT-301 D. _ _ _ (As available)

Question Source: Bank#

Modified Bank # _--,--,-_ (Note changes or attach parent)

New Question History: Last NRC Exam (Optional: Questions validated at the facility since 10195 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of evety question.)

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 (5)

Comments:

Question #84 Three hours ago the Plant had experienced a loss of coolant accident in the Drywell.

Current conditions are as follows:

Chemistry reports the following:

o Drywell Hydrogen concentration is 12%

o Drywell Oxygen concentration liS 8%

o Torus Hydrogen and Oxygen concentrations are not completed yet

Offsite dose rates are projected to exceed release limits if the drywell is vented.

Drywell Gas Control Drywell Oxygen Concentration I .. 5% or cannot be determined to be below 5%

<5% Torus Hydro~enConcentraticm c ;( 0.6%

.g f&Qu!,od

~

0

'" I0

<.)

?O.5't, AND'" 6%

!:~~~;~

DW/Q*1 DWJG-3 Torus Gas Control Torus Oxygen Con:::entratlon

! ;:: 5% or cannol be dewrmined 10 be below 5%

<5% Drywall Hydrogen Concentration

<.: 0,6%

No 1)clioo Nll attiel'!

,g requiMd required

~ ~ 1---1--'="--1--'="--'

TfG-3

... : TIG*!

o u

Based on the above conditions, which one of the following correctly lists the status of (1) Primary Containment (PC) combustion limits and (2) the correct EOP-4 and EOP-4a action(s) for the drywell?

(1) PC combustion limits (2) action A. Below Vent and purge regardless of offsite dose rates.

B. Below Do NOT vent and purge.

c. Above Vent and purge regardless of offsite dose rates.

D. Above Do NOT vent and purge.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level SRO Tier# 1 Group # L KIA # 500000 EA2.03 Importance Rating 3.8

>'~"r,FHn Cone, to dei£}rmine and/or the to HIGH PRiMAI'XY CONTAiNMENT HYDROGEfI,I CONCENTRA nONS:

for Three hours ago the Plant had experienced a loss of coolant accident in the Drywell.

Current conditions are as follows:

Chemistry reports the following:

o Drywell Hydrogen concentration is 12%

o Drywell Oxygen concentration is 8%

o Torus Hydrogen and Oxygen concentrations are not completed yet

Offsite dose rates are projected to exceed release limits if the drywell is vented.

Drywall Gas Control Drywall Oxygnn Concentration

?: 5% or cannot be determined to be below 5%

<5% TorusHydrogenConcentration

< 0.6% I ~o 6% AND < 6'10 be determined o < 0.6%

No action required No aC,tlon reqUired I to be below ~%

-o~ 10 1-------1I---"='----t---"'=--' OW/G-2 DW/G-3

DW/G-l bedelermlned C,.) 10 I.e t>elow 5%

Torus Gas Control Torus Oxygen Concentration

~ 5% or cannol be determined 10 be below 5%

<5% Drywell Hydrogen Concentration

<: 0.6% I :.06% AND < 6% bedelermlned 10 be below 6%

~

.~

< 0.6% No a~tliln I 1---_---1---";:;;;;;:;:,.,-+---=,.'::::":::,.,_-' T/G-2

~ ~ 1-------1 T/G-3 T/G*1

(.) t:et,:e:;~n:

Based on the above conditions, which one of the following correctly lists the status of (1) Primary Containment (PC) combustion limits and (2) the correct EOP-4 and EOP-4a action(s) for the drywell?

(1) PC combustion limits (2) action A. Below Vent and purge regardless of offsite dose rates.

B. Below 00 NOT vent and purge.

C. Above Vent and purge regardless of offsite dose rates.

O. Above 00 NOT vent and purge.

Proposed Answer: C, Above, Vent and purge regardless of offsite dose rates Explanation: This question evaluates Candidates knowledge of Containment combustion limits and what to do if limits are exceeded with offsite dose rates exceeding limits. EOP-4 combustion limits for OW are 02 ,:::5% and H2 ,:::6%. EOP-4a requires that if OW 02 .::5% and H2 .::6% are present then Vent and Purge the OW, defeat isolation interlocks and exceed offsite rad release rate limit if necessary. Torus value do not have to be known for this to occur, Torus values are an "or" condition in EOP-4a. Protecting PC takes precedence. Additionally, parameters given do not exceed PCPL therefore venting is allowable.

A. incorrect- see above B. incorrect- see above C. correct- see above O. incorrect- see above Technical Reference(s): - EOP-4' 4a Proposed references to be provided to applicants during examination: none Learning Objective: MIT-301 E 4.05 _ _ _ (As available)

Question Source: Bank #

Modified Bank # _ _ _ (Note changes or attach parent)

New X Question History: Last NRC Exam (Optional: Questions validated at the facility since 10195 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 (5)

Comments:

Question #85 The Plant is operating at full power with two evolutions in progress:

1. ISFSI fuel cask is being loaded

2. HPCI surveillance test in progress The following events occur:

A spent fuel bundle is dropped resulting in the Spent Fuel Pool Area Radiation Monitor reading above Max Safe.

A HPCI steam leak has resulted in the East Crescent Area Radiation Monitor reading above Max Safe.

HPCI Steam Supply valves 23MOV-15 and 23MOV-16 failed to isolate.

Both values are confirmed to be accurate and valid by RP.

Which one of the following correctly lists the appropriate order to be directed by the CRS and the reason why?

CRS order A Maintain Rx power steady Minimize crew actions and distractions until further reports are received B. Override Rx Bldg Vent radiation To allow ventilation to restore building habitability isolation and personnel access.

C. Commence a normal Rx shutdown In an attempt to lower radiation levels in the East HPCI area D. Scram and Emergency To place the Rx in its lowest energy state to depressurize the Rx minimize the driving head of the leak

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level SRO Tier # 1 Group # .L KJA# 295033 G2.1.32 Importance Rating 4.0 Proposed Question: 85 The Plant is operating at full power with two evolutions in progress:

ISFSI fuel cask is being loaded

HPCI surveillance test in progress The following events occur:

A spent fuel bundle is dropped resulting in the Spent Fuel Pool Area Radiation Monitor 4

reading 1xi 0 mr\hr.

A HPCI steam leak has resulted in the, East Crescent Area Radiation Monitor reading 1xi 04 mr\hr.

HPCI Steam Supply valves 23MOV-1!5 and 23MOV-i6 failed to isolate.

Both values are confirmed to be accurate and valid by RP.

Which one of the following correctly lists the appropriate order to be directed by the CRS and the reason why?

CRS order A. Maintain Rx power steady Minimize crew actions and distractions until further reports are received B. Override Rx Bldg Vent radiation To allow ventilation to restore building habitability isolation and personnel access.

C. Commence a normal Rx shutdown In an attempt to lower radiation levels in the East HPCI area D. Scram and Emergency To place the Rx in its lowest energy state to depressurize the Rx minimize the driving head of the leak

Proposed Answer: 0, Scram and Emergency depressurize the Rx, To place the Rx in its lowest energy state to minimize the driving head of the leak Explanation: This question evaluates a Candidates knowledge of appropriate actions should 2 or more parameters in the RB exceed their maximum safe value and the reason for the action.

If two values of the SAME parameter (even if unrelated) exist above max safe, it requires at a minimum a Rx shutdown. However one of the area max safe values is from a Primary system discharging. EOP-5 requires an Emergency Depressurization under this condition in order to place the primary system in its lowest possiblE~ energy state, reject heat to the torus in preference to outside the containment, and reduce the driving head and flow of primary systems that are un isolated and discharging into the secondary containment. The criteria of "two or more" specified in EOP-5 identifies the rise in reactor building radiation level as a wide-spread problem which may pose a direct and immediate threat to plant equipment and to personnel both on and off the site. One parameter (e.g., radiation) above its maximum safe operating value in one area and a different parameter (e.g., temperature or water level) above its maximum safe operating value in the same or another area is not a condition which requires emergency RPV depressurization. A combination of parameters exceeding maximum safe operating values in one area does not necessarily indicate that control of a given parameter cannot be maintained or that previous actions have not been effective in confining the trouble to one area. Expanding the application to encompass multiple parameters might lead to depressurization of the RPV when such action is not needed.

A. incorrect- see above B. incorrect- the LOCA overrides can be bypassed but not the radiation one.

C. incorrect- see above D. correct- see above Technical Reference(s): _EOP-S_

Proposed references to be provided to applicants during examination: EOP-11 graphs Learning Objective: MIT -301.11 F 1.07 _ _ _ (As available)

Question Source: Bank#

Modified Bank # _ _ _ (Note changes or attach parent)

New Question History: Last NRC Exam (Optional: Questions validated at the facility since 10195 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis _X_

10 CFR Part 55 Content: 55.41 55.43 (2)

Comments:

Question #86 The Plant just entered a Forced Outage due to a fault on the 10600 Bus and an unexplained rise in Oryweilleakage. The following conditions exist:

Time: 0800

Both Orywell Personnel Airlock doors have just been opened.

RHR Loop 'A' is in a Shutdown Cooling lineup.

Reactor Coolant temp currently 200F.

RHRSW Pump 'C' is danger tagged out of service.

Time: 0805 The following alarms are received:

09-3-1-25 RHRSW PMP 10P-1A TRIP

09-3-1-26 RHRSW PMP 10P-1A OVERCURRENT Time: 0810

Only one Orywell Personnel Airlock door can be closed I!!:n£.:. 0815

RPV temp is 213'F.

Based on the above information, which one o'f the following is a correct statement?

A. At Time 0800 Primary Containment was inoperable ONLY B. At Time 0805 RHR Loop 'A' was inoperable ONLY C. At Time 0805 Primary Containment AND RHR Loop 'A' was inoperable O. At Time 0815 Primary Containment AND RHR Loop 'A' was inoperable

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level SRO Tier # -.L Group # _1_

KIA # 205000 2.4.31 4.1 or Proposed Question: 86 The Plant just entered a Forced Outage due to a fault on the 10600 Bus and an unexplained rise in Dryweilleakage. The following conditions exist:

Time: 0800

Both Drywell Personnel Airlock doors have just been opened.

RHR Loop 'A' is in a Shutdown Cooling lineup.

Reactor Coolant temp currently 200'F.

RHR Pump 'B' is danger tagged out of service.

Time: 0805 The following alarms are received:

09-3-1-25 RHRSW PMP 10P-1A TRIP

09-3-1-26 RHRSW PMP 10P-1A OVERCURRENT Time: 0810

Only one Drywell Personnel Airlock door can be closed Time: 0815

RPV temp is 213'F.

Based on the above information, which one of the following is a correct statement?

A. At Time 0800 Primary Containment was inoperable ONLY B. At Time 0805 RHR Loop 'N was inoperable ONLY C. At Time 0805 Primary Containment AND RHR Loop 'A' was inoperable D. At Time 0815 Primary Containment AND RHR Loop 'A' was inoperable Proposed Answer: B, At Time 0805 RHR Loop 'A' was inoperable ONLY

Explanation: This question evaluates a Candidates knowledge of Annunciators and their impact on Plant conditions and based on that, evaluate the operability of Primary Containment and RHR SOC. Conditions given are indicative of a loss of shutdown cooling. (10600 bus de energized. RHR pump 'B' out of service, and RHRSW pump 'A' trip leads to no cooling to Loop A SOC. Per TS 3.4.8 Bases. RHRSW is a component necessary for operability of SOC.

Primary Containment is not required to be operable when ,:::212F. Therefore both OW Equipment hatches can be open at the same time. However, the hatches are included in the TS Bases for operability determination. One hatch must be shut to declare the Pri Cont operable when >212F.

A. incorrect- see above B. correct- see above C. incorrect- see above O. incorrect- see above Technical Reference(s): _TS 3.4.8 RHR SOC, TS 3.6.1.1 Pri Containment_

Proposed references to be provided to applicants during examination:

Learning Objective: _SOLP-16A 1.18_S0LP-10 1.0ge 1.18.d_ (As available)

Question Source: Bank #

Modified Bank # _-:-:-_ (Note changes or attach parent)

New Question History: Last NRC Exam (Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge -

Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 Comments:

Question #87 The plant was at 100% power when the following happens:

A LOCA occurs

The LOCA is eventually terminated by the closure of the recirc suction & discharge valves (02MOV-43A1B & 02MOV-53A1B)

Torus Level - 18 ft and rising

HPCI - in auto and injecting into the vessel

RCIC - in auto and injecting into the vessel

Condensate and Feedwater are unavailable

Rx Level - 140" and rising 20"/min

Reactor Pressure - 1000 psig and steady

OW Pressure - 4 pSig and steady

MSIV's are closed In regards to HPCI and RCIC, what order should the CRS give?

A. Control level with HPCI and place RCIC in standby B. Control level with HPCI and place RCIC in pressure control mode C. Ensure HPCI is aligned to the CST, control level with both HPCI and RCIC O. Secure HPCI, align RCIC for both injection and pressure control mode

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level SRO Tier #

Group # _1_

KJA# 217000 A2.17 Importance Rating 3.4 Proposed Question: 87 The plant was at 100% power when the following happens:

A LOCA occurs

The LOCA is eventually terminated by the closure of the recirc suction & discharge valves (02MOV-43A1B & 02MOV-53A1B)

Torus Level - 18 ft and rising

HPCI - in auto and injecting into the vj~ssel

RCIC - in auto and injecting into the vessel

Condensate and Feedwater are unavailable

Rx Level 140" and rising 20"/min

Reactor Pressure - 1000 psig and steady

OW Pressure - 4 psig and steady

MSIV's are closed In regards to HPCI and RCIC, what order should the CRS give?

A. Control level with HPCI and place RCIC in standby B. Control level with HPCI and place RCIC in pressure control mode C. Ensure HPCI is aligned to the CST, control level with both HPCI and RCIC O. Secure HPCI, align RCIC for both injection and pressure control mode

Proposed Answer: A, Control level with HPCI and place RCIC in standby Explanation: The crew is in EOP-2 and EOP-4. EOP-2 gives direction to restore level to 177 222.5" with Group 1 systems. The Group 1 section includes HPCI and RCIC with directions to align to CST if available. However, EOP-4, for high torus water level, directs that if torus water level and RPV pressure can not be maintained below the SRV Tail Pipe Level limit (approximately 18 ft) and adequate core cooling can be assured, then terminate injection from sources external to primary containment. Normally, HPCI and RCIC are aligned to the CST.

When torus water level reaches 14.5 ft HPCI automatically aligns suction to the torus.

Adequate core cooling is assured because the stem states the LOCA has been secured and the fill rate for the core indicates virtually all HPCI flow (4,300 gpm) is going into the vessel.

Therefore, RCIC should be placed in standby (to secure external injection) and HPCI allowed to bring level into the EOP-2 level band A. correct- see above.

B. incorrect- Plausible because EOP-4 directs RPV pressure be maintained below SRV Tail Pipe level limit. However, although RCIC itself can be aligned for pressure control, it relies on a HPCI valve to accomplish this. Because HPCI has an auto initiation signal, that valve can not be opened.

C. incorrect- Plausible because EOP-2 directs injection with the CST (if possible) and EP-2 has an override that will allow HPCI suction to be transferred from Torus to CST. But the candidate would have failed to recognize the actions required for SRV Tail Pipe level Limit.

D. incorrect- Plausible because the fill rate is rapid and HPCI will trip on high level. If RCIC is placed in both level and pressure control then RCIC can run at a constant GPM and be adjusted periodically so it maintains level in band without tripping. However, the candidate would not recognize that the needed HPCI valve could not be opened and that there should not be an external injection source going into the vessel.

Technical Reference(s): _EOP-2, EOP-4_

Proposed references to be provided to applicants during examination: !\lone Learning Objective: __MIT-301.11E 4.05__ (As available)

Question Source: Bank #

Modified Bank # _ _ _ (Note changes or attach parent)

New x Question History: Last NRC Exam (Optional: Questions validated at the facility since 10195 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis _X_

10 CFR Part 55 Content: 55.41 55.43 Comments:

Question #88 The Plant is in a Refuel Outage with the following activities in progress:

Control Rod Drive Mechanism (CRDM) replacements

'B' Battery system is tagged out of service for battery cell replacements The following occurs:

Annunciator 09-8-1-21 (125VDC BAn" CHGR A DC GROUND) alarms

Ground Detector 'A' meter indicates (+)150 VDC (pegged high)

Battery Charger "A' trips

Battery 'A' Bus voltage reads 127 VDC and lowering slowly Concerning Technical Specification 3.B.5 (DC Sources-Shutdown), which one of the following correctly states whether or not LCO 3.8.5 is currently satisfied and the required action to perform?

LCO 3.8.5 is... required action to perform ...

A currently satisfied Start 71 BC-9 (temporary battery charger) per OP-43A "125VDC Power System""

B. is NOT currently satisfied Immediately suspend CRDM replacements C. currently satisfied If ground is not isolated, a 10CFR50.59 is required for long term plant operation.

D. is NOT currently satisfied Enter AOP-45, Loss of DC Power System 'A'.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level SRO Tier#

Group # _1_

KIA # 263000 A2.01 Importance Rating 3.2 Proposed Question: 88 The Plant is in a Refuel Outage with the following activities in progress:

Control Rod Drive Mechanism (CRDM) replacements

'B' Battery system is tagged out of service for battery cell replacements The following occurs:

Annunciator 09-8-1-21 (125VDC BATT CHGR A DC GROUND) alarms

Ground Detector 'A' meter indicates (+)150 VDC (pegged high)

Battery Charger 'A' trips

Battery 'A' Bus voltage reads 127 VDC and lowering slowly Concerning Technical Specification 3.8.5 (DC Sources-Shutdown), which one of the following correctly states whether or not LCO 3.S.5 is currently satisfied and the required action to perform?

LCO 3.S.5 is... required action to perform ...

A. currently satisfied Start 71 BC-9 (temporary battery charger) per OP-43A "125VDC Power System".

B. is NOT currently satisfied Immediately suspend CRDM replacements C. currently satisfied If ground is not isolated, a 10CFR50.59 is required for long term plant operation.

D. is NOT currently satisfied Enter AOP-45, Loss of DC Power System 'A'.

Proposed Answer: B, is NOT currently satisfied, immediately suspend CRDM replacements Explanation: Question evaluates a Candidates knowledge of Operability determination of the 125VDC Battery. Tech Spec 3.8.5, Surveillance Requirement and Bases requires Battery voltage ~ 128VDC. Therefore LCO is NOT met. CRDM replacements are considered an OPDRV which are required to be Immediately suspended. Also, the 125V Battery is a 'support' system to RCIC which would have to be declared Inop however, RCIC LCO is not applicable in Mode 5. The actions of distractors A & C come from AOP-45, Loss of DC Power System A and are correct by themselves. However the first part of the distractors is wrong (the LCO is NOT satisfied).

A. incorrect- see above B. correct- see above C. incorrect- see above D. incorrect- see above Technical Reference(s): TS3.8.5_

Proposed references to be provided to applicants during examination:

Learning Objective: __ ~ILP-OPS-ITS02 1.02 __ (As available)

Question Source: Bank #

Modified Bank # _ _ _ (Note changes or attach parent)

New Question History: Last NRC Exam (Optional: Questions validated at the facility since 10195 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of eve/)' question)

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 55.43 (5)

Comments:

Question #89 The Plant is operating at power when Annunciator 09-6-2-4 (ESW SYS A OVERLOAD OR CNTRL PWR LOSS) alarms.

The Control Room Operator reports:

o Green light for 46MOV-101A (ESW SYS A INJ VLV) is extinguished.

With regards to the ESW 'A' system and Emergency Diesel Generators (EDGs) 'A' and 'C',

which one of the following correctly lists the Tech Spec operability status of the two systems?

ESW'A' EDG 'A' and 'c' A. Operable Operable B. Inoperable Operable C. Operable Inoperable D. Inoperable Inoperable

ES-401 Sample Written Examination Form ES*401*5 Question Worksheet Examination Outline Cross-

Reference:

Level SRO Tier# --.L Group # _1_

KIA # 400000 G2.2.44 Importance Rating 4.4 Proposed Question: 89 The Plant is operating at power when Annunciator 09-6-2-4 (ESW SYS A OVERLOAD OR CNTRL PWR LOSS) alarms.

The Control Room Operator reports:

o Green light for 46MOV-101A (ESW SYS A INJ VLV) is extinguished.

With regards to the ESW 'A' system and Emergency Diesel Generators (EDGs) 'A' and 'C',

which one of the following correctly lists the Tech Spec operability status of the two systems?

ESW'A' EDG 'A' and 'C' A. Operable Operable B. Inoperable Operable C. Operable Inoperable D. Inoperable Inoperable Proposed Answer: B, Inoperable. Operable

Explanation: This question evaluates the Candidates knowledge of understanding the impact that a system component (valve) failure has on Tech Spec operability status of that system and a system supported by that degraded system. Annunciator 09-6-2-4 (ESW SYS A OVERLOAD OR CNTRL PWR LOSS) means that 46MOV-101A (ESW SYS A INJ VLV) will not open if an ESW Lockout Matrix signal is generated. 46MOV-101A supplies ESW water to various components but does not affect cooling to the EDGs. TS 3.7.2 (ESW system and UHS) requires two ESW subsystems to be operable. TS 3.7.2 Bases defines operable as: the subsystem has an operable UHS, one operable pump and an operable flow path capable of taking suction from the intake structure and transferring the water to the appropriate equipment.

Therefore since the transfer of water is impacted, the ESW 'A' subsystem is inoperable. Now, entering Condition A of TS 3.7.2 the Required Action statement is: NOTE: enter applicable Conditions and Required Actions of LCO 3.8.1 AC Sources-Operating, for emergency diesel generator subsystem made inoperable by ESW. Candidate may believe this direction is applicable due to 46MOV-101A not being able to be opened, however, ESW flow to EDG 'A' and 'C' is not impacted therefore EDG 'A' and 'C' are still operable.

A. incorrect- see above B. correct- see above C. incorrect- see above D. incorrect- see above Technical Reference(s): _OP-21 and TS 3.7.2 and Bases_

Proposed references to be provided to applicants during examination: TS 3.7.2 Learning Objective: __ SDLP-468 1.05, 1.16,17,18_ _ (As available)

Question Source: Bank#

Modified Bank # _ _ _ (Note changes or attach parent)

New X Question History: Last NRC Exam (Optional: Questions validated at the facility since 10195 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 (5)

Comments:

Question #90 The Plant was operating at power when a Seismic event occurred.

The following conditions exist:

Unisolable Reactor Water Cleanup leak in the Reactor Building

Drywell pressure is 1.9 psig and rising slowly

RPV water level is 65 inches and lowering slowly

RPV pressure is 250 psig and lowering slowly

LPCI Pump 'A' is the only injection source with power to it

LPCI Pump 'A' is not running

Annunciator 09-3-1-18 (RHR A or B DISCH NOT FULL) is in alarm Concerning LPCI Loop 'A'. which one of the following correctly states the (1) Operability, (2) ) when the pump can be started and used for injection?

(1) Operable (2) when the pump can be started and used for injection A. Yes Can be started with precautions taken to protect against water hammer.

B. Yes Can NOT be started until "RHR A or B DISCH NOT FULL" alarm is cleared.

C. No Can be started with precautions taken to protect against water hammer.

D. No Can NOT be started until "RHR A or B DISCH NOT FULL" alarm is cleared.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level SRO Tier # L Group # _1_

KJA# 203000 A2.17 Importance Rating 3.5 Proposed Question: 90 The Plant was operating at power when a Seismic event occurred.

The following conditions exist:

Unisolable Reactor Water Cleanup leak in the Reactor Building

Drywell pressure is 1.9 psig and rising slowly

RPV water level is 65 inches and lowering slowly

RPV pressure is 250 psig and lowering slowly

LPCI Pump 'A' is the only injection source with power to it

LPCI Pump 'A' is not running

Annunciator 09-3-1-18 (RHR A or B DISCH NOT FULL) is in alarm Concerning LPCI Loop 'A', which one of the following correctly states the (1) Operability and (2) when the pump can be started and used for injection?

(1) Operable (2) when the pump can be started and used for injection A. Yes Can be started with precautions taken to protect against water hammer.

B. Yes Can NOT be started until "RHR A or B DISCH NOT FULL" alarm is cleared.

C. No Can be started with precautions taken to protect against water hammer.

D. No Can NOT be started until "RHR A or B DISCH NOT FULL" alarm is cleared.

Proposed Answer: C: No, Can be started with precautions taken to protect against water hammer.

Explanation: Question evaluates a Candidates knowledge of Operability and the usage of that component if needed for injection. Tech Spec 3.5.1 requires (per SR 3.5.1.1) that ECCS injection piping be filled with water from pump discharge valve to the injection valve to prevent water hammer damage when the pump is started. Annunciator 09-3-1-18 (RHR A or B DISCH NOT FULL) means the piping is not full of water therefore the ECCS Loop is Inoperable. The Caution in the ARP states: Starting an RHR pump in an RHR loop that is not full could result in severe water hammer and equipment damage. RHR loop piping shall be full prior to manually starting an RHR pump in that loop. There are two ways to fill the system: 1 - with the Keepfili/Condensate transfer system 2- in an emergency. by starting the pump with the discharge valve closed and slowly opening the valve to fill the system. In either case. flow can not be maximized until the system is filled.

A. incorrect- see above B. incorrect- see above C. correct- see above D. incorrect- see above Technical Reference(s): _TS 3.5.1. ARP 09-3-1-18 Proposed references to be provided to applicants during examination:

Learning Objective: __SDLP-10 1..10.1.14,1.16 (As available)

Question Source: Bank#

Modified Bank # (Note changes or attach parent)

New Question History: Last !\IRC Exam (Optional: Questions validated at the facility since 10195 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge -

Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 Comments:

Question #91 The plant is at 75% power when the following alarms come in:

09-6-3-14 FDWTR HTR 1A DRN LOOP LVL HI 09-6-3-04 FDWTR HTR 1A DRN LOOP LVL HI-HI The STA reports RPV FW Inlet temperatures as follows:

Epic Point A-3411: 385 deg F

Epic Point A-3412: 383 deg F Which one of the below choices completes the following statement?

MCPR limits ____ and/but the ATe operator must be directed to reduce power to a maximum of power.

A. MAY be violated 54%

B. MAY be violated 24%

C. are NOT violated 54%

D. are NOT violated 24%

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

LElvel SRO Tier#

Group # .L KJA# 256000 G2.4.11 Importance Rating 4.2 Proposed Question: 91 The plant is at 75% power when the following alarms come in:

09-6-3-14 FDWTR HTR 1A DRN LOOP LVL HI 09-6-3-04 FDWTR HTR 1A DRN LOOP LVL HI-HI The STA reports RPV FW Inlet temperatures as follows:

Epic Point A-3411: 385 deg F

Epic Point A-3412: 383 deg F Which one of the below choices completes the following statement?

MCPR limits ____ and/but the A TC operator must be directed to reduce power to a maximum power.

A. MAY be violated 54%

B. MAY be violated 24%

C. are NOT violated 54%

D. are NOT violated 24%

Proposed Answer: B, may be violated, 24%.

Explanation: The Feedwater High alarm is an indication of a condensate tube leak and the FW temps indicate a loss of FW heating. AOP-62, "Loss of FW Heating" is entered. To determine MCPR limits, the EPIC temperatures must be compared to AOP-62 graph, Attachment 1. The temperatures are outside the bounds of the graph. The procedures says that temperatures outside the graph means the limits may be violated. The power reduction to 20% lower than pre-transient levels, than 54% would be correct if just the Hi level alarm comes in. However the Hi-Hi level causes an isolation and requires a further power reduction to 24%.

A. incorrect- see above B. correct - see above C. incorrect- see above D. incorrect- see above Technical Reference(s): _AOP-62 Proposed references to be provided to applicants during examination: Attachment 1, AOP-62 Learning Objective: _ _ _ (As available)

Question Source: Bank #

Modified Bank # _ _ _ (Note changes or attach parent)

New Question History: Last NRC Exam (Optional: Questions validated at the facility since 10195 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detai/ed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 (5)

Comments:

Question #92 The Plant is operating at 75% power with Standby Gas Treatment (SBGT) Fan 'A' out of service for maintenance when the following occurs:

Drywell pressure reaches 2.8 psig

SBGT Fan 'B' fails to start

Reactor Building Ventilation Radiation Monitors read 145 cpm

Reactor Building dP is 0.1 inches water and steady Based on the above given information. which one of the following choices correctly states whether or not EOP-5 and \ or EP-2 is required to be entered \ executed and whether or not Secondary Containment is Operable or Inoperable?

EOP-5 entry required? EP-2 executed? Secondary Containment Operable?

A. No Yes No B. Yes Yes No C. Yes No Yes D. No No Yes

ES-401 Sample Written Examination Form E5-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level SRO Tier #

Group # _1_

KJA# 290001 G2.4.1 Importance Rating 4.8 Proposed Question: 92 The Plant is operating at 75% power with Standby Gas Treatment (SBGT) Fan 'A' out of service for maintenance when the following occurs:

Drywell pressure reaches 2.8 psig

SBGT Fan 'B' fails to start

Reactor Building Ventilation Radiation Monitors read 145 cpm

Reactor Building dP is 0.1 inches water and steady Based on the above given information, which one of the following choices correctly states whether or not EOP-5 and \ or EP-2 is requirE!d to be entered \ executed and whether or not Secondary Containment is Operable or Inoperable?

EOP-5 entry required? EP-2 executed? Secondary Containment Operable?

A. No Yes No B. Yes Yes No C. Yes No Yes D. No No Yes Proposed Answer: B, Yes, Yes, No Explanation: Question evaluates Candidates knowledge of EOP entry conditions, procedures used and Operability of Secondary Containment. 2.8 psig in DW results in an auto isolation of RB Vent and auto start of SGBT. However, stem states one train SBGT OOS and the other train failed to start. This results in no vacuum being drawn on RB, hence dP of 0.1 inches water. While not a "leg" in EOP-5, a dP 0.1 inches water is an entry condition to EOP-5. The

override section in EOP-5 states: IF RB ventilation isolates AND RB Vent Exhaust Rad level is below 104 cpm THEN Restart RB Ventilation. Defeat high drywell pressure and low RPV water level isolation if necessary (EP-2). RB Vent Rad levels given are <104 cpm therefore defeating isolation per EP-2 is warranted. TS 3.6.4.1 (Secondary Containment) requires a dP of ~0.25 inches vacuum water gage for the Sec Cont to be considered Operable.

A. incorrect- see above B. correct- see above C. incorrect- see above D. incorrect- see above Technical Reference(s): _EOP-5 TS 3.6.4.1 Proposed references to be provided to applicants during examination:

Learning Objective: _SDLP-66A 1.14, 18_ MIT-301.11F 1.03, 04_ (As available)

Question Source: Bank#

Modified Bank # _-:-:-_ (Note changes or attach parent)

New X Question History: Last NRC Exam (Optional: Questions validated at the facility since 10195 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of evety question.)

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 (5)

Comments:

Question #93 The Plant is operating at 94% power in the month of August.

Record heat wave for central NY.

Outside air temperature is 112F.

Control Room temperature is 91F.

Lake water temperature is 83F.

The following occurs:

Both Refrigeration Water Chiller Units (70RWC-2A & B) trip and cannot be restarted.

Emergency Service Water is supplied to the Air Handler Units.

Which one of the following choices correctly lists the status of Control Room HVAC?

A. Operable until Control Room temperature exceeds 104 F B. Operable until Lake water temperature exceeds 87F C. Inoperable due to Control Room temperature exceeding 75F D. Inoperable due to Lake water temperature exceeding 77F

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level SRO Tier # ....L Group # _1_

KJA# 290003 A2.02 Importance Rating 3.4 Proposed Question: 93 The Plant is operating at 94% power in the month of August.

Record heat wave for central NY.

Outside air temperature is 112'F.

Control Room temperature is 91 'F.

Lake water temperature is 83'F.

The following occurs:

Both Refrigeration Water Chiller Units (70RWC-2A & B) trip and cannot be restarted.

Emergency Service Water is supplied to the Air Handler Units.

Which one of the following choices correctly lists the status of Control Room HVAC?

A. Operable until Control Room temperature exceeds 104'F B. Operable until Lake water temperature exceeds 87'F C. Inoperable due to Control Room temperature exceeding 75'F D. Inoperable due to Lake water temperature exceeding 77'F Proposed Answer: A, Operable until Control Room temperature exceeds 104 'F

Explanation: Tech Spec 3.7.4:

3.7.4 Control Room Air Conditioning (Ae) System LeO 3.7.4 Two oontrol room AC subsystems ahal. be OPERABLE.

APPLlCABIUTY; MODES 1, 2, and 3, During movement of recentty Irradiated fuel assemblies In the secondary containment, During operations with a potential for draining the reactor vessel (OPDRVs).

ACTIONS CONOlTlON REQUIRED ACTION COMPLETION TIME A. One control room AC A.l Restore control room At:. 30 days subsystem Inoperable. subsptem to OPERABLE statUI.

B. Two centrolroom AC B.l Venfy centro:! room area Once per 4 houl'$

subsystems Inoperable. tenlpenrture < 90

F.

6NQ 8.2 Restore one control 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months room AC subsystem to OPERASLE status.

C. Required Action and C.l Be In MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 8S8OCtated Completion nme of Condition Aor B AtQ not met In MODE 1. 2, or

3. C.:2 Be In MODE 4. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months (continued)

SURVEILLANCE REQUfREMENTS SURVEILl.ANCE FREQUENCY SR 3.7.4.1. Verify each control room /1£ subsystem has the 24 months capabUlt,ytoremove the assumed neat toad.

The Control Room AC System Is designed to provide a controlled environment under both normal and accident conditions. A single subsystem provides the requIred temperature control to malnta,in a suitable control room environment for 8 sustaIned occupancy of 20 persons. The design conditions for the control room environment are 75°F and 50% relative humid~. This can be accomplished when a control room chiller Is provlding the cooling medium to the cooling colis of an air handling unit. The control room chilliers are non-safety related; however the Control Room .AC System stili meets safety..

related QA Category I requirements when the Emergency Service Water System is aligned to directly supply the cooling coUs. "rhe resulting maximum control room environmental conditions when the Emergency Service Water System is supplying the air handling unit cooling coils is 104°F assuming a lake temperature of 85°F. This satisfies the OPERABIUTY requIrements of the control room equipment. The Control Room ACSystem operation in maintatning the control room temperature is discussed In the UFSAR, Section 9.9.3.11 (Ref. 1)..

The Control Room AC System is considered OPERABLE when the Individual components necessary to maintain the control room temperature are OPERABLE In both subsystems. These components include the air handUng units, recirculation exhaust fans, air handUng unit fans, ductwork, dampers, and assOCiated Instrumentation and controls. The cooUng colts of the air handling units may be cooled by the control room chillers, but to satisfy this leO the Emergency Service Water System must be capable of alignment to provide coonng water directly to the cooling coils.

A. correct- per TS bases 104'F is the max temperature with ES W in service.

A. incorrect- 87'F is the max FSAR temperature for RHRSW. (FSAR 6.5.1) ESW (and Control Room Vent) would be inop if lake temp reached 85'F.

B. incorrect- 75'F is the design temperature with t he chillers in service but ESW (not chillers) is required for operability C. incorrect- 77'F is the maximum temperature for t he lake epilimnion (upper layer of stratified water) per FASR 2.3.5.1.

Technical Reference(s): _OP-55B, Tech Spec 3.7.4 _

Proposed references to be provided to applicants during examination:

Learning Objective: _SDLP-70 1.16,1.18 (As available)

Question Source: Bank#

Modified Bank # _ _ _ (Note changes or attach parent)

New X Question History: Last NRC Exam (Optional: Questions validated at the facility since 10195 will generalfy undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 Comments:

Question #94 Technical Specification 3.4.6 "RCS Specific Activity" limits the allowable specific activity of 1-131 in the Reactor Coolant to certain values.

Which one of the following selections correctly lists the Bases behind the imposed limits?

limit the exposure ______ during a DBA_ _ _ _ __

A. Personnel at the Site boundary and Main Steam line Break Control Room B. Personnel at the Site boundary only Recirc Loop Break C. Personnel at the Site boundary only Main Steam line Break D. Personnel at the Site boundary and Recirc Loop Break Control Room

ES*401 Sample Written Examination Form ES-401*5 Question Worksheet Examination Outline Cross-

Reference:

Level SRO Tier # 2 Group #

KIA # G 2.1.34 Importance Rating 3.5 Proposed Question: 94 Technical Specification 3.4.6 "RCS Specific Activity" limits the allowable specific activity of 1-131 in the Reactor Coolant to certain values.

Which one of the following selections correctly lists the Bases behind the imposed limits?

Limit the exposure during a DBA'--_ _ _ __

A. Personnel at the Site boundary and Main Steam Line Break Control Room B. Personnel at the Site boundary only Recirc Loop Break C. Personnel at the Site boundary only Main Steam Line Break D. Personnel at the Site boundary and Recirc Loop Break Control Room Proposed Answer: A, Personnel at the Site boundary and Control Room, Main Steam Line Break

Explanation: TS 3.4.6 Bases: This Main Steam Line Break (MSLB) release forms the basis for determining offsite and control room doses. The limits on the specific activity of the primary coolant ensure that the 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months thyroid and whole body doses at the site boundary. Resulting from an MSLB outside containment during steady state operation will not exceed 10% of the dose guidelines of 10 CFR 100. The limits on the specific activity of the primary coolant also ensure the thyroid dose to the control room operators, resulting from an MSLB outside containment during steady state operation will not exceed the limits specified in GDC 19 of 10 CFR 50. Appendix A.

A. correct- see above B. incorrect- see above C. incorrect- see above D. incorrect- see above Technical Reference(s): _TS 3.4.6 Proposed references to be provided to applicants during examination:

Learning Objective: __ JLP-OPS**ITS02 1.05_NET-238.3 1.04_ (As available)

Question Source: Bank#

Modified Bank # _ _ _ (Note changes or attach parent)

New x Question History: Last NRC Exam (Optional: Questions validated at the facility since 10195 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 (5)

Comments:

Question #95 The Plant is in Mode 5 conducting refueling operations. The following conditions are present:

A fuel bundle is in transit from the RPV cavity to the Spent Fuel Pool.

The Reactor Analyst on the Refuel Bridge informs the Refuel SRO that the last bundle placed in the Core was in the wrong orientation.

Based on the above information, which one of the following correctly lists the course of action the Refuel Bridge SRO is required to do?

First ... Then ...

A. Complete the current bundle move Re-orientate the mis-orientated bundle to its Spent Fuel Pool location and continue fuel moves B. Return the current bundle to the Stop further fuel moves until GMPO Core position it was removed from permission is granted to continue C. Complete the current bundle move Stop further fuel moves until GMPO to its Spent Fuel Pool location permission is granted to continue D. Return the current bundle to the Re-orientate the mis-orientated bundle Core position it was removed from and continue fuel moves

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level SRO Tier # -L Group #

KJA# G 2.1.35 Importance Rating 3.9 of Proposed Question: 95 The Plant is in Mode 5 conducting refueling operations. The following conditions are present:

A fuel bundle is in transit from the RPV cavity to the Spent Fuel Pool.

The Reactor Analyst on the Refuel Bridge informs the Refuel SRO that the last bundle placed in the Core was in the wrong orientation.

Based on the above information, which one of the following correctly lists the course of action the Refuel Bridge SRO is required to do?

First. .. Then ...

A. Complete the current bundle move Re-orientate the mis-orientated bundle to its Spent Fuel Pool location and continue fuel moves B. Return the current bundle to the Stop further fuel moves until GMPO Core position it was removed from permission is granted to continue C. Complete the current bundle move Stop further fuel moves until GMPO to its Spent Fuel Pool location permission is granted to continue D. Return the current bundle to the Re-orientate the mis-orientated bundle Core position it was removed from and continue fuel moves Proposed Answer: C, Continue moving the current bundle to its Spent Fuel Pool location. Stop further fuel moves until GMPO permission is granted to continue

Explanation: This question evaluates an SRO knowledge of fuel move requirements should a refuel error occur. A mis-orientated fuel bundle is defined as a refuel error. Per OSP-66.001 (Management of Refueling Activities) should at refuel error occur, refueling activities shall be immediately stopped. Refueling activities shall not resume until the condition is resolved and GMPO has granted permission.

A. incorrect- see above B. incorrect- see above C. correct- see above D. incorrect- see above Technical Reference(s): _OSP-66.001 Proposed references to be provided to applicants during examination:

Learning Objective: ___ (As available)

Question Source: Bank#

Modified Bank # _ _ _ (Note changes or attach parent)

New X Question History: Last NRC Exam (Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 (7)

Comments:

Question #96 A plant startup is in progress with the following:

The Reactor is critical and reached the point of adding heat at time 0715

The following heatup data has been recorded:

Time

Reactor pressure (psig) Recirc loop suction (hhmm) temperature (deg F) 0730 10 88 0800 0 118 0830 0 169 0900 0 187 0930 0 211 1000 36 1030 53 1264 1100 74 292 Which one of the following describes the Technical Specification implications of this heatup with respect to heatup rate and Minimum Reactor Coolant System PressurefTemperature Limits?

Heatup Rate Limit Minimum Reactor Coolant System PressurelTemperature Limits A. Not Exceeded Not Exceeded B. Not Exceeded Exceeded C. Exceeded Not Exceeded D. Exceeded Exceeded

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level SRO Tier# 3 Group #

KIA # G 2.2.40 Importance Rating 4.7 Proposed Question: 96 A plant startup is in progress with the following:

The Reactor is critical and reached the point of adding heat at time 0715

The following heatup data has been recorded:

I Time I Reactor pressure (psig) I Recirc loop suction I (hhmm) i temperature (deg F) 0730 0 88

, 0800 0 118 0830 0 169 0900 0 187 0930 0 211 1000 36 232 1030 53 264 1100 74 ?~2 Which one of the following describes the Technical Specification implications of this heatup with respect to heatup rate and Minimum Reactor Coolant System Pressure/Temperature Limits?

Heatup Rate Limit Minimum Reactor Coolant System Pressure/Temperature Limits A. Not Exceeded Not Exceeded B. Not Exceeded Exceeded C. Exceeded Not Exceeded D. Exceeded Exceeded Proposed Answer: B Not Exceeded, ExceedE~d

Explanation (Optional):

A Incorrect -Minimum reactor vessel coolant temperature curve is violated at time 0730 since the reactor is critical with temperature less than 90F.

B. Correct -Heatup rate limit is not exceeded because no one hour temperature change is above 1OOF. Minimum reactor vessel coolant tempera ture curve is violated at time 0730 since the reactor is critical with temperature less than 90'F.

C. Incorrect -Heatup rate limit is not exceeded because no one hour temperature change is above 1OOF. Minimum reactor vessel coolant temp erature curve is violated at time 0730 since the reactor is critical with temperature less than 90F.

D. Incorrect -Heatup rate limit is not exceeded because no one hour temperature change is above 1OOF.

A. incorrect- see above B. correct- see above C. incorrect- see above D. incorrect- see above Technical Reference(s): TS 3.4.9 and TRM App F PTLR (Attach if not previously provided)

Proposed References to be provided to applicants during examination: TS 3.4.9 and TRM App F PTLR Learning Objective: ITS-LP-007 1.04 ___ (As available)

Question Source: Bank # _X_ _

Modified Bank # _ _ _ (Note changes or attach parent)

New Question History: Last NRC Exam 2010 9-Mile Exam (Optional: Questions validated at the facility since 10195 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 (2)

Question #97 The plant was at rated power with Offgas Radiation Monitor 17RM-150A failed downscale due to an electrical failure.

An event then occurred that resulted in the following Annunciators alarming over the next 3 minutes:

09-3-2-27 OFF GAS RAD MON HI 09-3-2-38 OFF GAS RAD MON HI-HI 09-3-1-40 RX BLDG ARM RAD HI 09-3-2-10 OFF GAS TIMER INITIATED The Operator reports the following observations:

RWCU PUMP AREA ARM 18RIA-05119 indicates 40 mr/hr and lowering All Reactor Water Cleanup System isolation valves indicate green light on Offgas Radiation Monitor 17RM-150B indicates 1100 mr/hr and rising slowly Stack Gas Radiation Monitors 17RM-050A&8 are rising slowly Which of the following states the required direction the SRO is to provide to mitigate the radiation hazard?

Offgas Direction Reactor Power Direction A. Offgas will automatically isolate and is Reduce reactor power lAW AOP-3 High to be verified when isolated. Activity In Reactor Coolant or Off-Gas B. Offgas will automatically isolate and is Shutdown the reactor lAW the to be verified when isolated. Secondary Containment Control EOP C. Offgas will NOT automatically isolate Scram the reactor lAW the Secondary and is to be manually isolated. Containment Control EOP D. Offgas will NOT automatically isolate Scram the reactor lAW the and is to be manually isolated. Radioactivity Release Control EOP

ES-401 Sample Written Examination Form ES*401*5 Question Worksheet Examination Outline Cross-

Reference:

L~~vel SRO Tier # ~

Group #

KJA# G 2.3.14 Importance Rating 3.9 Proposed Question: 97 The plant was at rated power with Offgas Radiation Monitor 17RM-150A failed downscale due to an electrical failure.

An event then occurred that resulted in the following annunciators alarming over the next 3 minutes:

09-3-2-27 OFF GAS RAD MON HI 09-3-2-38 OFF GAS RAD MON HI-HI 09-3-1-40 RX BLDG ARM RAD HI 09-3-2-10 OFF GAS TIMER INITIATED The Operator reports the following observations:

RWCU PUMP AREA ARM 18RIA-05119 indicates 40 mr/hr and lowering All Reactor Water Cleanup System isolation valves indicate green light on Offgas Radiation Monitor 17RM-150B indicates 1100 mrlhrand rising slowly

Stack Gas Radiation Monitors 17RM-050A&B are rising slowly Which of the following states the required direction the SRO is to provide to mitigate the radiation hazard?

Offgas Direction Reactor Power Direction A. Offgas will automatically isolate and is Reduce reactor power lAW AOP-3 High to be verified when isolated. Activity In Reactor Coolant or Off-Gas B. Offgas will automatically isolate and is Shutdown the reactor lAW the to be verified when isolated. Secondary Containment Control EOP C. Offgas will NOT automatically isolate Scram the reactor lAW the Secondary and is to be manually isolated. Containment Control EOP D. Offgas will NOT automatically isolate Scram the reactor lAW the and is to be manually isolated. Radioactivity Release Control EOP Proposed Answer: A, Offgas will automatically isolate and is to be verified when isolated.

Reduce reactor power lAW AOP-3 High Activity In Reactor Coolant or Off-Gas

Explanation:

The plant was at rated power with an offgas radiation monitor failed downscale due to a power failure. An event then occurs which shows a leak in the RWCU System, high radiation in the offgas system, an area radiation monitor in alarm, and increased radiation in the stack. Several procedures could be entered: the ARP for the alarms, AOP*3, Secondary Containment Control EOP, and possibly the Radioactive release Control EOP. The Operator observations show that the RWCU System isolated, as designed. The offgas radiation monitor is above the high setpoint and the RWCU ARM radiation indication is lowering (confirming that the RWCU leak has been stopped). The logic for offgas radiation monitors for isolating the offgas system is both hi*hi, or one downscale and 1 hi*hi. Thus, with the information provided, 1 radiation monitor is hi*hi and 1 is downscale. This started the isolation timer and in 15 minutes off-gas will be isolated. For the conditions provided, AOP-3 directs lowering reactor power in an attempt to clear the alarms which, in turn, prevents the o'ffgas system isolation. The Secondary Containment Control EOP directs a scram prior to exceeding a max normal value, provided that a primary system is discharging into the Secondary Containment. From the question, it shows that a primary system was discharging into the Secondary Containment but is no longer and radiation levels are lowering. Thus a scram is not the appropriate action. Indications show that an offsite radiological release is in progress, but it does not rise to the point of the entry condition for the Radioactive Release Control EOP. If it had, then a scram would be appropriate. A reactor shutdown is appropriate in the Secondary Containment Control EOP if a non-primary system were discharging into the Secondary Containment and two radiation areas were above max safe levels.

A. correct* see above B. incorrect- see above C. incorrect- see above D. incorrect- see above Technical Reference(s):

Proposed references to be provided to applicants during examination:

Learning Objective: LP-AOP E.O. 1.02, 1.03 ___ (As available)

Question Source: Bank #

Modified Bank # _ _ _ (Note changes or attach parent)

New Question History: Last NRC Exam 2010 Oyster Creek_ _ _ __

(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X_

10 CFR Part 55 Content: 55.41 55.43 (5)

Comments:

Question #98 The plant has declared a General Emergency with a large release of radioactive material.

All Emergency Facilities are fully manned.

Two people have been selected who might be able to terminate the release and prevent general population overexposure to the release.

It is expected that the person who tries to stop the release will be exposed to 65 REM TEDE.

The two peoples pertinent information is as follow:

Volunteer Status Person 1 Volunteered 32 Person 2 Will go if directed 48 Which one of the below choices indicates the person that should be selected to stop the release and the person who can authorize the exposure?

A. Person 1 Emergency Director OR Shift Manager B. Person 1 Emergency Director ONLY C. Person 2 Emergency Director OR Shift Manager D. Person 2 Emergency Director ONLY

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level SRO Tier # ~

Group #

KIA # G 2.3.4 Importance Rating 3.7 Proposed Question: 98 The plant has declared a General Emergency with a large release of radioactive material.

All Emergency Facilities are fully manned.

Two people have been selected who might be able to terminate the release and prevent general population overexposure to the release.

It is expected that the person who tries to stop the release will be exposed to 65 REM TEDE.

The two peoples pertinent information is as follow:

Volunteer Status Age Person 1 Volunteered 32 Person 2 Will go if directed 48 Which one of the below choices indicates the person that should be selected to stop the release and the person who can authorize the exposure?

A. Person 1 Emergency Director OR Shift Manager B. Person 1 Emergency Director ONLY C. Person 2 Emergency Director OR Shift Manager D. Person 2 Emergency Director ONLY Proposed Answer: B Person 1, Emergency Director ONLY

Explanation:

Normally, a person 45 or older is selected, however with a dose that exceeds 25 Rem, only a volunteer will be allowed (the selection of 45 or older is a "should" statement). The Shift Manager is initially the Emergency Director. But will all facilities manned, he has turned over the responsibility of being the Emergency Director and thus can no longer approve it.

A. incorrect- see above B. correct- see above C. incorrect- see above D. incorrect- see above Technical Reference{s): EAP-15 Proposed references to be provided to applicants during examination:

Learning Objective: ___ (As available)

Question Source: Bank#

Modified Bank # _~_ (Note changes or attach parent)

New Question History: Last NRC Exam (Optional: Questions validated at the facility since 10195 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of eve!}' question.)

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 55.43 (4)

Comments:

Question #99 Question contains sensitive information - not for public disclosure.

Question #100 The Plant has experienced an emergency radiological event with the following conditions \

timeline:

Time Condition

0800: General Emergency (GE) declared.

0815: Part 1 Notification sent with the following Protective Action Recommendations (ERPAs) circled: 1,2,3,4,5

0900: GE still is still in effect however, environmental conditions now dictate ERPAs 3,4,5,6,7 warrant evacuation.

Which one of the following correctly lists the latest time a new Part 1 Notification is required to be transmitted by and the ERPAs that should be circled?

Latest time ERPAs A. 0915 3,4,5,6,7 B. 0915 1,2,3,4,5,6,7

c. 0930 3,4,5,6,7 D. 0930 1,2,3,4,5,6,7

ES*401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level SRO Tier # -.L Group #

KJA# G 2.4.44 Importance Rating 4.4 Proposed Question: 100 The Plant has experienced an emergency radiological event with the following conditions \

timeline:

Condition

0800: General Emergency (GE) declared.

0815: Part 1 Notification sent with the following Protective Action Recommendations (ERPAs) circled: 1,2,3,4,5

0900: GE still is still in effect however, environmental conditions now dictate ERPAs 3,4,5,6,'7 warrant evacuation.

Which one of the following correctly lists the latest time a new Part 1 Notification is required to be transmitted by and the ERPAs that should be circled?

Latest time ERPAs A. 0915 3,4,5,6,7 B. 0915 1,2,3,4,5,6,7 C. 0930 3,4,5,6,7 D. 0930 1,2,3,4,5,6,7 Proposed Answer: B, 0915, 1,2,3,4,5,6,7

Explanation: From EAP-1.1: Part 1 Notifications are required to be updated every 30 minutes unless: reclassification of EAL, initial PARs or a PAR changes occur. If PARs change, an updated Part 1 is required to be transmitted within 15 minutes. From EAP-4: Do not delete ERPAs from a PAR that previously recommended the ERPA for evacuation or sheltering unless an error was made and recognized prior to the County taking action to implement the PAR.

A. incorrect- see above B. correct- see above C. incorrect- see above D. incorrect- see above Technical Reference(s): _EAP-1.1, EAP-4 _

Proposed references to be provided to applicants during examination:

Learning Objective: ___ (As available)

Question Source: Bank #

Modified Bank #: _ _ _ (Note changes or attach parent)

New X Question History: Last NRC Exam (Optional: Questions validated at the facility since 10195 will generaffy undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of eve/)' question.)

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 (5)

Comments:

v t e Letter Sequence Draft Other
TAC:U01843, (Open)
Initiation Acceptance, Acceptance Administration Questions Answers Results Approval... Other: ML12116A225, ML12116A242, ML12116A252, ML12116A259, ML12116A275, ML122710026, ML122750357, ML122750389, ML12356A250
MONTHYEAR2012-04-03 [Table View]

ML12072A356: Difference between revisions

Latest revision as of 12:13, 10 March 2020

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