Text

Initial Exam 2013-301 Draft RO Written Exam
	ML13156A030
Person / Time
Site:	Turkey Point
Issue date:	06/05/2013
From:	; NRC/RGN-II
To:	; Florida Power & Light Co
References
	Download: ML13156A030 (196)
	v • d • e

Examination Outline Cross-reference: Level RO SRO Tier# 1 Group# 1 KIA# 007 EA1.02 Importance Rating 3.8 Ability to operate and monitor the following as they apply to a reactor trip: MEW System Proposed Question: RO Question # 1 Plant conditions:

Unit 3 has just tripped from 35% power.

Steam Generator Narrow Range Levels are as follows:

3A=5%

3B=6%

3C=7%

AFW is NOT running and CANNOT be started.

The crew is addressing the Steam Generator Levels in 3-EOP-ES-0.1, Reactor Trip Response.

Which ONE of the following describes actions required regarding the operation of the Main Feedwater System?

Establish greater than 400 gpm total feed flow to the Steam Generators and maintain this flow rate until...

A. the next procedural check of Steam Generator levels; THEN Establish feed flow to maintain Steam Generator Levels between 21-50% NR.

B. at least one Steam Generator level is greater than 7% NR; THEN Establish feed flow to maintain Steam Generator Levels between 21-50% NR.

C. the next procedural check of Steam Generator levels; THEN Establish feed flow to maintain all Steam Generator Levels at a MINIMUM of 7% NR.

D. at least one Steam Generator level is greater than 7% NR; THEN Establish feed flow to maintain all Steam Generator Levels at a MINIMUM of 7% NR.

Proposed Answer: B

Explanation (Optional):

A. Incorrect. 400 GPM is required to be maintained until at least one SG is >7% NR, regardless of where the crew is in the procedure. Plausible because its is logical that when you check in a procedure, if 7% is exceeded, the crew would throttle AFW flow B. Correct. The first time 3-EOP-ES-0.1 is checked, the crew checks AFW flow and SG level. The statement for 7% is a continuous action step that requires monitoring of SG level for throttling AFW to the control band of 21-50%.

C. Incorrect. 1 st part wrong, 2 nd part wrong. This is incorrect because the control range is 21-50% not >7%%. This is plausible because the applicant can misinterpret the requirement for AFW flow, knowing that minimum required heat sink is 1 SG at >7%

D. Incorrect. s 1 t part correct, 2nd part wrong. See B and C.

3-EOP-ES-0.1 (step 13)

BD-EOP-ES-0.1 (p22; Rev Technical Reference(s): .

(Attach if not previously provided) 6/13/07)

Proposed References to be provided to applicants during examination: None 6910323 Objective 5, 8 and 10 Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam: NA Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

The KA is matched because the operator must demonstrate the ability to operate and monitor

the MEW System as it applies to a reactor trip.

Examination Outline Cross-reference: Level RO SRO Tier# I Group# 1 KJA# 008 AK2.02 Importance Rating 2.7 Knowledge of the interrelations between the Pressurizer Vapor Space Accident and the following: Sensors and detectors Proposed Question: RO Question # 2 Plant conditions:

Unit 4 tripped from 100% power.

A LOCA is in progress.

The crew is performing 4-EOP-E-1, Loss of Reactor or Secondary Coolant.

RCS pressure is 1000 psig and lowering.

All equipment is operating as designed.

Which of the following identifies the indications seen if the LOCA occurred in the pressurizer vapor space?

A. Pressurizer level trending up or off-scale high; QSPDS RVLMS indication constant throughout the event B. Pressurizer level trending up or off-scale high; QSPDS RVLMS indication lowering throughout the event C. Pressurizer level off-scale low; QSPDS RVLMS indication constant throughout the event D. Pressurizer level off-scale low; QSPDS RVLMS indication lowering throughout the event Proposed Answer: B Explanation (Optional):

A. Incorrect. At 1000 psig, the RCPs are off and the vessel head is hot enough to start forming a bubble and pushing indicated PZR level higher.

B. Correct. With the conditions presented, PZR level will rise because of the bubble being formed under the head. As the bubble under the head grows, RVLMS will show

uncovered thermocouples, resulting in a lowering level for RVLMS C. Incorrect. This is incorrect for conditions given but would be correct for a SBLOCA where SI flow is able to overcome break flow. There would never be indication of RVLMS lowering where the flow from SI was higher than break flow once the pressurizer was emptied and SI either automatically or was manually actuated.

D. Incorrect. This is incorrect because pressurizer level would be rising for the conditions given in the stem of the question. Second part could be true if PZR/RCS pressure continues to lower, meaning that break flow is exceeding injection flow. This would ultimately result in a lowering RVLMS level Lesson Plan 6902918 (p26; Technical Reference(s): (Attach if not previously provided)

Drawing 5614-M-3041, Sheet 1 Proposed References to be provided to applicants during examination: None 6902918 Objectives 2.A and 2.B Learning Objective: (As available)

Question Source: Bank #

Modified Bank # PTN 69028180107 (Note changes or attach parent)

New Question History: Last NRC Exam: NA Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 14 55.43 Principles of heat transfer, thermodynamics and fluid mechanics.

Comments:

The KA is matched because the operator must demonstrate Knowledge of the interrelations between the Pressurizer Vapor Space Accident and plant sensors and detectors, specifically RVLMS and pressurizer level. The question is at the Comprehension/Analysis (3PEO) cognitive level because the operator must recall bits of information (When is Pressurizer level not a measure of RCS mass on a LOCA event? What drives RCS/Pressurizer pressure in a

LOCA event in which the leaking liquid is water?), and then apply this information to a set of plant conditions to answer the question correctly.

Examination Outline Cross-reference: Level RO SRO Tier# 1 Group# 1 KIA# 009 EA2.02 importance Rating 3.5 Ability to determine or interpret the following as they apply to a small break LOCA: Possible leak paths Proposed Question: RO Question # 3 Plant conditions:

Unit 3 has tripped.

Subsequently, a LOCA has developed.

The crew has entered 3-EOP-ECA-1 .2, LOCA Outside Containment.

Which ONE of the following is a likely source of the RCS leak?

A. RCP Thermal Barrier B. CVCS Regenerative Heat Exchanger C. Downstream of MOV 744A, RHR Discharge to Cold Leg Isolation Valve D. Upstream of MOV-3-843A, SI to Cold Leg Isolation Valve Proposed Answer: D Explanation (Optional):

A. Incorrect. Plausible because an applicant may wrongly believe that an inter-system LOCA would constitute a LOCA 0/S Containment. If a TB leak occurred, the leak would auto isolate on high flow. Conditions would not exist for entry to ECA-1 .2.

B. Incorrect. Plausible because an applicant may wrongly believe that an inter-system LOCA would constitute a LOCA 0/S Containment. If a RHX leak occurred, the leak would be in the Letdown System and be contained. Conditions would not exist for entry to ECA-1.2.

C. Incorrect. Plausible because these valves would be manually isolated in ECA-1 .2 as a potential source of leakage. Incorrect because these valves are located in containment and a leak downstream would be contrained, not in Aux Building.

D. Correct. A leak upstream on SI Cold Leg Isolation valves could lead to high radiation in the Aux Building and these valves are listed in ECA-1 .2 for manual isolation

3-EOP-E-0, step 20 Technical Reference(s):

(Attach if not previously provided) 3-EOP-ECA-1 .2, step 2.e Proposed References to be provided to applicants during examination: None

* . 6902212 Objective 7 Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam: NA Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

The KA is matched because the operator must demonstrate the ability to determine or interpret the possible leak paths as they apply to a small break LOCA. This is accomplished by giving the operator a set of plant conditions and requiring the operator to identify the one possible leak path.

The question is at the Comprehension/Analysis (2DR) cognitive level because the operator must recall bits of information and then relate this information to itself by recognizing which procedure is in use and the conditions required to get to that procedure NOTE: The question could be related to 10CFR55.41 (7).

Examination Outline Cross-reference: Level RO SRO Tier# 1 Group# 1 KIA# 011 EK1.O1 Importance Rating 4.1 Knowledge of the operational implications of the following concepts as they apply to the Large Break LOCA: Natural circulation and cooling, including reflux boiling.

Proposed Question: RO Question #4 What is the PRIMARY method of decay heat removal for large break LOCNs?

A. The condensation of reflux boiling in the SIGs B. Heat transfer between the RCS and the SIGs due to natural circulation flow C. The injection of water from the ECCS and leakage of steam/water out the break D. Convection Cooling from the Reactor Vessel to the water in the Containment sumps.

Proposed Answer: C Explanation (Optional):

A. Incorrect. Reflux cooling is a mechanism for cooling during a small break LOCA before the RCS has drained enough to clear the loop seal B. Incorrect. Natural circulation provides a viable method of heat removal post trip and even for the initial phase of a small break LOCA but the primary side of the Steam generators are drained relatively soon in a large break LOCA C. Correct. The background document for E-1 discusses the injection of water from the RWST and the removal of steam/water out the break as the primary method of heat removal for a large break LOCA D. Incorrect. Convection cooling is a method of decay heat removal, and is cited in the SACRG procedure network, but is not the primary heat removal mechanism.

Westinghouse background Technical Reference(s): document for E-1, pgs. 16-17 (Attach if not previously provided)

(4/30/2005)

Proposed References to be provided to applicants during examination: N Learning Objective: (As available)

Question Source: Bank # WTSI 95967 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: 2009 Seabrook Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 14 55.43 Principles of heat transfer, thermodynamics and fluid mechanics.

Comments:

Question matches KA because the applicant must choose RCS heat removal mechanism during a large break LOCA

Examination Outline Cross-reference: Level RO SRO Tier# 1 Group# 1 KIA# 015 AK1.04 Importance Rating 2.9 Knowledge of the operational implications of the following concepts as they apply to the Reactor Coolant Pump Malfunctions (Loss of RC Flow): Basic steady state thermodynamic relationship between RCS loops and SIGs resulting from unbalanced RCS flow Proposed Question: RO Question # 5 Plant conditions:

Unit 3 tripped from 100% power following trip of RCP 3A.

The crew is performing actions of 3-EOP-ES-0.1, Reactor Trip Response.

SG pressures are stable at approximately 1000 PSIG.

Which ONE of the following describes plant status related to the trip of RCP 3A five (5) minutes following the reactor trip?

A. RCS Loop 3A differential temperature will be significantly higher than RCS Loops 3B and 30 differential temperatures due to lower flow through the loop.

B. RCS Loops 3B and 3C differential temperature will be significantly higher than RCS Loop 3A differential temperature due to heat input from the running 3B and 3C RCPs.

C. Feedwater requirements to maintain SGs 3B and 3C levels stable will be significantly higher than requirements to maintain SG 3A level stable because decay heat is being removed from RCS Loops 3B and 3C.

D. Feedwater requirements to maintain SG 3A level stable will be significantly higher than requirements to maintain SGs 3B and 30 levels stable because SG 3A sustained a larger amount of SG Shrink on the reactor trip.

Proposed Answer: C Explanation (Optional):

A. Incorrect. Plausible because if reactor was at power, Loop Delta T would be higher and may show larger difference. Post-Trip, loop Delta T is negligible. Also plausible because if natural circ was setting up in an idle loop, Delta T would be rising

B. Incorrect. Plausible because heat removal requirements are higher for ioop B, but Delta T is negligible for post trip conditions C. Incorrect. Plausible because SG A would have more shrink with reduced heat input from RCP operation, but 5 minutes post trip, AFW will have already stabilized SG level.

The steady state requirement would be higher for the loop with RCP heat input D. Correct. RCP heat is required to be steamed off and AFW flow will be higher in the loop that requires heat removal LP6910916(pgs.5-6 Technical Reference(s): (Attach if not previously provided)

(05/15/12))

Proposed References to be provided to applicants during examination: No Learning Objective: (As available)

Question Source: Bank # 95807 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: 2010 Ginna Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 5 55.43 Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.

Comments:

KA matched because the applicant must choose the correct option for post trip thermodynamic relationship between RCS loops and each SG after an RCP trip

Examination Outline Cross-reference: Level RO SRO ller# 1 Group# 1 KIA# 022 2.1.20 Importance Rating 4.6 Conduct of Operations: Ability to interpret and execute procedure steps.

Proposed Question: RO Question # 6 Plant conditions:

Unit 3 is operating at 100% power.

A loss of all charging flow occurs and attempts to start a charging pump are unsuccessful.

The crew has entered 3-ONOP-047.i, Loss of Charging Flow In Modes i Through 4.

Which ONE of the following identifies the action required in accordance with 3-ONOP-47.i?

A. CLOSE Letdown Orifice Isolation Valves, CV-3-200A, B, C; if charging flow cannot be restored, trip the reactor and enter 3-EOP-E-0, Reactor Trip or Safety Injection.

B. CLOSE Letdown Orifice Isolation Valves, CV-3-200A, B, C; if charging flow cannot be restored, perform a controlled plant shutdown using 3-GOP-i 03, Power Operation to Hot Standby.

C. CLOSE MOV-3-38i, Excess Letdown and RCP Seal Return Valve; if charging flow cannot be restored, trip the reactor and enter 3-EOP-E-0, Reactor Trip or Safety Injection.

D. CLOSE MOV-3-381, Excess Letdown and RCP Seal Return Valve; if charging flow cannot be restored, perform a controlled plant shutdown using 3-GOP-103, Power Operation to Hot Standby.

Proposed Answer: A Explanation (Optional):

A. Correct. In accordance with ONOP-047.i, the crew will close letdown orifice isolation valves in an attempt to maintain RCS inventory B. Incorrect. Actions are correct to isolate the RCS, but if Charging cannot be restored, a reactor trip is required

C. Incorrect. Second half is correct, but closing MOV-3-381 will only partially stop RCS inventory loss by a very small amount. Action to isolate letdown is the critical element to maintaining RCS Inventory D. Incorrect. See explanations for B and C 3-ONOP-047.1 (p5-6; Rev 11/4/08)

Technical Reference(s): BD-ONOP-047.1 (p5; Rev (Attach if not previously provided) 11/4/08)

Proposed References to be provided to applicants during examination: None 6902234 Objectives 4 and 6 Learning Objective: (As available)

PTN 69022340107 Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: NA Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

The KA is matched because the operator must demonstrate the ability to interpret and execute procedure steps in the ONOP associated with a loss of Charging flow.

The question is at the Comprehension/Analysis cognitive level because the operator must recall bits of information (What action required when all Charging flow is lost), and then apply this information to a set of plant conditions, understanding flowpaths of water from the RCS to answer the question correctly.

Examination Outline Cross-reference: Level RO SRO Tier# 1 Group# 1 KIA# 025 AK3.02 Importance Rating 3.3 Knowledge of the reasons for the following responses as they apply to the Loss of Residual Heat Removal System: Isolation of RHR low-pressure piping prior to pressure increase above specified level Proposed Question: RO Question # 7 Plant conditions:

Unit3isin Mode4.

RHR Train 3A is in service.

OMS is in service.

Subsequently:

RCS Pressure rising.

A PZR PORV opens before operators can determine the cause of the pressure increase.

The operator observed RCS pressure peaked at 540 psig and is now lowering.

Which ONE of the following identifies the status of RHR Loop Inlet Isolation Valves MOV-3-750 and MOV-3-751, AND the reason for this condition?

MOV-3-750 and MOV-3-751 are A. fully OPEN because the PZR PORV actuation has terminated the transient in progress.

B. fully OPEN to ensure an adequate pressure relief path exists through letdown.

C. CLOSED or Closing to prevent overpressurization of the RHR System piping.

D. CLOSED or Closing to prevent RHR cooling from exceeding flow limits and exceeding technical specification cooldown rate limits Proposed Answer: C Explanation (Optional):

A. Incorrect. This is not correct because the valves will automatically close under the

stated conditions. This is plausible because if the valves close, NPSH would be lost to the RHR Pumps. In fact, according to 3-ONOP-050 (6; Rev 12/3/07) Step 1.c, the operator will check the RHR pumps immediately, and upon observing that they are cavitating, will stop them. The operator may be unaware of the interlock to close the valves, or believe that they close at a higher pressure. The operator may be unaware of the interlock to close the valves, or believe that they close at a higher pressure.

Additionally, plausibility is gain because according to GOP-305 (p45; Rev 2/23/12) Step 5.19.2, under certain conditions (Mode 5, CM approval) the interlock can be defeated.

Consequently, the operator may incorrectly believe that the interlock is defeated and the valves remain open for the specified reason.

B. Incorrect. This is not correct because the valves will automatically close under the stated conditions. This is plausible because the conditions have established that there is an increase in RCS pressure for unknown reasons, and that closing MOV-3-750 and MOV-3-751 will isolate a pressure relief point via RHR Letdown. Additionally, plausibility is gain because according to GOP-305 (p45; Rev 2/23/1 2) Step 5.19.2, under certain conditions (Mode 5, CM approval) the interlock can be defeated.

Consequently, the operator may incorrectly believe that the interlock is defeated and the valves remain open for the specified reason.

C. Correct. According to SD-021 (p27; Rev 11/30/12) MOV-3-750 & 751, provide isolation between the RCS and the RHR pump suctions when the RHR System is not in use.

The RHR System maximum design pressure is 600 PSIG. Both isolation valves and the upstream piping are rated at the RCS design pressure of 2485 PSIG. There are several interlocks associated with these valves : (1) MOV-3-750 cannot be opened and will automatically close when hot leg pressure (loop B) exceeds 515-535 PSIG (as sensed by PC-403), and (2) MOV-3-751 cannot be opened and will automatically close when hot leg pressure (loop A) exceeds 515-535 PSIG (as sensed by PC-405). Since RCS pressure is 540 psig, both valves will be closed or traveling closed in order to protect the RHR System piping from overpressure, which could lead to an intersystem LOCA.

D. Incorrect. This is not correct because the valves will not auto close to avoid a PTS condition. This is plausible because the RHR System is associated with RCS cooling, and there are limits placed on the allowable flow within the system. The operator may incorrectly believe that the overpressure condition will lead to overflow condition leading to a overcooling event.

SD-021 (p27; Rev 11/30/11)

ONOP-050 (6 Rev 12/3/07)

Technical Reference(s): . .

(Attach if not previously provided)

GOP-305 (p47; Rev 8/4/1 2)

Proposed References to be provided to applicants during examination: None Learning Objective: 6902121A Objectives 8.b and 9.f (As available)

Question Source: Bank# PTN 69021210628 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: NA Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Comments:

The KA is matched because the operator must demonstrate Knowledge of the reasons for the isolation of RHR low-pressure piping prior to pressure increase above specified level as they apply to the Loss of Residual Heat Removal System.

The question is at the Memory (il/F) cognitive level because the operator must recall bits of information (valve interlock, basis for interlock) to answer the question correctly.

Examination Outline Cross-reference: Level RO SRO Tier# 1 Group# 1 KIA# 027 2.1.28 Importance Rating 4.1 Conduct of Operations: Knowledge of the purpose and function of major system components and controls.

Proposed Question: RO Question # 8 Plant conditions:

Both units are at 100% power.

A fire occurred in the 4B 4KV Switchgear Room and has been extinguished.

Unit 4 RCS pressure began to slowly decrease, and the crew entered 4-ONOP-041 .5, Pressurizer Pressure Control Malfunction.

CV-4-31 1, Auxiliary Spray Isolation Valve, has BOTH opened and closed (dual light) indication in the Control Room, and the crew believes that the fire has caused this valve to inadvertently open.

All Pressurizer Heaters are energized, and both PZR Spray valves are in MANUAL and CLOSED.

RCS pressure is currently 2160 psig, and lowering at 10 psi per minute.

The 4B and 4C Charging Pumps are running.

The 4A Charging Pump is OFF.

Which ONE of the following identifies an action that is required by 4-ONOP-041 .5?

A. Trip the reactor and go to E-0, Reactor Trip or Safety Injection.

B. Ensure the 4B and 4C Charging Pumps B are operating at minimum speed and locally vent instrument air from CV-4-31 1.

C. Reduce Charging to one pump operating at minimum speed and close Charging to RCS Control Valve HCV-4-121.

D. Reduce Charging to one pump operating at minimum speed and open Alternate Charging Isolation Valve, CV-4-3 1 OB.

Proposed Answer: C Explanation (Optional):

A. Incorrect. This is incorrect because under the current conditions a reactor trip is not required. This is plausible because according to 4-ONOP-041 .5 (p11 & Foldout; Rev 12/17/12) if Pzr pressure cannot be maintained > 2000 psig the operator must trip the reactor and go to E-0. The operator may incorrectly believe that the threshold is higher, such as 2200 psig, or interpret the stated conditions in such a way that causes the incorrect belief that these conditions will be achieved before any other mitigation action has a chance to correct the current course of current plant conditions.

B. Incorrect. This is incorrect because ONOP-041 .5 requires that Charging Pumps be reduced to one operating at slow speed, and does not direct the operator to vent instrument air from CV-4-31 1. This is plausible because the operator may not know that the Foldout Page requires single Charging Pump operation at slow speed. Additionally, according to SD-013 (p95; Rev 9/20/11) CV-4-31 1 fails closed on a loss of air, and if this action can be taken, the RCS pressure trend will stabilize. However, this valve is in the Containment and will require an entry to made, and it is NOT directed by 4-ONOP-041.

C. Correct. According to 4-ONOP-041 .5 (p 1 5; Rev 12/17/12) Steps 18-19, the operator will evaluate if PZR Spray Valve leakage is preventing RCS Pressure stabilization, If so, which is the case under the stated conditions, the operator is directed to verify that the PZR Spray valves are closed. If the valve cannot be closed, RCS pressure will continue to lower, as is the case in the stated conditions. According to the Foldout Page (pFoldout; Rev 12/1 7/1 2) Item 5.a, IF pressurizer pressure is decreasing and Auxiliary Spray Valve, CV-4-31 1, is suspect, THEN reduce charging to one charging pump on slow speed AND close charging to RCS Control Valve HCV-4-121.

D. Incorrect. This is incorrect because ONOP-041 .5 does not direct the operator to open CV-4-310B. This is plausible because according to SD-013 (p12-13; Rev 9/20/11), CV-4-11 OB is a 3 line rather than the 2 line associated with CV-4-31 1. The operator may incorrectly believe that the procedure directs this action to create a path of least resistance allowing flow into the RCS through the Charging connection rather than the Pzr Steam Space. However, this action is NOT directed by 4-ONOP-041.

4-ONOP-041.5 (p15 & Foldout;

Rev 12/17/12)

Technical Reference(s): * *

(Attach if not previously provided) sD-01 3 (p12-13; Rev 9/20/11)

Proposed References to be provided to applicants during examination: None

6918204 Objective 7 Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam: NA Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

The KA is matched because the operator must demonstrate Knowledge of the purpose and function of major system components (HCV-1 21) and controls (Reduce Speed of Charging Pump).

The question is at the Comprehension/Analysis (2DR) cognitive level because the operator must recall bits (ONOP-041 .5 Foldout Page Items) of information, and then apply this information to a set of plant conditions to answer the question correctly.

Examination Outline Cross-reference: Level RO SRO Tier# 1 Group# 1 K/A# 029 EK2.06 Importance Rating 2.9 Knowledge of the interrelations between the following and ATWS: Breakers, relays, and disconnects Proposed Question: RD Question # 9 Which ONE of the following describes how the AMSAC (ATWS Mitigating System Actuation Circuit) trips the reactor?

A. Energizes both Control Rod MG set input breaker trip coils.

B. Energizes both Control Rod MG set output breaker trip coils.

C. Energizes the Shunt Trip Coils on both Reactor Trip Breakers.

D. Deenergizes the Undervoltage Trip Coils on the Reactor Trip Breakers and Bypass Breakers.

Proposed Answer: B Explanation (Optional):

A. Incorrect. This is incorrect because AMSAC does not affect the Control Rod MG set input breakers. This is plausible because according to EOP-FR-S.1, Step 7, the NB MG set motor input breakers is one set of four sets of breakers that are identified as breakers to locally trip in the event that the reactor will not rip form the Control Room.

The operator may incorrectly believe that AMSAC operates the input breakers rather than the output breakers.

B. Correct. According to Drawing 5614-T-L1 Sheets 33A/B and SD-063 (p77; Rev 9/10/11), when AMSAC actuates the reactor is tripped by energizing the control rod MG set Output Breaker Trip Coil for both the A and B Control Rod MG sets.

C. Incorrect. This is incorrect because AMSAC does not affect the Shunt Trip Coils on both Reactor Trip Breakers. This is plausible because the Reactor Protection System uses UV Coils and Shunt Trip Coils for the Reactor Trip and Bypass Breakers to trip the reactor, rather than tripping breakers associated with the Control Rod MG sets. The operator may incorrect believe that the AMSAC trips the reactor via a subset of the RPS.

D. Incorrect. This is incorrect because AMSAC does not affect the Undervoltage Trip Coils on the Reactor Trip Breakers and Bypass Breakers. This is plausible because the Reactor Protection System uses UV Coils and Shunt Trip Coils for the Reactor Trip and Bypass Breakers to trip the reactor, rather than tripping breakers associated with the Control Rod MG sets. The operator may incorrect believe that the AMSAC trips the reactor via a subset of the RPS.

Drawing 5614-T-L1 Sheets Technical Reference(s): (Attach if not previously provided)

SD-063 Rev 9/10/11)

Proposed References to be provided to applicants during examination: None 6902163 Objectives 5.c, 5.d and 9 Learning Objective: (As available)

Question Source: Bank # PTN 69021630658 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: NA Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Comments:

The KA is matched because the operator must demonstrate Knowledge of the interrelations between breakers Reactor Trip Breakers), relays (Shunt, UV Coils), and disconnects (MG Set breakers) used in the ATWS event. This is accomplished by requiring the operator to identify how a circuit (AMSAC) designed specifically for protecting the reactor against the ATWS event will trip the reactor when it is actuated.

The question is at the Memory (1 F) cognitive level because the operator must recall bits of

information (How the AMSAC trips the reactor) to answer the question correctly.

Examination Outline Cross-reference: Level RO SRO Tier# 1 Group# 1 KIA# 038 EKI.02 Importance Rating 3.2 Knowledge of the operational implications of the following concepts as they apply to the SGTR:

Leak rate vs. pressure drop Proposed Question: RO Question # 10 Plant conditions:

A Steam Generator Tube Rupture has occurred on 3C S/G.

RCS cooldown and depressurization is complete.

Preparations are made to transition to 3-EOP-ES-3.1, Post SGTR Cooldown using Backfill.

Pressurizer Level is 38%.

All RCPs are OFF, and unavailable to start.

RCS Subcooling is 43° F.

3A Charging Pump is running.

Letdown is45gpm.

3C S/C Narrow Range Level is 79% and slowly rising.

Which ONE of the following identifies the required action and the reason for that action in accordance with 3-EOP-E-3, Steam Generator Tube Rupture?

OPEN...

A. PCV-3-455A or PCV-3-455B, Pressurizer Spray valve, to refill the PZR.

B. PCV-3-455A or PCV-3-455B, Pressurizer Spray valve, to minimize RCS leakage.

C. CV-3-31 1, Auxiliary Spray Isolation Valve to refill the PZR.

D. CV-3-31 1, Auxiliary Spray Isolation Valve to minimize RCS leakage.

Proposed Answer: D Explanation (Optional):

A. Incorrect. This is incorrect because the Pzr spray cannot be used under the stated conditions, (RCPs off) and the depressurization is not done to raise Pzr level. This is plausible because if RCPs were in service, the spray would be used; and because an

earlier depressurization in the SGTR strategy is done in part to raise PZR level. (refill pzr) According to 3-EOP-E-3 (p 1 8; Rev 4) Step 22, the operator is directed to depressurize the RCS to minimize break flow and refill the pressurizer. The operator may confuse the reasons for the different depressurizations.

B. Incorrect. This is incorrect because the Pzr spray valves are not used under the stated conditions. This is plausible because if RCPs were in service, this would be the correct response.

C. Incorrect. This is incorrect because the depressurization is not done to refill the PZR.

This is plausible because an earlier depressurization in the SGTR strategy is done in part to raise PZR level. According to 3-EOP-E-3 (p18; Rev 4) Step 22, the operator is directed to depressurize the RCS to minimize break flow and refill the pressurizer. The operator may confuse the reasons for the different depressurizations.

D. Correct. According to 3-EOP-E-3 (p24; Rev 4) Step 34 provides the operator direction to control RCS pressure and Charging flow to minimize RCS-to-Secondary Leakage.

Because the content of this step is placed in a separate Attachment (Attachment 2), it is clear that this action is a Continuous Action, and will be implemented as needed, as the crew continues with the procedure. The selection of the recovery procedure is the last step in the body of the procedure, and yet, this action is still applicable. Action is taken by the operator based on two key plant parameters, Pzr Level and Ruptured SG Level.

Based on these parameters, the operator must depressurize the RCS, while referring to a Note prior to the AttachmentlStep 34 Table. The Note reads as follows: When RCS depressurization is required, normal spray should be used whenever possible. If normal spray is NOT available and letdown is in service, auxiliary spray should be used. If normal spray and auxiliary spray are NOT available, one PRZ PORV should be used.

Since no RCPs are running or available to start, normal spray is not an option. Since Letdown is 45 gpm, Auxiliary Spray is the preferred method of pressure reduction over the PORV.

3-EOP-E-3 (p18 & p24; Rev 4)

Technical Reference(s):

(Attach if not previously provided)

Proposed References to be provided to applicants during examination: None 6910339 Objective 5.1 Learning Objective: (As available)

Question Source: Bank #

Modified Bank # X (Note changes or attach parent)

New

Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

The KA is matched because the operator must demonstrate Knowledge of the operational implications of the Leak rate vs. pressure drop concepts as they apply to the SGTR. This is accomplished by requiring that when given a set of point post-RCS cooldown/depressurization conditions in a SGTR, the operator identify the preferred method for conducting the depressurization and the basis for it.

The question is at the Comprehension/Analysis (3PEO) cognitive level because the operator must recall bits of information (Preferred means of depressurizing), and then apply this information to a set of plant conditions to answer the question correctly.

The question is significantly modified because at least one significant change was made in the conditions (No RCPs Running, Letdown in service); and two answers, including the correct are different from the parent question. According to NUREG-1021, ES-401 Section D.2.f, paragraph 1, bullet 4; to be considered a significantly modified question, at least one pertinent condition in the stem and at least one distractor must be changed from the original bank question. Changing the conditions in the stem such that one of the three distractors in the original question becomes the correct answer would also be considered a significant modification.

NOTE: The parent question was used as Question 10 on the 2011 Turkey Point NRC Exam.

Examination Outline Cross-reference: Level RO SRO Tier# 1 Group# 1 KIA# 040 AAI.19 Importance Rating 3.8 Ability to operate and I or monitor the following as they apply to the Steam Line Rupture:

Postaccident monitoring panel indicators Proposed Question: RO Question # I I A large Main Steam Line Break has occurred on Unit 3.

Which ONE of the following identifies an instrument listed in Technical Specification 3.3.3.3, Accident Monitoring Instrumentation that will be used to determine if a Pressurized Thermal Shock condition exists in accordance with 3-EOP-F-0, Critical Safety Function Status Trees?

A. RCS Wide Range Thot B. RCS Wide Range Tcold C. Steam Generator Wide Range Pressure D. Core Exit Thermocouples Proposed Answer: B Explanation (Optional):

A. Incorrect. Thot is used in EOPs, specifically in ECA-2.1 which may lead to a PTS condition. It is also a monitor required by TS 3.3.3.3 B. Correct. Tcold is required by TS 3.3.3.3 and is also evaluated for PTS in the CSF Status Trees C. Incorrect. SG WR pressure will give indication of a main steam line break (decreasing uncontrollably) but is not considered for TS 3.3.3.3 and F-0 for CSF status.

D. Incorrect. CETs are accident monitoring instruments, but F-0 does not require this parameter for the Integrity CSF status tree. It is required for the Core Cooling CSF Status tree

3-EOP-F-0 (plO; Rev 7/17/1 2)

Technical Reference(s): TS 3.3.3.3 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None 6910334 Objective 5 Learning Objective: 6910353 Objective 2 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam: NA Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

The KA is matched because the operator must demonstrate the ability to monitor post-accident monitoring panel indicators as they apply to the Steam Line Rupture. According to TS LCO 3.3.3.3 both the RCS Wide Range Thot, Tcold, and CETs are Accident Monitoring Instrumentation.

Examination Outline Cross-reference: Level RO SRO Tier# 1 Group# 1 K/A # 054 AK3.04 Importance Rating 4.4 Knowledge of the reasons for the following responses as they apply to the Loss of Main Feedwater (MFW): Actions contained in EOPs for loss of MFW Proposed Question: RO Question # 12 Which ONE of the following is the primary reason for stopping all RCPs in 3-EOP-FR-H.1, Response to Loss of Secondary Heat Sink?

A. To establish natural circulation conditions and mitigate the transient by establishing a delta T across the core.

B. To reduce the heat added from the RCPs, thereby delaying the need for bleed and feed and gaining time to establish a means of supplying FW to a SIG.

C. To prevent the heat added by the RCPs from adversely affecting indications used to determine whether or not RCS bleed and feed will be required.

D. To reduce RCS pressure enough to ensure bleed and feed is adequate for RCS cooling requirements.

Proposed Answer: B Explanation (Optional):

A. Incorrect, because Natural Circulation will not mitigate the event. Natural circulation will be inhibited in FR-H.1 due to the loss of secondary mass in the SGs B. CORRECT. According to BD-EOP-FR-H.1 (p14.; Rev 6/1 5/1 2) RCP operation results in heat addition to the RCS water. By stopping the RCPs, the effectiveness of the remaining water inventory in the S/Cs is extended, which extends the time at which the operator action to initiate bleed and feed must occur. This extension of time is additional time for the operator to restore feedwater flow to the S/Cs.

C. Incorrect because although RCPs do affect heat input, they do not affect the indication for determination of bleed and feed requirements. (RCS pressure and SC level)

D. Incorrect because the action to open PORVs will reduce RCS pressure enough to ensure adequate flow. Tripping RCPs does not reduce RCS pressure, with the

exception of the head delivered by the RCP BD-EOP-FR-H.1 (p14; Rev Technical Reference(s): 6/1 5/1 2) (Attach if not previously provided)

Proposed References to be provided to applicants during examination: No Learning Objective: (As available)

Question Source: Bank# 91919 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: 2006 VC Summer Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

KA is matched because the applicant must choose the reason for an action that is taken (Trip RCPs) during a loss of Main Feedwater/Secondary Heat Sink Question is at Memory level because the applicant must recall the reason for an action that is taken in the EOP network

Examination Outline Cross-reference: Level RO SRO Tier# 1 Group# 1 KIA# 056 2.4.11 Importance Rating 4.0 Emergency Procedures I Plan: Knowledge of abnormal condition procedures.

Proposed Question: RO Question # 13 3A 4KV Bus is de-energized.

Which ONE of the following is the preferred source of power to re-energize 3A 4KV bus in accordance with 3-ONOP-004.2, Loss of 3A 4KV Bus?

A. 3C Bus B. SBO Tie Line C. Unit 3 Startup Transformer D. Unit 4 Startup Transformer Proposed Answer: C Explanation (Optional):

A. Incorrect. 3C bus would be used to re-energize 3A bus if other power was not available from startup transformers and SBO tie line was not available due to 4kv bus 3B being energized.

B. Incorrect. SBO Tie line would be used in the event of a station blackout but would not be the preferred source of power for restoration for just a de-energized bus.

C. Correct. This is the preferred source. If off-site power was lost this source would not be available, but since the initial condition is a loss of bus 3A only, this is the first source used to attempt a re-energization.

D. Incorrect. This would be the preferred source of power for re-energization if the Unit 3 startup transformer was available

Technical Reference(s): 3-ONOP-004.2 (p6-iD, Rev 2) (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None 6902113 Objective 4.a 6902138 Objective 4.q Learning Objective: (As available) 6918256 Objectives 4 and 6 Question Source: Bank # 69022560103 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: NA Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

The KA is matched because the operator must demonstrate Knowledge of abnormal condition procedures, specifically 3-ONOP-004.2, which would be implemented following a Loss of the 3A 4KV bus.

The question is at the Comprehension/Analysis (2R1) cognitive level because the operator must match a specific condition with an ONOP that specifies power restoration based on equipment that is available

Examination Outline Cross-reference: Level RO SRO Tier# 1 Group# 1 K/A# 057 AAI.04 Importance Rating 3.5 Ability to operate and / or monitor the following as they apply to the Loss of Vital AC Instrument Bus: RWST and VCT valves Proposed Question: RO Question # 14 Plant conditions:

Unit 3 has lost 120 Volt Vital Instrument Panel 3P07.

VCT Level Indicator LI-3-112 indicates 25%.

VCT Level indicator LI-3-1 15 indicates 0%.

Annunciator A 4/6, VCT HI/LO LEVEL, is in alarm.

Which ONE of the following describes the effect on VCT auto makeup and charging pump suction alignment?

VCT auto makeup A. initiates and charging pump suction remains aligned to the VCT.

B. initiates and charging pump suction automatically swaps to the RWST.

C. is disabled and charging pump suction remains aligned to the VCT.

D. is disabled and charging pump suction automatically swaps to the RWST.

Proposed Answer: A Explanation (Optional):

A. Correct. 1 st part is correct, 2nd part correct. According to SD-01 3 (p45; Rev 9/20/11) and 561 0-T-D-1 9 sheet 1, the VCT has two level detectors, LT-3-1 12 and LT3-3-1 15.

According to SD-013 (p45; Rev 9/20/11) and 5610-T-D-19 sheet 1, during normal operation, LC-3-1 150 initiates automatic makeup on a low level in the VCT. A preselected blend of boric acid and pure water is injected into the VCT outlet line.

Automatic makeup is initiated when VCT level, as sensed by LT-3-1 15, reaches 17%

decreasing and stops when level reaches 31% increasing. A failure of the VCT level control loop 115 low would result in auto makeup starting and staying on. According to

3-ONOP-003.7, Enclosure 1 (plO; Rev 11/1/10), on a loss of 3P07, LT-3-115 indication will fail low, and auto VCT makeup will occur.

B. Incorrect. 1 st part is correct, 2 nd part wrong. This is incorrect because the Charging Pump suction will not swap to the RWST. This is plausible because according to SD-013 (p27-p28; Rev 9/20/Il), VCT Outlet Isolation Valve LCV-3-1 1 5C is a 4, motor operated, gate valve. It receives its electrical power from 480V MCC 3B. Valve operation is through a CLOSE-AUTO-OPEN, spring-return to AUTO switch on VPA.

According to SD-013 (p29; Rev 9/20/11), the suction of the Charging Pumps can be supplied from two connections from the RWST. The first is through LCV-3-l 1 5B, an air to open, fail closed, 4, solenoid actuated butterfly valve. The second is a manual valve (3-358). According to SD-013 (p44-p45; Rev 9/20/11) and 5610-T-D-19 sheet 1, the VCT has two level detectors, LT-3-1 12 and LT-3-1 15. If the level falls to 4%, the emergency low level, a signal from both channels opens LCV-3-1 15B and closes LCV 3-1 1 5C. When level is restored to 11% increasing, the charging pump suction is automatically realigned to the VCT. LT-3-112 is at 25% and rising with makeup.

Swapover will not occur. The operator may incorrectly believe that the logic requires only one instrument below setpoint.

C. Incorrect. 1 st part is wrong, 2nd part correct. This is incorrect because auto makeup is NOT disabled. This is plausible because the operator may incorrectly believe that this is an energize-to-function signal associated with LT-3-1 15, in which case the makeup would not initiate without electrical power.

D. Incorrect. 1 st part is wrong, 2nd part wrong. See B and C.

SD-013 (p27-29, 44-45; Rev 9/20/11) 5610-T-D-l9sheetl Technical Reference(s): .

(Attach if not previously provided) 3-ONOP-003.7, Enclosure 1 (plO; Rev 11/1/1 0)

Proposed References to be provided to applicants during examination: N 6902113 Objectives 5.e, 6, 7.c, 8.f Learning Objective: and 11 .a (As available)

WTSI 7125 Question Source: Bank #

PTN 69022600313 Modified Bank # (Note changes or attach parent)

New

Question History: Last NRC Exam: 2009-2 Turkey Point (Not last 2)

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Comments:

The KA is matched because the operator must demonstrate the ability to monitor the RWST and VCT valves as they apply to the Loss of Vital AC Instrument Bus. This is accomplished by establishing that one Vital AC Bus has de-energized and requiring the operator indicate the impact on the automatic swapover of the Charging Pump suction from the VCT to the RWST.

The question is at the Comprehension/Analysis cognitive level because the operator must recall bits of information (i.e. VCT Instrument logic!setpoints, how auto rn/u signal generated) and choose between plausible alternatives to answer the question correctly.

Examination Outline Cross-reference: Level RO SRO Tier# 1 Group# 1 KIA# 062 AA2.06 Importance Rating 2.8 Ability to determine and interpret the following as they apply to the Loss of Nuclear Service Water: The length of time after the loss of SWS flow to a component before that component may be damaged Proposed Question: RO Question # 15 Plant conditions:

Unit 4 is operating at 100% power.

An operator has been dispatched to isolate CCW/ICW strainers due to suspected fouling.

While isolating BS-4-1402, 4A ICW/CCW Basket Strainer, ICW flow to all 3 CCW heat exchangers is determined to be less than the minimum required in accordance with 4-NOP-019, Intake Cooling Water System.

In accordance with 4-NOP-019, which ONE of the following identifies (1)the minimum ICW flow to each CCW heat exchanger, and (2) the amount of time allowed at less than minimum required flow prior to entering T.S. 3.0.3?

A. (1)3500GPM (2) One hundred twenty (120) minutes B. (1)3500GPM (2) Five (5) minutes C. (1)11,000GPM (2) One hundred twenty (120) minutes D. (1) 11,000 GPM (2) Five (5) minutes Proposed Answer: B Explanation (Optional):

A. Incorrect. 1 st part correct, 2nd part wrong. 120 minutes is plausible based upon ICW operability after a loss of instrument air

B. Correct. If below minimum flow through CCW heat exchangers for >5 mm, TS 3.0.3 must be entered to make plans for shutdown to hot standby C. Incorrect. 1st part wrong, 2nd part wrong. 1 1,000 GPM is plausible because that is minimum required total flow for lOW through COW heat exchangers. 120 minutes plausible for same reason as in A D. Incorrect. See explanation in C for first part. Second part of this option is correct 4-ONOP-019 (Foldout; Rev 09/14/08)

Technical Reference(s): .

(Attach if not previously provided)

Proposed References to be provided to applicants during examination: None 6918277 Objectives4and6 Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam: NA Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

The KA is indirectly matched because the operator must demonstrate the ability to determine and interpret the length of time after the loss of SWS flow to a component before that component may be damaged as they apply to the Loss of Nuclear Service Water. This is accomplished by placing the operator in a situation where a strainer is isolated and the applicant must apply foldout page criteria of ONOP-019 to the situation.

NOTE: The difficulty with directly matching the KA, is that there is no reference information that

identifies a definitive time period when an lOW Load will overheat. However, there is definitive information regarding the situation presented and a Time Critical Action is based on this.

Examination Outline Cross-reference: Level RO SRO Tier# 1 Group# 1 KIA# 065 AK3.03 Importance Rating 2.9 Knowledge of the reasons for the following responses as they apply to the Loss of Instrument Air: Knowing effects on plant operation of isolating certain equipment from instrument air Proposed Question: RO Question # 16 Plant conditions:

Both units are operating at 100% power.

Unit 3 Instrument Air pressure drops to 75 psig as indicated on PI-3-144A (VPA).

The crew enters 0-ONOP-Ol 3, Loss of Instrument Air.

The crew determines that an Auxiliary Building Header rupture has occurred and closes 3-40-339, Auxiliary Building Header Isolation Valve.

Unit 3 Instrument Air pressure stabilizes.

Which ONE of the following identifies whether or not a reactor trip is required, and the reason for the action taken?

A. Reactor trip is required; Instrument Air Header pressure is below reactor trip criteria.

B. Reactor trip is not required; Instrument Air Header pressure is stable and components have not repositioned.

C. Reactor trip is required; failure of critical components in the Auxiliary Building have the potential to place the unit in a more severe transient.

D. Reactor trip is not required; Instrument Air pressure is stable and air operated components in the Auxiliary Building may be operated manually Proposed Answer: C Explanation (Optional):

A. Incorrect. This is incorrect because there is no direction to the crew to trip the plant in anticipation at the current header pressure. This is plausible because according to the 0-ONOP-013 Foldout Page Major Component Impact Item 1, Bullet 2 (Foldout; Rev 11/30/11), when IA pressure is less than 90 psig, the MSIVs at U3 ONLY may result in a

loss of function if the N2 Backup Systems cannot be maintained. The operator may incorrectly believe that this information is directing the crew to trip the plant.

B. Incorrect. This is incorrect because a plant trip is required both by steps in the body of the procedure and per Foldout Page direction. This is plausible because this would be the correct response if the leak were in the Intake Area or the Steam Dump to Condenser Header. Both of these are areas in which the operator will be directed to look for leaks in Step 16 of 0-ONOP-013, but neither are bound by the preceding Caution as is a leak in the Auxiliary Building Header.

C. Correct. According to 0-ONOP-013 (p23; Rev 4) Step 16, the operator is directed to check if the Instrument Air Line Leak is identified. If it is not, the RNO must be performed. One of the RNO actions is to close the Auxiliary Building Header Isolation Valve, and to leave it closed if it isolates the leak. A caution provided prior to step 16 states that if a main air header isolation valve to the Auxiliary Building is required to be closed to isolate a leak or line rupture, then the affected unit is required to be tripped, and its associated 314-EOP-E-0 entered prior to proceeding. This requirement is also repeated on the Foldout Page (Foldout; Rev 4). According to 0-BD-ONOP-013 (p6; Rev 12/23/02), this action is taken because the compounded failures place the plant in a condition outside the operating design basis and a unit trip is required to minimize the impacts of the failures.

D. Incorrect. This is incorrect because a plant trip is required both by steps in the body of the procedure and per Foldout Page direction. This is plausible because the operator may know that the plant trip is not required until header pressure drops to 65 psig, but be unaware of the requirement to trip when the Aux Building Header is isolated. If so, the operator may incorrectly believe that the purpose of the Major Component Impact statements on the Foldout Page is to have the operator consider plant vulnerabilities when IA pressure is between 75-90 psig. The ability to operate components manually is not considered in the background doc 0-ONOP-013 (pZ3, 28 &

Foldout; Rev 4)

Technical Reference(s): 0-BD-ONOP-013 (p6; Rev (Attach if not previously provided) 12/23/02)

Proposed References to be provided to applicants during examination: None 6910286 Objectives4and6 Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam: NA Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

The KA is matched because the operator must demonstrate Knowledge of the reasons for the knowing effects on plant operation of isolating certain equipment from instrument air as they apply to the Loss of Instrument Air. This is accomplished by requiring that the operator know whether or not a plant trip is required if a certain IA Header is isolated during the performance of the Loss of lAS ONOP; and requiring that they identify the reason for this response.

The question is at the Comprehension/Analysis (2DR) cognitive level because the operator must recall bits of information (Thresholds for Plant Trip not based on a memorized setpoint),

and then recognize the relationship between the action and the basis for the action (i.e. Why the trip is needed) to answer the question correctly.

Examination Outline Cross-reference: Level RO SRO Tier# 1 Group# 1 KIA# E04 EK2.2 Importance Rating 3.8 Knowledge of the interrelations between the (LOCA Outside Containment) and the following:

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

Proposed Question: RO Question # 17 Unit 3 was initially at 100% power and has experienced the following events:

The crew has responded to a LOCA into the Auxiliary Building using 3-EOP-ECA-1 .2, LOCA Outside Containment.

The reported leakage was from the RHR Heat Exchanger Room and was stopped by closing RHR Cold Leg Injection Valves 3-MOV-744A and 3-MOV-744B.

RCS pressure is now rising.

Which ONE of the following describes the RCS decay heat removal methodology which will be in place when the crew exits 3-EOP-ECA-1 .2?

A. AFW supplying steam generators, dumping steam.

B. RCS Cooldown using PZR PORVs with one train of ECCS injecting.

C. RHR normal cooling lineup.

D. RCS cooldown using break flow with one train of ECCS injecting.

Proposed Answer: A Explanation (Optional):

A. Correct. Heat sink will be established in 3-EOP-E-0 using AFW and either condenser or atmospheric dump valves. That will be unaffected by actions taken in 3-EOP-ECA-1 .2.

B. Incorrect. Applicant may misapply depressurization/cooldown actions required by subsequent procedures, such as remainder of 3-EOP- E-1 and transition to 3-EOP- ES-1.2. By the time 3-EOP-ECA-1 .2 is complete, these actions have not been considered

C. Incorrect. Cannot use normal RHR cooling until RCS is depressurized to the RHR entry conditions. Applicant may also misapply RHR flowpath, not recognizing that MOV-744A and B are required for cooldown flow path as well as injection D. Incorrect. SI flow is not reduced in 3-EOP-ECA-1 .2, it will be reduced in downstream EOPs such as 3-EOP-ES-1 .2. RCS cooldown will happen because of ECCS injection, but will be controlled by secondary heat sink as long as RCS pressure remains above SG pressure 3-EOP-E-0 (p 1 1-12; Rev 6/16/12)

Technical Reference(s): BD-EOP-E-0 (19-21; Rev (Attach if not previously provided) 6/1 6/1 2)

Proposed References to be provided to applicants during examination: No Learning Objective: LP 6910321 obj 5 (As available)

Question Source: Bank # 92172 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: 2004 Sequoyah Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

KA is matched because it requires the applicant to consider the situation and what led to entry to the EOP and plant status based upon actions previously taken, taken in the EOP, and resulting plant condition.

Question is at comprehension/analysis level because the applicant must consider actions that are taken in previous procedures as well as the procedure identified and determine plant condition.

Examination Outline Cross-reference: Level RO SRO Tier# 1 Group# 1 KIA# E05 EA2.2 Importance Rating 3.7 Ability to determine and interpret the following as they apply to the (Loss of Secondary Heat Sink) Adherence to appropriate procedures and operation within the limitations in the facilitys license and amendments.

Proposed Question: RO Question # 18 Plant conditions:

A Steam Line Break outside Containment resulted in a Reactor Trip and Safety Injection on Unit 3.

ALL Auxiliary Feedwater Pumps are TRIPPED.

DWDS-3-012, SSGFP DISCH TO UNIT 3 ISOL, is Danger Tagged CLOSED.

All Main Steam Isolation valves are closed.

The operating crew has implemented 3-EOP-FR-H.1, Response to Loss of Secondary Heat Sink.

Restoration attempts are in progress to initiate AFW flow.

RCS Pressure is 2255 psig and slowly rising.

All RCPs are TRIPPED.

Wide Range SIG levels are:

3A-45%

3B-46%

3C-44%.

Based on plant conditions, which ONE of the following identifies the NEXT mitigation strategy to be used in 3-EOP-FR-H.1 to restore AFW flow?

Establish...

A. Main Feedwater flow not to exceed 100 gpm per SG B. Main Feedwater Flow at maximum rate to restore heat sink.

C. Standby Feedwater Flow not to exceed 100 gpm per SG.

D. Standby Feedwater Flow at maximum rate to restore heat sink

Proposed Answer: B Explanation (Optional):

A. incorrect. This is incorrect because there is no flow limitation for feedwater based upon the current SG levels B. Correct. According to 3-EOP-FR-H.1 (p7-16; Rev 6/1 5/1 2), Step 2 will direct the operator to Try to establish AFW flow to at least one SG. Since the conditions indicate that attempts to restore AFW are in progress, it is acceptable to continue with the procedure. A caution prior to step 2 directs the operator to initiate bleed and feed under specific conditions. The Caution states that if wide range level in any SIG is less than 33% [narrow range level in all S/Gs less than 27%] or PRZ pressure is greater than or equal to 2335 psig due to loss of secondary heat sink, Steps 11 through 19 should be initiated immediately for bleed and feed. None of the three SGs are < 33% Wide Range; nor is RCS pressure is> 2335 psig. Consequently, Bleed and Feed criteria are NOT met. With this in mind, the operator must look for another means to establish a secondary heat sink. The order of preference is identified in Steps 4 (MFW), 6 (SBFW),

and 8 (Condensate). Consequently, the next method of heat sink restoration is using the MFW System.

C. incorrect. This is incorrect because MEW is preferred over using either the Standby Eeedwater System or the Condensate System. Additionally, even if this were the next mitigation strategy used, DWD-3-012 cannot be opened, and therefore this method is not available. This is plausible because the procedure identifies the Standby Feedwater System as a potential method to re-establish a secondary heat sink. The operator may miss the fact that the method is not available and incorrectly believe that this is preferred over the MEW System under the stated conditions.

D. Incorrect. This is incorrect because MEW is preferred over using either the Standby Feedwater System or the Condensate System. This is plausible for same reason as C but in this option, the flow is correct 3-EOP-ER-H.1 (p7-12; Rev Technical Reference(s): 6/1 5/1 2) (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None 6902337, Objective 5 Learning Objective: (As available)

Question Source: Bank #

Modified Bank # X (Note changes or attach parent)

New Question History: Last NRC Exam: NA Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

The KA is matched because the operator must demonstrate the Ability to adhere to appropriate procedures and operation within the limitations in the facilitys license and amendments as they apply to the Loss of Secondary Heat Sink. This is accomplished by presenting the operator with a set of plant conditions and requiring the operator to determine the procedurally preferred means for re-establishing the secondary heat sink The question is at the Comprehension/Analysis (2R1 or 3SPK) cognitive level because the operator must recall bits of information (i.e. equipment available, SG level), and then apply this information to a set of plant conditions to answer the question correctly.

Examination Outline Cross-reference: Level RO SRO Tier# 1 Group# 2 KIA# 003 AK3.07 Importance Rating 3.8 Knowledge of the reasons for the following responses as they apply to the Dropped Control Rod: Tech-Spec limits for T-ave Proposed Question: RO Question # 19 Unit 3 is at 3% power during a plant startup prior to entering MODE 1.

ONE Control Group C rod drops to the bottom of the core.

Tave indicates the following

Loop 3A 548° F

Loop 3B-546.5°F

Loop 3C-540.5°F Which ONE of the following identifies whether RCS temperature is above or below the minimum temperature for criticality, and the action and reason for action required?

A. BELOW minimum temperature for criticality. Restore Tave within 15 minutes or be in Hot Standby within the following 15 minutes because MTC may no longer be within the analyzed range for accident analysis.

B. BELOW minimum temperature for criticality. Immediately reduce power below the P-6 setpoint and stop all positive reactivity additions.

C. ABOVE minimum temperature for criticality. Trip the reactor and enter 3-EOP-E-0, Reactor Trip or Safety Injection, because Tave cannot be stabilized by adjusting steam demand.

D. ABOVE minimum temperature for criticality. Stabilize Tave in accordance with 3-ONOP-28.3, Dropped RCC, to ensure reactor trip system instrumentation remains operating within its analyzed range.

Proposed Answer: A

Explanation (Optional):

A. CORRECT B. Incorrect. Correct effect but there is no allowance for 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months prior to shutdown C. Incorrect. Any Tave below the limit is outside the LCO. Credible because normal mode of temperature stabilization is with turbine D. Incorrect. Any Tave below the limit is outside the LCO. If Tave was not exceeded, this could be a correct answer.

TS 3.1.1.4 Technical Reference(s): (Attach if not previously provided)

Proposed References to be provided to applicants during examination: N Learning Objective: (As available)

Question Source: Bank # WTSI 96605 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: 2007 McGuire Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 2 Facility operating limitations in the technical specifications and their bases.

Comments:

KA is matched because the applicant must determine whether a limit is exceeded and the reason (basis) for that limit Question is written at comprehension/analysis cognitive level because the applicant must choose an action based upon understanding of the technical specification LCO

Examination Outline Cross-reference: Level RO SRO Tier# 1 Group# 2 KIA# 005 AA1.01 Importance Rating 3.6 Ability to operate and / or monitor the following as they apply to the Inoperable / Stuck Control Rod: CRDS Proposed Question: RO Question # 20 Plant conditions:

Unit 3 Reactor power is 80%.

Bank D Group 2 Control Rod H8 is 20 steps lower than the rest of its bank.

The Shift Manager has directed the crew to realign Rod H8.

Annunciator B 9/4, ROD CONTROL URGENT FAILURE, alarms when the rod realignment begins.

Which ONE of the following identifies which lift coil disconnect switches were disconnected, AND the source of the B 9/4 alarm?

A. All Bank D switches except Rod H-8 were disconnected; Group 1 is the source of the B 9/4 alarm.

B. Only Bank D Group 2 switches except Rod H-8 were disconnected; Group 1 is the source of the B 9/4 alarm.

C. All Bank D switches except Rod H-8 were disconnected; Group 2 is the source of the B 9/4 alarm.

D. Only Bank D Group 2 switches except Rod H-8 were disconnected; Group 2 is the source of the B 9/4 alarm.

Proposed Answer: A Explanation (Optional):

A. Correct. 1st part correct, 2nd part correct. According to 3-ONOP-028.1 (p9; Rev 8/2/10) Step 5.9.4, the operator will be directed to place all the lift coil disconnect switches for the misaligned rod bank to the disconnect position (toggle switch down)

EXCEPT the misaligned RCC switch which is left in the connect position (toggle switch up). According to 3-ONOP-028.1 (plO; Rev 4/7/09), the Note prior to Step 5.9.8 that

Annunciator B9/4, ROD CONTROL URGENT FAILURE, and the RCC power cabinet URGENT FAILURE will alarm for the group with the lift coils disconnected.

B. Incorrect. 1st part wrong, 2nd part correct. This is incorrect because control rods in Bank D Groups 1 and 2 have been disconnected. According to Lesson Plan 6902105 (p21; Rev 9/18/07), there are five control rods in Bank D. Two are powered through Cabinet 1BD (D8, M8) and makeup the Bank D Group I rods. Three are powered through Cabinet 2BD (H4, H8, H12) and makeup Bank D Group 2 rods. Consequently, two rods in both Group I and 2 have been disconnected. This is plausible because the operator may incorrectly believe that only the control rods in the same group as the misaligned rod needs to be disconnected. The fact that the step counters for Group 1 and 2 are separate raises the plausibility.

C. Incorrect. 1st part correct, 2nd part wrong. This is incorrect because Group 1 is the source of the B 9/4 alarm. This is plausible because the operator may incorrectly believe that the cause of the alarm is associated with the one rod still connected.

D. Incorrect. jst part wrong, 2nd part wrong. See B and C.

3-ONOP-028.1 (p 9 (8/2/10) 10 (4/7/09))

Technical Reference(s): Lesson Plan 6902105 (p21; (Attach if not previously provided)

Rev 9/1 8/07)

Proposed References to be provided to applicants during examination: N

* . 6902207 Objectives 4 and 6 Learning Objective: (As available)

Question Source: Bank # WTSI 66696 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: 2009 Turkey Point (Not last 2)

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

55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

The KA is matched because the operator must demonstrate the ability to operate/monitor the Control Rod Drive System as it applies to the Inoperable/Stuck Control Rod event.

The question is at the Memory (1 P) cognitive level because the operator must recall bits of information, and then apply this information to a set of plant conditions to answer the question correctly.

NOTE: This question was used on the 2009 PTN NRC RO Exam. (not last 2)

Examination Outline Cross-reference: Level RO SRO Tier# I Group# 2 KIA# 024 AA2.01 Importance Rating 3.8 Ability to determine and interpret the following as they apply to the Emergency Boration:

Whether boron flow and/or MOVs are malfunctioning, from plant conditions Proposed Question: RO Question # 21 The Unit 3 RO is Emergency Borating in accordance with 3-ONOP-046.1, Emergency Boration.

The 3A Charging Pump is running.

Emergency Boration valve, MOV-3-350, is OPEN.

Emergency Borate Flow indicator, Fl-3-110, displays 45 gpm flow.

Charging Flow indicator, Fl-3-122A indicates 30 gpm.

Which ONE of the following identifies the action needed, if any, to establish emergency boration flow per 3-ONOP-046.1?

A. NO additional action is needed. Flow indication is within required limits.

B. Must raise boration flow as indicated on FI-3-1 10 by >15 gpm ONLY.

C. Must raise charging flow as indicated on FI-3-122A by >15 gpm ONLY.

D. BOTH Charging and Emergency Boration flows must be raised by >15 gpm.

Proposed Answer: D Explanation (Optional):

A. Incorrect. This is incorrect because the flow of 45 gpm on FI-3-1 10 and 30 gpm on Fl 3-122A is inadequate. This is plausible because according to Technical Specification 3.1.1.1 (p3!4 1-1; Amendment 247 and 243) the Shutdown Margin shall be within the limits specified in the COLR while in Modes 1-4. If not, the operator must immediately initiate and continue boration at greater than or equal to 16 gpm of a solution containing greater than or equal to 3.0 wt% (5245 ppm) boron or equivalent until the required SHUTDOWN MARGIN is restored. The operator may incorrectly believe that 16 gpm is the minimum requirement for emergency boration, and conclude that no additional action is needed.

B. Incorrect. This is incorrect because the flow of 45 gpm on FI-3-1 10 and 45 gpm on Fl 3-122A is inadequate. This is plausible because the operator may be unaware of the required flowrates, and incorrectly believe that this adjustment is acceptable.

C. Incorrect. This is incorrect because the flow of 60 gpm on FI-3-110 and 30 gpm on Fl 3-122A is inadequate. This is plausible because the operator may be unaware of the required flowrates, and incorrectly believe that this adjustment is acceptable.

D. Correct. According to 3-ONOP-046.1 (p5; Rev 2) Step 1 .g, after aligning for Emergency Boration, the operator is directed to check that emergency boration flow is established by verifying FI-3-1 10 > 60 gpm and Fl-3-1 22A> 45 gpm. If not, which is the case under the stated conditions, the operator is directed to start additional charging pumps and align valves as necessary to establish emergency boration flow.

Consequently, flow on both Fl-3-1 10 and Fl-3-1 22A must be raised 15 gpm.

3-ONOP-046.1 (p5; Rev 2)

Technical Specification 3.1.1.1 Technical Reference(s): 4 1-1; Amendment 247 and (Attach if not previously provided) 3 (p

243)

Proposed References to be provided to applicants during examination: None 6902232 Objectives 4 and 6 Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam: NA Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

The KA is matched because the operator must demonstrate the ability to determine and

interpret whether boron flow and/or MOVs are malfunctioning, from plant conditions as they apply to the Emergency Boration. This is accomplished by providing a set of actions/conditions that are observed when Emergency Boration is being established, and then requiring the operator to evaluate the flowrates, and take action if needed. It is the same mental exercise that is required of the operator when performing step 1 .g of 3-ONOP-041 .6.

The question is at the Comprehension/Analysis (2DR) cognitive level because the operator must recall bits of information (Required flowrates on FI 110/1 22A), and then relate this to itself by determining whether or not these flowrates exist, and identifying the action required to establish these flowrates, to answer the question correctly.

Examination Outline Cross-reference: Level RO SRO Tier# 1 Group# 2 K!A# 032 AA2.05 Importance Rating 2.9 Ability to determine and interpret the following as they apply to the Loss of Source Range Nuclear Instrumentation: Nature of abnormality, from rapid survey of control room data Proposed Question: RO Question # 22 Plant conditions:

A reactor startup is in progress.

The crew is verifying proper overlap and preparing to block Source Range High Flux Trips.

IR channel N-35 indicates 3 X 10-11 amps and slowly rising

IR channel N-36 indicates 8 X 10-11 amps and slowly rising

The reactor then trips.

Which ONE of the following conditions caused the reactor trip?

A. SR Channel N-31 Pulse Height Discrimination was lost, causing the SR High Flux Trip Bistable to trip.

B. SR Channel N-31 High Voltage power supply was lost, causing the SR High Flux Trip Bistable to trip.

C. IR Channel N-36 being undercompensated caused the trip prior to P-6 being satisfied.

D. lR Channel N-35 being overcompensated caused the trip prior to P-6 being satisfied.

Proposed Answer: A Explanation (Optional):

A. Correct. if compensation for pulse height is lost, then the source range detector would read gamma activity as well as neutron activity, resulting in an upwards spike in indicated power B. Incorrect. HV power does not trip the SR bistable. Loss of HV power will result in an

artificially low reading on the source range detector C. Incorrect. Both of the Intermediate Range channels indicate correctly, although an undercompensated channel would result in higher indication and would cause a trip if conditions were not stable and the unit power was at P-6 or above D. Incorrect. Both of the Intermediate Range channels indicate approximately correctly, and an overcompensated channel would result in lower indication, not a higher indication

Lesson Plan 6900104 (p47 Technical Reference(s): (Attach if not previously provided)

Rev 12/8/07)

Proposed References to be provided to applicants during examination: No

69001040bj14 Learning Objective: (As available)

Question Source: Bank # 94630 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: 2002 Beaver Valley Unit 1 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Comments:

KA is matched because the item simulates an operators rapid scan of the control room to determine why the reactor will trip under a specific set of conditions.

Item developed to comprehension/analysis level because the applicant must put several pieces of information together to eliminate distracters and to determine the correct answer based on combination of operation of plant equipment and plant conditions.

Examination Outline Cross-reference: Level RO SRO Tier# 1 Group# 2 KIA# 033 AK3.0i Importance Rating 3.2 Knowledge of the reasons for the following responses as they apply to the Loss of Intermediate Range Nuclear Instrumentation: Termination of startup following loss of intermediate- range instrumentation Proposed Question: RO Question # 23 Plant conditions:

Unit 3 plant startup is in progress in accordance with 3-GOP-301, Hot Standby to Power Operation.

Both Intermediate Range Channels are reading 11 8x10 amps.

Subsequently:

Intermediate Range Channel N36 fails LOW.

The crew enters 3-ONOP-059.7, Intermediate Range Nuclear Instrumentation Malfunction.

Which ONE of the following identifies the required action per 3-ONOP-059.7, AND the reason for this action?

A. Maintain power below the P-6 setpoint on N35; Both Intermediate Range channels are required to be OPERABLE under the current plant conditions.

B. Power may be raised to any value below the P-i 0 setpoint; Only ONE Intermediate Range Channel is required to be OPERABLE below P-b.

C. Power may be raised to any value below the P-b setpoint; Two Intermediate range channels are required to be OPERABLE to meet accident analysis assumptions at power levels above P-b.

D. Maintain power below the P-6 setpoint on N35; Only ONE Intermediate Range Channel is required to be OPERABLE below P-6.

Proposed Answer: A

Explanation (Optional):

A. Correct. According to 3-ONOP-059.7 (p8; Rev 10/3/05) Step 8, the operator is directed to check power level < P-6. According to SD-063 (p41; Rev 9/10/11) the P-6 setpoint is 10-10 amps. Since power level is < P-6, the operator is directed to Maintain power below the P-6 setpoint until both IR channels are Operable. According to BD-ONOP 059.7 (p8; Rev 10/3/05) the reason for this action is Tech Spec 3.3.1 Action 3a, which according to Technical Specification 3.3.1 (p3/4 3-6; Amendment 179 and 173), states Below the P-6 (Intermediate Range Neutron Flux Interlock) Setpoint, restore the inoperable channel to OPERABLE status prior to increasing THERMAL POWER above the P-6 Setpoint.

B. Incorrect. This is incorrect because the limit on power is the P-6 setpoint, not P-b.

This is plausible because the power limit established by the procedure is P-b if power level is above P-6. The operator may incorrectly believe that the power level is above P-6.

C. Incorrect. This is incorrect because the limit on power is the P-6 setpoint, not P-b 0, and because Technical Specifications do not allow increasing power to 10%. This is plausible because the power limit established by the procedure is 5% if power level is above P-6, and because according to BD-ONOP-059.7 (p7; Rev 10/3/05) the reason for limiting power to 5% until both IR Channels are restored to Operable when power level is above P-6 is because it is prudent to do so.

D. Incorrect. This is incorrect because Technical Specifications does not allow an inoperable IR channel at this power level. First part is correct.

3-ONOP-059.7 (p8; Rev 10/3/05)

BD-ONOP-059.7 (p8; Rev 10/3/05)

Technical Reference(s): Technical Specification 3.3.1 (Attach if not previously provided)

(p3/4 3-6; Amendment 179 and 173)

SD-063 (p41; Rev 9/10/11)

Proposed References to be provided to applicants during examination: None 6902163 Objective 7 Learning Objective: 6902206 Objective 4 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam: NA Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

The KA is matched because the operator must demonstrate knowledge of the reason for termination of startup following loss of intermediate-range instrumentation as it applies to the Loss of Intermediate Range Nuclear Instrumentation.

The question is at the Comprehension/Analysis (2DR) cognitive level because the operator must recall bits of information (power restrictions on plant), and then relate this information to itself by describing the reason why to answer the question correctly.

Examination Outline Cross-reference: Level RO SRO Tier# 1 Group# 2 K!A# 059 AK1.05 Importance Rating 2.6 Knowledge of the operational implications of the following concepts as they apply to Accidental Liquid Radwaste Release: The calculation of offsite doses due to a release from the power plant Proposed Question: RO Question # 24 Plant conditions:

Unit 3 is in Mode 5.

Unit 4 is at 100% power.

3A1 Circulating Water Pump is running.

There is an on-going unplanned liquid radioactive release in progress.

R-18, Liquid Waste Effluent Radiation Monitor, alarmed but RCV-018 remains OPEN.

The crew has entered 3-ONOP-067, Radioactive Effluent Release.

Which ONE of the following identifies an action required in accordance with 3-ONOP-067?

A. Direct Security to block access to the Circulating Water discharge from normal traffic.

B. Start Circulating Water Pump 3A2 to raise dilution flow.

C. Direct Chemistry to sample and analyze the effluent release to determine the extent of contamination and off-site dose rates.

D. Stop the release by stopping Circulating Water Pump 3A1.

Proposed Answer: C Explanation (Optional):

A. Incorrect. Plausible because this is a logical assumption due to the fact that there is an effluent release in progress B. Incorrect. Plausible because dilution flow is important to the calculated discharge rate for off-site discharge activity. Not correct because the actual activity is above the value set on the discharge permit and additional dilution flow will not change that.

C. Correct. 3-ONOP-067 directs the crew to direct Chemistry to sample for off-site release rates when an inadvertent discharge is occurring.

D. Incorrect. Stopping CWP3A1 will also close the CWP discharge valve, stopping Circulating Water flow to the outfall. The applicant may reasonably assume that stopping Circulating Water flow will stop all flow, but the effluent discharge ties in downstream of CW discharge valves, so the release would not be stopped by this action 4-ONOP-067 (p14; Rev Technical Reference(s): 11/10/10) (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None 6902242 Objectives 5 and 6 Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 12 55.43 Radiological safety principles and procedures.

Comments:

KA is matched because it evaluates the ROs knowledge of the requirement for off-site dose rate determination during an unplanned radioactive liquid release Question is developed to the memory level because the applicant must choose the action that applies based on the procedural requirement

Examination Outline Cross-reference: Level RO SRO Tier# 1 Group# 2 K/A# 061 AK2.01 Importance Rating 2.5 Knowledge of the interrelations between the Area Radiation Monitoring (ARM) System Alarms and the following: Detectors at each ARM system location Proposed Question: RO Question # 25 Fuel shuffle is being performed in the Unit 3 Spent Fuel Pool.

Spent Fuel Pool Area monitor, RI-3-1407B, alarm is received in the Control Room.

Which ONE of the following identifies the indication available to the operators in the spent fuel pooi that the monitor is in alarm?

A. Red LED on the local monitor and a horn B. Red flashing light on the local monitor and a horn C. Horn ONLY D. Red flashing light ONLY Proposed Answer: B Explanation (Optional):

A. Incorrect. Second part is correct but the red LED isnt part of the local alarm, it is for the remote alarm.

B. Correct.

C. Incorrect. Plausible because it is half of the local alarm function of the ARM D. Incorrect. Plausible because it is half of the local alarm function of the ARM Lesson Plan 6902168 (p39, 163 Technical Reference(s): and 165; Rev 2/26/1 0) (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None 6902168 Objectives 2 4and5 Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam: NA Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 11 55.43 Purpose and operation of radiation monitoring systems, including alarms and survey equipment.

Comments:

The KA is matched because the operator must demonstrate knowledge of the interrelations between the Area Radiation Monitoring (ARM) System Alarms and the detectors at each ARM system location, specifically, the local alarm function of the spent fuel pool monitor The question is at the Memory (IF) cognitive level because the operator must recall bits of information to answer the question correctly.

Examination Outline Cross-reference: Level RO SRO Tier# 1 Group# 2 K!A# E15 2.1.27 Importance Rating 3.9 Conduct of Operations: Knowledge of system purpose and / or function.

Proposed Question: RO Question # 26 Which ONE of the following identifies the systems checked for leakage into Containment by FR-Z.2, Response to Containment Flooding, AND the basis for limiting leakage from these systems into containment?

CCW and...

A. CVCS; To protect systems needed for recovery.

B. CVCS; To protect the Containment barrier.

C. Primary Makeup Water; To protect the Containment barrier.

D. Primary Makeup Water; To protect systems needed for recovery.

Proposed Answer: D Explanation (Optional):

A. Incorrect. Plausible because CVCS could spill into containment but the flow rates would not be a high concern for containment flooding. Second part is correct B. Incorrect. Plausible for same reason as A but the Containment Flooding FR is not concerned with the containment barrier as much as protecting vital equipment needed for recovery. Containment barrier protection is FR-Z.1, not FR-Z.2 C. Incorrect. Plausible because PMW is correct, and because the containment barrier protection is in the same series of FR procedures.

D. Correct. PMW is identified as a potential unintended source of water in containment

that can result in flooding 4-EOP-FR-Z.2 (p5; Rev 4/15/99)

Technical Reference(s): BD-EOP-FR-Z.2 (p7; Rev (Attach if not previously provided) 4/15/99)

Proposed References to be provided to applicants during examination: None 6910338 Objectives 12 l3andl4 Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam: NA Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

The KA is matched because the operator must demonstrate knowledge of system purpose and for function associated with the Emergency Plant Event of Containment Flooding. This is accomplished by identifying two of four systems that the flooding procedure directs the operator to check as a potential source of leakage into the containment. By knowing the purpose/functions of each system under consideration, the operator can arrive at the correct answer.

The question is at the Memory (1 F or 1 P) cognitive level because the operator must recall bits of information (Purposes/Functions of systems, reason for avoiding leakage from these systems) to answer the question correctly.

Examination Outline Cross-reference: Level RO SRO Tier# 1 Group# 2 KIA# E03 EKI.1 Importance Rating 3.4 Knowledge of the operational implications of the following concepts as they apply to the (LOCA Cooldown and Depressurization) Components, capacity, and function of emergency systems.

Proposed Question: RO Question # 27 Plant conditions:

Unit 3 has experienced a LOCA and a Loss of Offsite Power.

The crew is implementing 3-EOP-ES-1 .2, Post-LOCA Cooldown and Depressurization.

Which ONE of the following identifies the steady state loading limit for each Unit 3 Emergency Diesel Generator, AND the strategy used if insufficient diesel capacity exists to start a Charging Pump?

A. 2500 KW; Do not start a Charging Pump within 3-EOP-ES-1 .2.

B. 2500 KW; Shed non-essential load from the diesel in order to start a Charging Pump.

C. 2750 KW; Do not start a Charging Pump within 3-EOP-ES-1 .2.

D. 2750 KW; Shed non-essential load from the diesel in order to start a Charging Pump.

Proposed Answer: B Explanation (Optional):

A. Incorrect. 1 st part correct, 2nd part wrong. This is incorrect because the strategy employed by ES-i .2 under the stated conditions is to shed non-essential loads on the Emergency Diesel Generator in order to later start a Charging Pump. This is plausible because according to BD-EOP-ES-i .2, the Charging pump operation is not essential for recovery; and the operator may incorrectly believe that if there is insufficient capacity on the diesels, the Charging Pump is not started during the recovery.

B. Correct. 1 st part correct, 2nd part correct. According to 3-EOP-ES-1 .2, Attachment 2, a Caution is provided stating the Steady state loading on each Unit 3 emergency diesel generator shall NOT exceed 2500 KW. According to 3-EOP-ES-1 .2 Step 2, the operator is directed to check if all 4KV Busses are energized by Offsite power. Since a LOOP has occurred, the RNO will need to be implemented. The operator will be directed to check diesel capacity adequate to run charging pumps, and IF adequate diesel capacity is NOT available, THEN shed non-essential loads. The operator will then be directed to continue in the procedure. At step 4, the operator will be directed to check that at least one Charging Pump is running. Since a LOCA has occurred, SIS will have actuated and all Charging Pumps will be OFF. Again, the RNO will need to be implemented. The Step 4 RNO directs the operator to start at least one Charging Pump. According to BD-EOP-ES-i .2 maximum charging flow to the RCS is established in order to provide sufficient makeup so that high-head SI pumps can later be stopped.

If charging flow cannot be established, only a partial SI reduction will be possible since at least one high-head SI pump will be needed for RCS makeup. Charging pump operation is not essential for recovery, and actions to restore charging flow should not delay subsequent steps to cooldown and depressurize the RCS and reduce SI flow which are necessary to limit reactor coolant leakage. However, a previous step will ensure adequate diesel capacity to run the charging pumps if offsite power is not available.

C. Incorrect. s 1 t part wrong, 2nd part wrong. See A and D.

D. Incorrect. 1 5t part wrong, 2nd part correct. This is incorrect because the steady state loading limit for each Unit 3 Emergency Diesel Generator is 2500 KW. This is plausible because according to 3-EOP-ES-i .2, Attachment 2, a Caution is provided stating when starting additional equipment, diesel load is required to be monitored to ensure the transient limit of 2750 KW is NOT exceeded. The operator may incorrectly believe that the Transient Limit is the Steady-State Limit.

3-EO P-ES-i .2 (p7, 9 and 33; Rev 6/i 5/12)

Technical Reference(s): BD-EOP-ES-i .2 (p15; Rev (Attach if not previously provided) 6/15/12)

Proposed References to be provided to applicants during examination: None 6910329 Objectives 3, 4, 5, and 9 Learning Objective: 6902136 Objective 4 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam: NA Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

The KA is matched because the operator must demonstrate knowledge of the operational implications of the components, capacity, and function of emergency systems as they apply to a LOCA Cooldown and Depressurization. This is accomplished by requiring the operator to identify the rating (i.e. capacity) of the EDGs and the strategy employed (operational implications) if there is limited electrical capacity with respect to the Charging Pump.

The question is at the Memory (1 P) cognitive level because the operator must recall bits of information (i.e. EDG rating, Strategy for Charging Pump with limited electrical capacity) to answer the question correctly.

Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 1 K!A# 003 K5.02 Importance Rating 2.8 Knowledge of the operational implications of the following concepts as they apply to the RCPS: Effects of RCP coastdown on RCS parameters Proposed Question: RO Question # 28 Which ONE of the following describes a function of the flywheel on the RCPs?

A. Prolongs RCP coastdown time to aid in maintaining loop flow thus maintaining DNBR within acceptable limits during certain loss of flow events.

B. Minimizes acceleration on pump start to minimize the effects of core lift when the first RCP is started during an RCS heatup from Cold Shutdown.

C. Prolongs RCP coastdown time to aid in maintaining loop flow thus maintaining hot channel factors at an acceptable level during certain loss of RCS flow events.

D. Maintains constant RCP speed, minimizing the potential for spurious RCS low flow reactor trips and maintaining hot channel factors at an acceptable level during power operation.

Proposed Answer: A Explanation (Optional):

A. Correct. According to Lesson Plan 6902108 the Flywheel prevents DNB during trip and blackout conditions by supplying inertia to the RCP rotor which will extend pump coast down time. According to Lesson Plan 6902108, a flywheel attached to the pump motor provides coastdown inertia supplying sufficient flow to prevent DNB in the event of a plant loss of offsite power.

B. Incorrect. This is incorrect because the flywheel does not minimize the acceleration on pump start. This is plausible because the operator may incorrectly believe that the large size of the Flywheel is used to limit the acceleration of the RCP shaft, and thereby limit the differential pressure across the core during plant startup. According to 3-GOP-503 Step 5.5.2, the RCPs are started early in the transition from Cold Shutdown to Hot Standby, and therefore, at low temperature. The lower the temperature, the more

dense the RCS water will be, and the more dense water will lead to a larger tP across the Rx Core. According to Lesson Plan 6902108, Table 8, the Flywheel is 72 inches in diameter, and weighs 12,150 pounds.

C. Incorrect. This is incorrect because the flywheel is not used to maintain the hot channel factors. Evidence for this is that according to Technical Specification LCO3.2.2 and 3.2.3, these power distribution limits are only applicable in Mode 1. When the pump is coasting down, the plant will be in Mode 3, and the Hot Channel Factors are not applicable. According to 3-ADM-536 the limits on heat flux hot channel factor and nuclear enthalpy rise hot channel factor ensure that the design limits on peak local power density and minimum DNBR are not exceeded. In other words, when the operator maintains these measureable parameters, it is done, in part, for the same reason that the designer has installed a flywheel on the RCP. This is plausible because the operator may incorrectly believe that the flywheel is used on the RCP to maintain these parameters, rather than DNBR.

D. Incorrect. This is incorrect because the Flywheel is not used to maintain a constant speed of the RCP, nor is it used to maintain hot channel factors. This is plausible because the operator may incorrectly believe that the flywheel is used to maintain a constant speed during operation, rather than prolong the RCP shaft speed post-trip. If so, the operator may also incorrectly believe that this is done to ensure hot channel factor limits are maintained.

Lesson Plan 6902108 (p21, 22 and 59; Rev 3/30/201 0) 3-ADM-536 (p57; Rev 3)

Technical Specification Technical Reference(s):

(Attach if not previously provided)

LCO3.2.2 and 3.2.3 (p3/4 2-4 and 2-11; Amendments 191 and 185)

Proposed References to be provided to applicants during examination: No 6902108 Objective 4.b.9 Learning Objective: (As available)

Question Source: Bank# WTSI 55194 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: 2005 Beaver Valley Unit 2

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 3 55.43 Mechanical components and design features of reactor primary system.

Comments:

The KA is matched because the operator must demonstrate an understanding of why the Flywheel is included in the design of the RCP. By doing so, the operator demonstrates knowledge of the operational implications of the effects of RCP coastdown on RCS parameters, such as loop flow, as these effects apply to the RCPs.

The question is at the Comprehension/Analysis (2DS) cognitive level because the operator must demonstrate an understanding of the material (i.e. why the Flywheel is there) from three potential choices, by relating it to other known material (i.e. large flywheel, RCPs started at low temperature, Hot Channel Factors are operator maintained parameter) in order to answer the question correctly.

The Question was used at Beaver Valley 2 (2005)

Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 1 KIA# 004 K5.19 Importance Rating 3.5 Knowledge of the operational implications of the following concepts as they apply to the CVCS:

Concept of SDM Proposed Question: RO Question # 29 Unit 3 is operating at 100% power.

Which ONE of the following evolutions, by itself, will RAISE Shutdown Margin?

A. Placing a new Mixed Bed Demineralizer in service prior to rinsing.

B. Adjusting CCW flow through the NRHX such that Letdown temperature is lowered by 10° F.

C. Raising the setting of the boric acid totalizer from 20 to 50 gallons during an automatic blended makeup to the VCT.

D. Lowering the flow setting of the primary water flow controller (HIC-3-1 14) during an automatic blended makeup to the VCT from 70 to 60 gpm.

Proposed Answer: D Explanation (Optional):

A. Incorrect. See D. According to 3-OP-047.3, placing a CVCS Demineralizer in service to the VCT, prior to its effluent boron concentration being within 10 percent of the RCS boron concentration (not to exceed 50 ppm), may cause an inadvertent reactivity change. This Caution precedes a Section entitled Placing A Demineralizer in service without Rinse. According to Lesson Plan 6902147 until saturated the Anion resin will initially reduce the RCS boron concentration causing an increase in power level if not monitored closely. Consequently, the placement of an un-rinsed (non-saturated) Mixed Bed Demineralizer in service will result in lower RCS Boron concentration, which will LOWER, not raise Shutdown Margin.

B. Incorrect. See D. According to SD-01 3,the temperature of the letdown at the non regenerative heat exchanger outlet is controlled by varying CCW flow through the shell side. According to SD-013, a significant decrease in letdown temperature can result in a reduction in boron concentration at the demineralizer outlet. This is due to increased boron ion exchange by the anion resin at lower temperatures. Unless letdown flow is

diverted to the HUTs, a positive reactivity addition may result. Consequently, the cooler Letdown Temperature will result in lower RCS Boron concentration, which will LOWER, not raise Shutdown Margin.

C. Incorrect. See D. According to SD-013, in the automatic mode, a low level signal (17%) from the VCT level controller initiates makeup of a preselected blend of boric acid and water to the suction of the charging pumps as follows: (1) Starts the boric acid transfer pump, (2) Opens the primary water flow control valve FCV-3-114A and the boric acid flow control valve FCV-3-1 13A, and (3) Opens the makeup stop valve FCV I 13B to the charging pump suction header. Once opened FCV-3-1 I 3A will modulate to hold the boric acid flow rate setpoint which is dialed into its BLEND controller FC-3-1 13.

FCV-3-114A will modulate to hold the primary water flowrate setpoint which is dialed into HIC-3-1 14 Primary Water Auto Setpoint. This controller is normally set to maintain a constant flowrate of approximately 70 GPM. The boric acid flow rate is adjusted so that the concentration of the blend matches that in the RCS. The primary water and boric acid flows meet and are mixed in the boric acid blender. Makeup addition to the charging pump suction header causes the water level in the volume control tank to rise. At a preset level (31%), the boric acid transfer pump stops, the primary makeup water and boric acid flow control valves (FCV-3-1 I 3A & 11 4A) close, and the makeup stop valve (FCV-3-1 1 3B) closes. The action of raising the total gallons on the Totalizer does not affect the blend during an Auto Makeup. It would affect the total amount of Boric Acid added to the RCS if a manual boration, rather than an auto makeup, was occurring. Shutdown Margin will remain unchanged by this action.

D. Correct. According to Technical Specification Definition 1.25, SHUTDOWN MARGIN shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming all full-length rod cluster assemblies (shutdown and control) are fully inserted except for the single rod cluster assembly of highest reactivity worth which is assumed to be fully withdrawn.

Consequently SDM is a direct function of RCS Boron Concentration. Raising RCS Boron concentration will raise Shutdown Margin, and lowering RCS Boron concentration will lower Shutdown Margin. According to SD-013, in the automatic mode, a low level signal (17%) from the VCT level controller initiates makeup of a preselected blend of boric acid and water to the suction of the charging pumps. The PW controller is normally set to maintain a constant flowrate of approximately 70 GPM. The boric acid flow rate is adjusted so that the concentration of the blend matches that in the RCS. If the PW controller setting is lowered, and the BA flow setting remains unchanged, the blend entering the RCS during an auto makeup will be at a higher boron concentration than the RCS, and raise RCS boron concentration as well as Shutdown Margin.

Technical Specification Definition 1.25 (p1-5; Amendment 249)

SD-013 (p19-20, 22,41; Rev Technical Reference(s): 9/20/11) (Attach if not previously provided) 3-OP-047.3 (p14.; Rev 3)

Lesson Plan 6902147 (p54; Rev 9/18/07)

Proposed References to be provided to applicants during examination: N 6902147 Objective 9 6902113 Objectives 6.c, 8.g 9.g Learning Objective: (As available) and 13.d Question Source: Bank #

Modified Bank # WTSI 68614 (Note changes or attach parent)

New Question History: Last NRC Exam: 2008 Harris Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 1 55.43 Fundamentals of reactor theory, including fission process, neutron multiplication, source effects, control rod effects, criticality indications, reactivity coefficients, and poison effects.

Comments:

The KA is matched because the operator must demonstrate the knowledge of the operational implication the concept of Shutdown Margin as it applies to operator action taken on the CVCS.

The question is at the Comprehension/Analysis (3PEO) cognitive level because the operator must recall bits of information (i.e. what is SDM, how is it changed when RCS boron concentration is changed), and then apply this information to a set of plant actions predicting which action will result in a raised SDM to answer the question correctly.

The Question was used at Shearon Harris (2008).

Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 1 KIA# 005 K6.03 importance Rating 2.5 Knowledge of the effect of a loss or malfunction on the following will have on the RHRS: RHR heat exchanger Proposed Question: RO Question # 30 Plant conditions:

A Unit 4 cooldown is in progress in accordance 4-GOP-305, Hot Standby to Cold Shutdown.

RHR Pump 4B tripped.

The operating crew implemented 4-ONOP-050, Loss of RHR.

HCV-4-758, RHR Heat Exchanger Outlet Flow, was closed from the Control Room and failed to reopen.

RCS temperature is 190°F and rising.

RCS pressure is 150 psig and rising.

In accordance with 4-ONOP-050, which ONE of the following is the required method to re-establish cooling to the RCS?

A. Place RHR Pump 4A in service.

B. Establish Steam Generator Blowdown.

C. Establish a Secondary Heat Sink and dump steam to the Condenser.

D. Establish a Secondary Heat Sink and dump steam to the atmosphere.

Proposed Answer: D Explanation (Optional):

A. Incorrect: This is incorrect because starting RHR Pump 4A will NOT result in restored cooling since HCV-758 isolates this pump as well. Consequently, circulation of RCS water via RHR Pump 4A will bypass the RHR heat exchanger and NOT result in a change in the upward trend of RCS temperature. This is plausible because the operator may incorrectly believe that HCV-758 does NOT affect both pumps, and according to 4-ONOP-50 Step 6, starting the opposite train of RHR is a strategy used within the ONOP

to restore RCS cooling.

B. Incorrect: This is not correct because the next mitigating strategy in 4-ONOP-050 will be to establish a Secondary Heat Sink and dump steam to the atmosphere. There is no reason to believe that this will NOT be successful. It is plausible because according to 4-ONOP-50 Step 14, the operator will be directed to determine if blowdown should be established. However, it will only be established if dumping steam has not been successful. The operator may incorrectly believe that establishing blowdown is accomplished before dumping steam from a SG.

C. Incorrect: This is not correct because the steam dump will not be directed to the Condenser. This is plausible because steam dump to the condenser is an available option per design, and may appear attractive because it offers the advantage of demineralized water conservation. Because of this, the operator may incorrectly believe that steam will be directed to the Condenser.

D. Correct: Since restarting the A Train of RHR will not reverse the RCS temperature trend, the crew will evaluate the RCS temperature condition at Step 9 of 4-ONOP-050.

According to 4-ONOP-050 Steps 9-13, the operator will determine that the RCS temperature is still rising, Isolate Containment, and then establish a Secondary Heat sink by initiating feedwater to the Steam Generators and then opening the Atmospheric Steam Dumps as necessary to maintain temperature.

4-ONOP-050 (p9, 11-13; Rev Technical Reference(s): 12/3/07) (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None

* . LP6902210 Obj.4 Learning Objective: (As available)

Question Source: Bank # PTN Bank Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: NA Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X

10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

The KA is matched because the operator must demonstrate knowledge of the effect of a loss or malfunction on the RHR heat exchanger will have on the RHRS; specifically, the procedural actions that must be taken when the RHR HXs are isolated during RHRS operation.

The question is at the Comprehension/Analysis (3SPK) cognitive level because the operator must recall bits of information such as the major action strategies within an ONOP, and then sort through a set of existing plant conditions, applying the recalled information to answer the question correctly.

ROll on ILC-27 Audit.

R030 on ILC-27 NRC Exam is a suitable alternative.

Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 1 KIA# 006 A1.18 Importance Rating 4.0 Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the ECCS controls including: PZR level and pressure Proposed Question: RO Question # 31 Plant conditions:

A LOCA has occurred on Unit 3.

The crew is performing actions of 3-EOP-ES-1 .2, Post LOCA Cooldown and Depressurization.

Pressurizer level is 78% and stable.

RCS Pressure is stable at 1400 psig.

ONE Charging pump is running.

3B RCP is running.

The US determines that the first High Head SI pump can be stopped.

When the High Head SI Pump is stopped, which ONE of the following describes the Pressurizer pressure response within the next 60 seconds? (Assume no additional operator actions)

A. PZR pressure will remain at its current value until all High Head SI flow is stopped, then lower until saturation conditions are reached.

B. PZR pressure will remain at its current value, then slowly lower as the bubble under the reactor vessel head stops expanding.

C. PZR pressure will immediately drop until RCS saturation conditions are reached, and then it will stabilize.

D. PZR pressure will immediately drop until High Head SI flow equalizes with break flow, then it will stabilize.

Proposed Answer: D Explanation (Optional):

A. Incorrect. Pressure would not remain stable if I HHSI pump is stopped while injecting.

Inventory would decrease because with a break prior to stopping the pump, level and pressure is stable, so it has to lower once less injection water is flowing B. Incorrect. Pressure is stable in the conditions given. Flow in = flow out. Stopping 1 HHSI pump will cause less flow in, meaning pzr pressure will have to go down.

Additionally, there will be no bubble under the head with one RCP running.

C. Incorrect. In 3-EOP-ES-1 .2, RCS subcooling is checked prior to stopping one HHSI pump, so unless the break became bigger simultaneously with stopping the pump, conditions will not reach saturation.

D. Correct.

3-EOP-ES-1.2 (p17; Rev 6/15/12)

BD-3-EOP-ES-1.2 (p40; Rev Technical Reference(s): 6/1 5/1 2) (Attach if not previously provided)

Lesson Plan 6910329 (p8; Rev 3/21/12)

Proposed References to be provided to applicants during examination: N Learning Objective: (As available)

Question Source: Bank # WTSI 87011 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: 2007 Beaver Valley Unit 1 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 5 55.43 Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.

Comments:

KA is matched because the item evaluates the applicants understanding of presurizer pressure response when SI flow is reduced after a LOCA Question is written at comprehension level because the applicant must determine that plant conditions are stable and will change when SI flow is reduced

Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 1 K/A# 007 K1.01 Importance Rating 2.9 Knowledge of the physical connections and/or cause-effect relationships between the PRTS and the following systems: Containment system Proposed Question: RO Question # 32 Which ONE of the following correctly identifies the mechanism for Pressurizer Relief Tank (PRT) overpressure control and its predicted affect on containment atmosphere?

A. Relief valve relieves at 100 psig and containment pressure will remain the same.

B. Rupture discs burst at 100 psig and containment pressure will remain the same.

C. Rupture discs burst at 100 psig and containment pressure will increase.

D. Relief valve relieves at 100 psig and containment pressure will increase.

Proposed Answer: C Explanation (Optional):

A. Incorrect. The PRT does not have a relief valve. Plausible because every other enclosed tank inside containment has a relief valve.

B. Incorrect. Plausible because the PRT has rupture discs, but the rupture discs burst directly to containment atmosphere and will raise containment pressure C. Correct.

D. Incorrect. The PRT does not have a relief valve, but will raise containment pressure SD-009 (p38 and 66; Rev Technical Reference(s): (Attach if not previously provided) 3-EOP-FR-H.1 iJ14 Rev 5)

Proposed References to be provided to applicants during examination: None 6902109 Objectives 5.d, 6.g, 8, Learning Objective: 10.h and 11.a (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam: NA Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 3 55.43 Mechanical components and design features of reactor primary system.

Comments:

The KA is matched because the operator must demonstrate Knowledge of the physical connections and/or cause-effect relationships between the PRTS and the Containment. This is accomplished by requiring the operator to evaluate the operation of the PRT Rupture Disks and then predict the consequence of the Rupture Disk operation on the Containment Atmosphere.

The question is at the memory cognitive level because the operator must recall bits of information (Setpoint of Rupture Disk), and then relate this information to itself by recognizing the impact on the Containment Atmosphere of the Rupture Disks releasing to the Containment in order to answer the question correctly.

Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 1 K/A# 007 K4.01 Importance Rating 2.6 Knowledge of PRTS design feature(s) and/or interlock(s) which provide for the following:

Quench tank cooling Proposed Question: RO Question # 33 Plant conditions:

Unit 4 is operating at 100% power.

The crew is attempting to reduce PRT liquid temperature in accordance with 4-NOP-041.03, Pressurizer Relief Tank.

The crew has opened the following two valves in order to raise PRT level:

CV-4-519A, PRIMARY WATER CONTAINMENT ISOL VLV

CV-4-519B, PRT PRIMARY MAKE UP Subsequently, a Safety Injection signal actuates.

In accordance with 4-NOP-041 .03, which ONE of the following identifies the required action, if any, regarding the Primary Water Valves that have been opened?

A. No action is needed, both valves have automatically closed.

B. The operator must manually close ONLY CV-4-519B, PRT PRIMARY MAKE UP.

C. The operator must manually close ONLY CV-4-519A, PRIMARY WATER CONTAINMENT ISOL VLV.

D. The operator must manually close CV-4-519B, PRT PRIMARY MAKE UP and CV 519A, PRIMARY WATER CONTAINMENT ISOL VLV.

Proposed Answer: B Explanation (Optional):

A. Incorrect. This is incorrect because CV-4-519B must be manually closed. It is plausible because the operator may incorrectly believe that both valves have an automatic function.

B. Correct. According to SD-009 Primary Makeup Water can be sprayed into the PRT via air operated control valves CV-3!4-519A and CV-3/4-519B. The spray header, located in the top of the PRT, disperses the makeup water over the entire water surface to enhance condensation and cooling, following a relief or safety valve discharge. CV 519A, the primary makeup water containment isolation valve, is located outside containment and will automatically shut upon a phase A containment isolation signal (CIS). CV-519B, the PRT makeup valve, located inside containment is operated from the Console by means of a 2-position, CLOSED-OPEN, and hand control switch. This valve does not have an automatic function on CIS. Consequently, according to 4-NOP-041.03 Step 5.3.2, a specific conditional step is added to the PRT temperature reduction procedure stating that IF Containment Isolation or Safety Injection signals actuate, THEN CLOSE CV-4-519B to ensure Containment Integrity.

C. Incorrect. This is incorrect because CV-4-519B must be manually closed. It is plausible because only one valve has an automatic function, and the operator may incorrectly believe that CV-4-51 9B, rather than CV-4-51 9A, has the automatic function.

D. Incorrect. This is incorrect because only CV-4 519B must be manually closed. It is plausible because the operator may incorrectly believe that neither CV-4 519B nor CV-4 519A has the automatic function when the valves are manually open per this procedure.

SD-009 (p39; Rev 9/20/11) 4-NOP.-041.03(plO;Revll)

Technical Reference(s):

(Attach if not previously provided)

Step 5.3.2 Proposed References to be provided to applicants during examination: None 6902109 Objectives 8, 9.c, 10.d Learning Objective: and 10.g (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam: NA Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis

10 CFR Part 55 Content: 55.41 7 55.43 Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Comments:

The KA is matched because the operator must demonstrate Knowledge of PRTS design feature(s) and/or interlock(s) (i.e. how PWS valves function on an SI/CIS Phase A) which provide for Quench tank (PRT) cooling The question is at the Memory (ii) cognitive level because the operator must recall bits of information to answer the question correctly.

Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 1 KJA# 008 K1.01 Importance Rating 3.1 Knowledge of the physical connections and/or cause-effect relationships between the CCWS and the following systems: SWS Proposed Question: RO Question # 34 Plant conditions:

Unit 3 is operating at 100% power.

Three CCW Heat Exchangers are in service.

In accordance with 3-NOP-019, Intake Cooling Water System, under normal conditions which ONE of the following identifies the maximum allowable ICW flowrate to each CCW Heat Exchanger, AND the reason for this limit?

A. 10,000 gpm; Prevent runout of the ICW pump.

B. 10,000 gpm; Minimize long-term tube-side erosion.

C. 3,200 gpm; Prevent runout of the ICW pump.

D. 3,200 gpm; Minimize long-term tube-side erosion.

Proposed Answer: B Explanation (Optional):

A. Incorrect. 1 st part correct, 2nd part wrong. This is incorrect because the 10,000 gpm limit on flow through a single CCW Heat Exchanger is not based on ICW pump runout limits. It is plausible because according to 3-ONOP-019 and BD-ONOP-019 the ICW off normal procedure contains steps to limit total ICW flow when ther4e is only one ICW Pump operating to avoid runout, and the operator may incorrectly believe that the 10,000 gpm limit is related to ICW Pump runout.

B. Correct. 1 st part correct, 2 nd part correct. According to 3-NOP-019, the maximum lOW flowrate to each COW HX during normal operation should NOT exceed 10,000 gpm in order to minimize long term tube side erosion of the COW HXs.

0. Incorrect. s1t part wrong, 2nd part wrong. See A and D.

D Incorrect. 1st part wrong, 2nd part correct. This is incorrect because the maximum allowable lOW flowrate through the CCW Heat Exchanger is not 11,000 gpm. This is plausible because according to 3-NOP-019 Step 2.2.4.3, if three COW HXs are in service, then the minimum required ICW flow rate is 11,000 gpm TOTAL, AND at least 3500 gpm through each COW HX. The operator may confuse the 10,000 and the 11,000 gpm limits. Plausibility is enhanced by maintaining the higher incorrect flowrate below the maximum allowable limit during times of strainer backwash. According to 3-NOP-019, the lOW flowrate for each COW HX may be increased to 12,850 gpm for up to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months period to accommodate HX or Basket Strainer cleanings.

3-NOP-019 (p9; Rev hA) 3-ONOP-019 (p7; Rev 7/23/09)

Technical Reference(s): BD-ONOP-019 (p5; Rev (Attach if not previously provided) 7/23/09)

Proposed References to be provided to applicants during examination: None 6902154 Objective 8.a and 9.b Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam: NA Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 OFR Part 55 Content: 55.41 4 55.43 Secondary coolant and auxiliary systems that affect the facility.

Comments:

The KA is matched because the operator must demonstrate Knowledge of the physical connections between the CCWS and the SWS. Specifically, the operator must know that during normal operation, ICW flowrate through any of the CCW Heat Exchangers is limited to 10,000 gpm to minimize long-term tube-side erosion.

The question is at the Memory (1 P) cognitive level because the operator must recall bits of information (i.e. max ICW flowrate through CCW HX, the reason for the limit), to answer the question correctly.

Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# I KIA# 010 K5.02 Importance Rating 2.6 Knowledge of the operational implications of the following concepts as the apply to the PZR PCS: Constant enthalpy expansion through a valve Proposed Question: RO Question # 35 Plant conditions:

Unit 4 is in the process of a plant heatup.

RCS pressure is 1835 psig.

A Pressurizer Code Safety Valve is leaking by.

PRT pressure is 6 psig.

Which ONE of the following correctly completes the statement below?

The Safety Valve tailpipe temperature on VPA will indicate approximately (1) . With no operator action, the Safety Valve tailpipe temperature indication (2) to determine which Safety Valve is leaking by.

A. (1)230°F (2) can be used B. (1)230°F (2) can NOT be used C. (1)400°F (2) can be used D. (1)400°F (2) can NOT be used Proposed Answer: A Explanation (Optional):

A. Correct. 1 st part correct, 2nd part correct. Expansion from 1850 psia and 624°F in the Pressurizer to 6 psig in the tailpipe will result in a tailpipe temperature of approximately 230°F due to isenthalpic expansion through the Safety Valve. According to SD-009, an

RTD is installed in the discharge pipe of each individual code safety valve and supplies its own individual temperature indicator on VPA. When any one of these three RTDs senses a temperature of 200°F t, it triggers the PRESSURIZER SAFETY VALVE LINE A, B, C HIGH TEMP alarm, common to all three RTDs and located on annunciator A-7/3. The RTD, alarm setpoint, and temperature indicator associated with each safety valve is tabulated in the System Data Section. The alarm is one indication that a safety valve has lifted or is leaking. Generally speaking, the leaking valve can be determined by observing which has the highest tailpipe temperature.

B. Incorrect. 1st part correct, 2nd part wrong. This is incorrect because the Safety Valve tailpipe temperature indication can be used to determine that the valve is leaking. This is plausible because it would be correct if the PORV was leaking. According to SD-009, an RTD, TE-3/4-463, is installed in the combined relief line from both PORVs. It supplies a temperature indicator, TI-3/4-463, on VPA. At 250°Ft, this RTD triggers the PRESSURIZER RELIEF LINE HIGH TEMP alarm on annunciator A, window 7/2. This is an indication that one of the PORVs has lifted or is leaking.

C. Incorrect. 1 st part wrong, 2nd part correct. This is plausible if the operator incorrectly believed that the leakage through the code safety was NOT an isenthalpic process. If so, the operator may conclude that the tailpipe temperature would be closer to the saturation temperature of the fluid being discharged from the valve.

D. Incorrect. jst part wrong, 2nd part wrong. See B and C.

SD-009 (p29, 32-33; Rev Technical Reference(s): 9/20/11) (Attach if not previously provided)

Proposed References to be provided to applicants during examination: N 6902109 Objective 11 Learning Objective: (As available)

Question Source: Bank #

Modified Bank # WTSI 71385 (Note changes or attach parent)

New Question History: Last NRC Exam: 2010-2 Turkey Point Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X

10 CFR Part 55 Content: 55.41 14 55.43 Principles of heat transfer, thermodynamics and fluid mechanics.

Comments:

The KA is matched because the operator must demonstrate Knowledge of the operational implications of a constant enthalpy expansion process through a valve as this apply to the PZR PCS, such as through a leaking Pzr Code Safety Valve.

The question is at the Comprehension/Analysis (2DR) cognitive level because the operator must recall bits of information (i.e. how are the Safety Valve tailpipe RTDs arranged, What type of thermodynamic process is leakage through a valve when the fluid in the system is at saturation, and leaks into to a subcooled system), and then relate this information to its larger setting (i.e. conclude its implications) by using the Steam Tables/Mollier Diagram to answer the question correctly.

The question is significantly modified because the first fill-in choice includes new conditions (RCS & PRT pressure) and the second fill in is now regarding Pzr Safety Valve, rather than a PORV. According to NUREG-1 021, ES-401 Section D.2.f, paragraph 1, bullet 4; to be considered a significantly modified question, at least one pertinent condition in the stem and at least one distractor must be changed from the original bank question. Changing the conditions in the stem such that one of the three distractors in the original question becomes the correct answer would also be considered a significant modification. This question has changed conditions so that a new correct answer exists.

Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 1 KJA# 010 K6.01 Importance Rating 2.7 Knowledge of the effect of a loss or malfunction of the following will have on the PZR PCS:

Pressure detection systems Proposed Question: RO Question # 36 Plant conditions:

Unit 3 is operating at 100% power.

ALL Pressurizer pressure controls are in AUTO.

PT-3-445, Pressurizer Pressure Transmitter, fails HIGH.

Which ONE of the following identifies the effect on the unit?

A. ONLY PZR PORV PCV-3-455C will open. The open PORV will automatically close when pressurizer pressure drops to 2000 psig.

B. ONLY PZR PORV PCV-3-456 will open. The open PORV will automatically close when pressurizer pressure drops to 2000 psig.

C. PZR Spray valves open, PZR PORV PCV-3-455C opens; Reactor trip on low pressurizer pressure.

D. PZR Spray valves open, PZR PORV PCV-3-456 opens; Reactor trip on low pressurizer pressure.

Proposed Answer: B Explanation (Optional):

A. Incorrect. 1 st part wrong, 2nd part correct. PORV (PCV-3-456) will open and be automatically closed at 2000 psig, and this is above the RCS Low Pressure Trip setpoint of 1835 psig. The transient will terminate with the plant at power. This is plausible because the operator may be unaware that PT-3-445 sends a signal to PCV 3-456 B. Correct. 1st part correct, 2nd part correct. According to Lesson Plan 6902109A, PT 444 provides for operation of the spray valves, heaters, and PORV-3-455C while PT-

445 provides input for PORV-3-456. According to Lesson Plan 6902109A the absence of a signal from the low pressure safety injection block permissive circuit representing a pressure of 2000 psig, is essentially an open permissive interlock for the PORVs. The requirement for such an open permissive interlock protects the RCS from a severe depressurization incident should either (or both) pressurizer pressure control channels (PT-3-445 and/or PT-3-444) fail high. If such a failure occurs with the PORV in AUTO, the PORV will open and reduce RCS pressure unnecessarily. However, the open permissive interlock will shut the PORV as soon as pressure falls to 2000 PSIG. The ultimate purpose of interlocking the PORVs with the pressurizer pressure protection channels (PT-3-455, PT-3-456, and PT-3-457) is to keep the RCS in a subcooled condition and thereby preclude voiding in the core. According to Lesson Plan 6902109A, failure of PT-3-445 high would cause PORV (PCV-3-456) to open. Pressure would drop to 2000 PSIG and cycle around that setpoint. Since this pressure is higher than the RCS low pressure trip setpoint, the plant will remain at power.

C. Incorrect. 1 5t part wrong, 2nd part wrong. See A and D.

D. Incorrect. jst part correct, 2nd part wrong. This is incorrect because according to Lesson Plan 6902109A, PT-444 provides for operation of the spray valves, heaters, and PORV-455C. This is plausible because each of the two transmitters controls a PORV, and the operator incorrectly reverse the PORV that each transmitter controls.

Additionally, the reactor will not trip because the PORV will close on the permissive block 6902109A (p25, 27, 51, 53, 69; Rev 11/24/09) 3-ONOP-041.5, (p7; Rev Technical Reference(s): 12/17/07) (Attach if not previously provided) 561 0-T-D-1 6B,Pressurizer Pressure Control Proposed References to be provided to applicants during examination: None 6902109A Objective 8.a Learning Objective: (As available)

Question Source: Bank #

Modified Bank # WTSI 61060 (Note changes or attach parent)

New Question History: Last NRC Exam: Turkey Point 2011

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Comments:

The KA is matched because the operator must demonstrate Knowledge of the effect of a loss or malfunction of PT-445 will have on the PZR PCS.

The question is at the Comprehension/Analysis (3PEO) cognitive level because the operator must recall bits of information (i.e. what pressure transmitter controls what PPCS component),

and then apply this information to a set of plant conditions that have already occurred to answer the question correctly.

The question is significantly modified because the question is looking at the failure of a different transmitter, in a different direction. According to NUREG-1021, ES-401 Section D.2.f, paragraph 1, bullet 4; to be considered a significantly modified question, at least one pertinent condition in the stem and at least one distractor must be changed from the original bank question. Changing the conditions in the stem such that one of the three distractors in the original question becomes the correct answer would also be considered a significant modification. This question has changed conditions so that a new correct answer exists. This question may be considered a new question.

Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 1 KIA# 012 K2.01 Importance Rating 3.3 Knowledge of bus power supplies to the following: RPS channels, components, and interconnections Proposed Question: RO Question # 37 Unit 3 is at 100% power when the following occurs:

Multiple annunciators simultaneously alarm.

The bottom two rows of Reactor Protection Logic Status Lights on 3C05 on VPB go DARK.

Which ONE of the following identifies the bus that has been lost?

A. 3P06 B. 3P07 C. 3P08 D. 3P09 Proposed Answer: C Explanation (Optional):

A. Incorrect. This is plausible because the Channel I instruments are powered from Panel 3P06; and two rows of status lights are powered from one Vital Instrument Bus. The operator may incorrectly believe that this panel powers the bottom two rows of lights.

B. Incorrect. This would be correct if the top two rows of lights were lost. According to 3-ONOP-003.7, Enclosure I is provided to identify the indications lost during a loss of 120V Vital Instrument Panel 3P07. This enclosure lists the top two rows of Reactor Protection Logic Status Lights will be lost if Panel 3P07 is deenergized. This is plausible because the operator may incorrectly reverse the power supplies.

C. Correct. According to Lesson Plan 6902163, status light panels A & B on VPB indicate the status of all of the instrument ioop protection bistables. The light is energized when the bistable reaches the trip setpoint. According to 3-ONOP-003.8, Enclosure I is provided to identify the indications lost during a loss of 120V Vital Instrument Panel

3P08. This enclosure lists the bottom two rows of Reactor Protection Logic Status Lights will be lost if Panel 3P08 is deenergized.

D. Incorrect. This is plausible because the Channel IV instruments are powered from Panel 3P09; and two rows of status lights are powered from one Vital Instrument Bus.

The operator may incorrectly believe that this panel powers the bottom two rows of lights.

Lesson Plan 6902163 (p83; Rev 10/31/07)

Technical Reference(s): 3-ONOP-003.8 (p11; Rev 1) (Attach if not previously provided) 3-ONOP-003.7 (p 1 1; Rev 1)

Proposed References to be provided to applicants during examination: N 6902163 Objective 4 Learning Objective: (As available)

Question Source: Bank # PTN6902 1630509 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: NA Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

The KA is matched because the operator must demonstrate the Knowledge of bus power supplies to the RPS channels, components, and interconnections such as the Reactor Protection Logic Status Lights.

The question is at the comprehension cognitive level because the operator must recall bits of information, (what is on the bottom 2 rows of plant status indication) and then apply this information to determine how those components and status panel lights are powered to answer the question correctly.

Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 1 KIA# 013 K4.19 Importance Rating 3.0 Knowledge of ESFAS design feature(s) and/or interlock(s) which provide for the following Reason for opening breaker on high-head injection pump Proposed Question: RO Question # 38 Plant conditions:

Both units are operating at 100% power.

A spurious Safety Injection Signal occurs on Unit 4.

The crew enters 4-EOP-E-0, Reactor Trip or Safety Injection.

Subsequently:

An undervoltage condition occurs on 3A 4KV Bus.

SI on Unit 4 has NOT been reset.

Which ONE of the following identifies how the 3A HHSI Pump Breaker will respond and the reason for this response?

A. The breaker will remain closed so that when the EDG re-powers the 3A 4KV Bus SI flow will be restored as soon as possible.

B. The breaker will trip open and re-close after a time delay so that the 3A EDG is not overloaded when it re-powers the 3A 4KV Bus.

C. The breaker will trip open and not re-close until closed by the operator so that the operator can control the loading of the 3A EDG when it re-powers the 3A 4KV Bus.

D. The breaker will trip open and can only be manually re-closed if the SI signal on Unit 4 has been actuated for 60 seconds.

Proposed Answer: B Explanation (Optional):

A. Incorrect. This is incorrect because the breaker will auto open on the undervoltage condition. This is plausible because the operator may incorrectly believe that since

there is a LOCA the design of the Sequencer would be such that the HHSI Pump is restored to power as soon as possible after the EDG re-powers the bus; and that this would require that the HHSI Pump breaker remains closed so that the pump motor is simply re-energized upon power restoration.

B. Correct. According to SD-021, all four HHSI pumps will automatically start on a safety injection (5) signal from either unit. Additionally, according to EPU SD-170, Emergency Load Sequencer 3A will actuate loading relay 3413A-3 (Second Load Block) at 3.0 sec after the receipt of the LOCA signal to support the LOCA of the other unit (Train 4A or 4B). Consequently, after the spurious SI, all four HHSI Pumps are running.

According to EPU SD-170, to enable the EDGs to reliably come up to speed and voltage and close on to the 4kV buses all breakers connected to the 4kV buses must be opened (bus stripping). Breakers supplying the required equipment can then be sequentially closed with sufficient time delay between breaker closures to prevent overloading the EDG. According to EPU SD-i 70, Emergency Load Sequencer 3A will respond to a LOCA on the other unit by starting its respective HHSI pump (3A) via relay 34/3A-3. When the LOOP occurs, the bus is stripped by the actuation of the stripping relays. These relays will clear Train A after one second of the initial LOOP. EDG 3A will start and within i5 seconds will be at rated speed and voltage. With the EDG at rated frequency and voltage and the bus clear permissive established, the EDG breaker closes, enabling the Emergency Load Sequencer to commence reloading essential equipment required for emergency reactor shutdown, provided the SI has not been reset. If SI has not been reset, the Emergency Load Sequencer will load equipment for safe reactor shutdown and the HHSI pump 3A. According to EPU SD-i 70, Table 2, a LOCA on the other Unit followed by a LOOP later will initiate Block 2 loading relay 34/3A-3 (3A HHSI Pump), which means that according to Table i the 3A HHSI Pump breaker will automatically re-close at 3 seconds after the sequencing is initiated (i.e.

EDG Breaker Closing).

C. Incorrect. This is incorrect because the breaker will auto re-close upon sequencer actuation. This is plausible because it may be correct if the SI were Reset.

D. Incorrect. This is incorrect because the breaker will auto re-close upon sequencer actuation. This is plausible because the operator may incorrectly believe that such a design feature exists to prevent damage to the HHSI Pumps.

SD-02i (pio-ii; Rev 11/30/il)

EPU SD-i 70 (p32 34-36, 47-Technical Reference(s): . .

(Attach if not previously provided) 48; Rev 9/2i11 i)

Proposed References to be provided to applicants during examination: None 6902157 Objectives 3 6 7 and Learning Objective: (As available) 8.b.7

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam: NA Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Comments:

The KA is matched because the operator must demonstrate the knowledge of ESFAS design feature(s) and/or interlock(s) (i.e. what does the Sequencer do under a specific set of plant conditions) which provide for the reason for opening breaker on high-head injection pump (i.e.

questions requires why sequencer does what it does).

The question is at the Comprehension cognitive level because the operator must recall bits of information (i.e. what occurs to the breaker and why does it occur) and relate it to plant conditions (SI has not been reset) to answer the question correctly.

Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 1 KIA# 013 A3.01 Importance Rating 3.7 Ability to monitor automatic operation of the ESFAS including: Input channels and logic Proposed Question: RO Question # 39 Plant conditions:

Unit 3 is in Mode 3 performing a plant cooldown in accordance with 3-GOP-305, Hot Standby to Cold Shutdown.

The low pressurizer pressure, steamline high differential pressure and the high steam flow Safety Injection signals have been blocked.

Due to a minor transient the following plant parameters are observed:

Pressurizer pressure PT-455 2015 Pressurizer pressure PT-456 2005 Pressurizer pressure PT-457 1995 Channel I Tavg 540°F Channel II Tavg 542°F Channel III Tavg 544°F Which ONE of the following identifies the status of the following three Safety Injection Signals?

Low Pressurizer Steamline High AP High Steam Flow Pressure A. Active Active Blocked B. Active Blocked Blocked C. Blocked Blocked Active D. Blocked Blocked Blocked Proposed Answer: A Explanation (Optional):

A. Correct. According to SD-063 and 3-GOP-305 P&L 4.8.1, the low pressurizer pressure and high steam line differential pressure safety injection signals are allowed to be blocked if pressurizer pressure is less than 2000 PSIG to prevent inadvertent safeguards actuation during the performance of a controlled plant cooldown. These blocks are accomplished by turning both Train A and B manual Block/Unblock switches on VPB to the BLOCK position when 2/3 pressurizer pressure channels are below 2000 PSIG. The BLOCK LOW PZR PRESS SI status window on VPA warns the operator that the coincidence to block these signals has been satisfied. The LOW PZR PRESS SI BLOCKED status window on VPA will light when the signals are blocked. If 2/3 pressure channels sense pressurizer pressure greater than 2000 PSIG, the low pressurizer pressure and high steam line differential pressure safety injection signals will be automatically unblocked. In this case, 2 of 3 pressurizer pressure signals have moved above 2000 psig. Consequently, both the Low Pzr Pressure and High Steam Pressure ziP signals are unblocked, or active. Also according to SD-063, the high steam flow safety injection signal can be blocked using the manual Block/Unblock switches on VPB when 2/3 protection TAVG channels are below 543°F. The BLOCK LOW TAVG SI status window on VPA alerts the operator that the block coincidence has been satisfied. When TAVG is less than 543°F and both Train A and B SI block switches on VPB are taken to BLOCK, the LOW TAVE SI BLOCKED status window on VPA will be lit indicating the high steam flow safety injection signals are blocked.

Any 2/3 TAVG channels above 543°F will automatically unblock the high steam flow safety injection signals. Consequently, the high steam flow Signal is still blocked.

According to 3-GOP-305, procedural steps are taken to address the blocking of the signals.

B. Incorrect. This is incorrect because the Steamline High i2iP signal is Active. This is plausible because the operator may incorrectly believe that this signal is blocked by the Tavg signals, rather than the pressurizer pressure channels.

C. Incorrect. This is incorrect because the signals are reversed. This is plausible because the operator may not know the setpoint and/or logic necessary to automatically unblock each signal; and incorrectly believe that this choice correctly reflects the current signal status.

D. Incorrect. This is incorrect because the low pressurizer pressure and high steamline AP are active, or unblocked. This is plausible because the operator may not know the setpoint and/or logic necessary to automatically unblock each signal; and incorrectly believe that this choice correctly reflects the current signal status.

SD-063 (p68-69; Rev 9/10/11) 3-GOP-305 (p15 and 32-33; Technical Reference(s): 8/4/12) (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None 6902163 Objective 8 and 11 Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam: NA Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Comments:

The KA is matched because the operator must demonstrate the ability to monitor automatic operation of the ESFAS including Input channels and logic. This is accomplished by providing the operator with a set of conditions and pertinent plant parameters (i.e. three channels of Tavg/Pzr pressure); and requiring that the operator use knowledge of the input channels and logic associated with automatic unblocking signals to arrive at the answer.

The question is at the Comprehension/Analysis (3SPK) cognitive level because the operator must recall bits of information (i.e. input channels and logic), and then apply this information to a set of plant existing plant parameters to answer the question correctly.

Examination Outline Cross-reference: Level RD SRO Tier# 2 Group# 1 K!A# 022 A4.02 Importance Rating 3.2 Ability to manually operate and/or monitor in the control room: CCS pumps Proposed Question: RD Question #40 Plant conditions:

Both units are operating at 100% power.

A LOCA occurs on Unit 3 resulting in a Safety Injection actuation.

Which ONE of the following describes which Emergency Containment Coolers will automatically start as a result of the Safety Injection?

A. ONLY3C B. ONLY3Aand3C C. ONLY 3B and 3C D. 3A, 3B, and 3C Proposed Answer: B Explanation (Optional):

A. Incorrect. This is incorrect because two fans minimum will start. This is plausible because the operator may incorrectly believe that only one fan is required to start.

B. Correct. According to SD-029 Unit 3 post EPU requires 2 ECCs operating within one minute of the event to ensure peak containment pressure remains below design. Both 3A and 3C ECCs auto start via the sequencers. 3B ECC (swing) will auto start in the event that either 3A or 3C ECC fails to start or the CCW outlet valve for either ECC fails to open.

C. Incorrect. This is incorrect because the 3A Fan will start. This is plausible because the operator may incorrectly believe that the 3B rather than the 3A Fan start.

D. Incorrect. This is incorrect because the 3B Fan will NOT start. This is plausible because the 3B Fan will start if either the 3A or 3C CCW Outlet valve fails to OPEN.

SD-029 (p23; Rev 2/23/1 2)

Technical Reference(s): (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None 6902129 Objective 7, 8 and 9 Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam: NA Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Comments:

NOTE: It is assumed that the CCS Pump identified in the K/A is the Emergency Containment Cooler, three of which are used at TPNPS during the DBA. The KA is matched because the operator must demonstrate the Ability to manually operate and monitor the ECCS in the control room. This is done predicting how many will start under a specific set of plant conditions (i.e.

monitor).

The question is at the Memory (ii) cognitive level because the operator must recall bits of information (i.e. How ECCs respond to an SI actuation,

Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# I KIA# 026 Al.0I Importance Rating 3.9 Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the CSS controls including: Containment pressure Proposed Question: RO Question # 41 Plant conditions:

A LOCA is occurring on Unit 3.

Containment pressure is 5.5 psig and rising.

All equipment is functioning as designed.

Which ONE of the following identifies the operation of the Containment Spray system as containment pressure rises throughout the event, and once actuation has occurred, the operation of Containment Spray as containment pressure is lowering?

Containment Spray pumps will automatically start...

A. directly on a Containment Pressure HI-HI signal; Both Containment Spray pumps will be stopped as soon as Containment Pressure is below 17 psig.

B. directly on a Containment Pressure HI-HI signal; Containment Spray flow will be reduced to one train running upon transition from E-I, Loss of Reactor or Secondary Coolant.

C. on a Sequencer signal if a Containment Pressure HI-HI signal is present; Both Containment Spray pumps will be stopped as soon as Containment Pressure is below 17 psig.

D. on a Sequencer signal if a Containment Pressure HI-HI signal is present; Containment Spray flow will be reduced to one train running upon transition from E-1, Loss of Reactor or Secondary Coolant.

Proposed Answer: D Explanation (Optional):

A. Incorrect. 1st part wrong, nd 2

part wrong. See B and C.

B. Incorrect. 1st part wrong, 2 part correct. This is incorrect because the CSPs do not start directly on a HI-HI- signal. The HI-HI- signal is received by the sequencer to allow starting on their associated load block. This is plausible because the operator may incorrectly believe that the CSPs are started directly off of the Containment Pressure Switches.

C. Incorrect. 1 st part correct, 2 nd part wrong. This is incorrect because if a LOCA exists the CSPs will remain running until the crew has transitioned out of 3-EOP-E-1. This is plausible because according to 3-EOP-E-1 (p11; Rev 4) Step 12, the crew will stop the CSPs if the Containment pressure was high due to a Secondary Break rather than a LOCA, when the Containment pressure drops to < 17 psig.

D. Correct. 1st part correct, 2 part correct. According to SD-025 (p12; Rev 2/23/1 2),

actual pump starting is via the emergency load sequencer. If HI (4 psi) and Hi-Hi (20 psi) containment pressure is present or occur during the 3rd load block on the sequencer, thesequencerwill startthe CSP (11 to 13 seconds). If HI and Hi-Hi containment pressure occur after the 3rd load block, the sequencer will not allow the CSP to start until the sequencer has timed out and start the pump in block 8 (44 to 46 seconds). During a LOCA under the stated conditions, the crew will enter 3-EOP-E-1, Loss of Reactor or Secondary Coolant, and then either enter 3-EOP-ES-1 .2, Post LOCA Cooldown and Depressurization or 3-EOP-ES-1 .3, Transfer to Cold Leg Recirculation. If 3-EOP-ES-1 .2 is entered first, the crew will eventually enter 3-EOP-ES-I .3 when RWST level drops low enough. According to 3-EOP-ES-1 .3 (p7; Rev 3B)

Step 3, the crew will reduce Containment Spray to one train prior to aligning the ECCS for Cold Leg Recirculation.

SD-025 (p12-1 3; Rev 2/23/1 2)

Technical Reference(s): 3-EOP-ES-1 .3 (p7; Rev 3B) (Attach if not previously provided) 3-EOP-E-I (p11; Rev 4)

Proposed References to be provided to applicants during examination: None 6902125, Obj7 9 Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Comments:

KA is matched because it evaluates whether the applicant knows that containment pressure is not the only criteria in the EOPs for securing containment spray flow.

Question is written at the memory cognitive level because it requires knowledge of the start signal as well as the procedural requirements for reducing spray flow.

Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 1 KIA# 026 2.2.39 Importance Rating 3.9 Equipment Control: Knowledge of less than or equal to one hour technical specification action statements for systems.

Proposed Question: RO Question #42 Plant conditions:

Unit 3 is at 100% power.

Both trains of the Unit 3 Containment Spray System are determined to be inoperable.

If the condition cannot be corrected, which ONE of the following describes the MAXIMUM amount of time for the Unit to be in MODE 3?

A. 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months B. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months C. 7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months D. 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Proposed Answer: C Explanation (Optional):

A. Incorrect. This is incorrect because the amount of time to be in Cold Shutdown is too little. This is plausible because the operator may incorrectly believe that there is no ACTION statement for such a condition; and because of that, LCO 3.0.3 is applicable, and misapply TS 3.0.3 B. Incorrect. This is incorrect because the amount of time to be in Hot Standby is too little. This is plausible because the operator may incorrectly believe that LCO 3.0.3 is applicable and forget that 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months to prepare prior to initiating shutdown will apply. If so, the hour which has already been counted in this case, would apply again, and the operator would incorrectly believe that there are 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months rather than 7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months available to get to Hot Standby.

C. Correct. According to Technical Specification 3.6.2.1 (p4 6-12; Amendment 137 and 3

132), two independent Containment Spray Systems shall be OPERABLE with each Spray System capable of taking suction from the RWST and manually transferring suction to the containment sump via the RHR System. This Technical Specification is applicable during Modes 1-4. If both trains are inoperable, the operator must restore at least one Spray System to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

D. Incorrect. This is incorrect because the amount of time to be in Hot Standby is too much. This is plausible because the operator may know that < 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months ACTION statement exists for the Containment Spray System, but incorrectly believe that it says If both trains are inoperable, the operator must restore at least one Spray System to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in at least HOT STANDBY within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months ..

Technical Specification 3.6.2.1 (p3/4 6-12; Amendment 137

and 132)

Technical Reference(s):

(Attach if not previously provided)

LCO 3.0.3 (p3/4 0-1; Amendment 235 and 230)

Proposed References to be provided to applicants during examination: None 6902125 Objective 11 Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam: NA Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

The KA is matched because the operator must demonstrate Knowledge of less than or equal to one hour technical specification action statements for systems such as Containment Spray.

The question is at the Memory (1 F) cognitive level because the operator must recall bits of information to answer the question correctly.

Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 1 K/A# 039 A4.01 Importance Rating 2.9 Ability to manually operate and/or monitor in the control room: Main steam supply. valves Proposed Question: RO Question #43 Plant conditions:

A plant startup is in progress on Unit 3 per 3-GOP-301, Hot Standby to Power Operation.

The crew is placing the MSRs in service per 3-NOP-072.01, Moisture Separator Reheaters.

The operator has just pressed the START MSR AUTO WARM-UP button on the MSR Warm-Up/Cool-Down Interface on the DCS.

Which ONE of the following identifies actions that will be taken by the operator to complete the start-up of the MSRs?

A. Manually open the MSR Timing Valves; AND Perform a 30 minute soak when the steam supply temperature reaches 50°F above the MSR outlet temperature.

B. Manually open the MSR Timing Valves; AND Manually close the MSR Purge Valves.

C. Manually open the MSR Main Steam Stop MOVs; AND Manually close the MSR Purge Valves.

D. Manually open the MSR Main Steam Stop MOVs; AND Perform a 30 minute soak when the steam supply temperature reaches 50°F above the MSR outlet temperature.

Proposed Answer: C Explanation (Optional):

A. Incorrect. 1 st part wrong, 2nd part correct. See B and D.

B. Incorrect. 1 st part wrong, 2nd part correct. This is incorrect because the timing valves are automatically opened when the operator presses this button. This is plausible because the operator may incorrectly believe that this button initiates the purging, and

that these valves need to be manually opened. Additionally, the Steam Stop MOVs have an AUTO function on their Control Switch, and the operator may incorrectly believe that these valves open when the button is pressed.

C. Correct. 1 5t part correct, 2nd part correct. According to SD-i 04 (p23; Rev 10/26/li) in the reheat steam supply line to each MSR are three valves; (l)the MSR Purge Valves, (2) The MSR Timing Valves, and (3) the MRS Main Steam Stop MOVs. According to 3-NOP-072.0l (p9; Rev 7), a Note prior to Step 4.1.2.4 states that the DCS Program will automatically open the timing valves to 11% OPEN, hold that position for a 30 minute soak, and then ramp MSR outlet temperature until timing valves are fully open.

However, the DCS program does not automatically open the MSR Main Steam Stop MOVs. According to 3-NOP-072.Oi (p ii; Rev 7), Step 4.1.2.10 directs the operator to PLACE MSR MAIN STEAM STOP MOVs control switch to OPEN on the console; after the MSR Timing Valves are fully open and the MSR temperatures have stabilized.

Additionally, according to Step 4.1.2.13, after the MSR Main Steam Stop MOVs have been opened, and their Control Switch placed in AUTO, the operator will be directed to close the MSR Purge Valves.

D. Incorrect. 1 st part correct, 2nd part wrong. This is incorrect because the 30-minute soak is performed automatically. This is plausible because according to 3-NOP-072.Oi (p9; Rev 7), Step 4.1.2.7 directs the operator such that IF a temporary delay during the warmup is desired, THEN DEPRESS the PAUSE button. It continues to direct stating that WHEN desired to continue the warmup, THEN DEPRESS the RESUME button.

The operator may incorrectly believe that the function of the Pause button is to manually perform the soak.

SD-i 04 (p23; Rev 10/26/li)

3-NOP-072.0l (pgs. 9-il; Rev Technical Reference(s): *

(Attach if not previously provided) 7)

Proposed References to be provided to applicants during examination: None

6902117 Objective 7 Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

The KA is matched because the operator must demonstrate the ability to manually operate the MSR Main Steam Supply Valves and Purge Valves to start up the MSRs. This is accomplished by placing the operator in a situation where specific steps have been completed in the MSR startup process, and then requiring that the operator identify steps that are yet to be performed by the operator. The two actions required to be performed are contrasted with automatic actions that are expected to take place.

The question is at the Comprehension/Analysis (2DR) cognitive level because the operator must recall bits of information (How the DCS Auto MSR Warm Up System operates), and then relate this information to a specific point within the MSR startup procedure to answer the question correctly.

Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 1 K/A# 059 K4.19 Importance Rating 3.2 Knowledge of MFW design feature(s) and/or interlock(s) which provide for the following:

Automatic feedwater isolation of MEW Proposed Question: RO Question #44 Unit 3 plant conditions:

A reactor trip occurred from 100% power.

RCS Tave is 547° F and stable Which ONE of the following identifies the Main Feedwater alignment?

A. Main Feedwater Regulating Valves throttled open in auto Main Feedwater Bypass valves open B. Main Feedwater Regulating Valves closed Main Feedwater Bypass valves closed C. Main Feedwater Regulating Valves throttled open in auto Main Feedwater Bypass valves closed D. Main Feedwater Regulating Valves closed Main Feedwater Bypass valves open Proposed Answer: B Explanation (Optional):

A. Incorrect. With a reactor trip signal and Tave below 554°F, the Main Feedwater Control Valves will automatically close. Bypass valves will be closed because on a trip from 100% power, AFW will actuate and feed SGs based on low SG level.

B. Correct. According to SD-063 (p61-62; Rev 9/10/11) the Main Feedwater Control Valves (FCV-478, 488, 498) are closed by any of the following signals: (1) the safety injection signal, (2)the steam generator high-high level signal, or (3)the reactor trip signal in coincidence with low TAVG of 554°B1-. Consequently, when Tavg lowers to less than 554°F, the FCVs close, and wHI not be able to be re-opened unless the FWIS

signal is reset. The Bypass valves will be closed because they are normally closed at power, and they will NOT open on a reactor trip. AFW will be supplying SGs post trip due to low SG levels.

C. Incorrect. See option A discussion of MFRV closure on low Tave with reactor trip signal present. 2nd part correct.

D. Incorrect. 1st part correct. 2nd part incorrect because AFW will be supplying SGs post trip due to low SG levels SD-063 (p61-62 Rev 9/10/11)

Technical Reference(s): (Attach if not previously provided)

Proposed References to be provided to applicants during examination: N 6902163 Objective 8 Learning Objective: 6002122 Objective 8.h (As available)

Question Source: Bank # WTSI 19206 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: 2004 Harris Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Comments:

The KA is matched because the operator must demonstrate the Knowledge of MEW design feature(s) and/or interlock(s) which provide for automatic feedwater isolation of MEW.

The question is at the Comprehension/Analysis (2R1) level because the operator must recall bits of information (i.e. Low Tavg setpoint), and then relate this information to itself to predict an outcome for plant conditions (i.e. reactor trip/SI actuation) to answer the question correctly.

NOTE: This Question was used at Harris in 2004. It has been modified for use at PTN.

Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 1 KIA# 061 A3.05 Importance Rating 2.5 Ability to monitor automatic operation of the AFW, including: Recognition of leakage, using sump level changes Proposed Question: RO Question # 45 Plant conditions:

Unit 3 is operating at 100% power.

The A AFW Pump is OOS.

AFW has been realigned to maintain 2 trains OPERABLE.

Subsequently:

A transient results in a plant trip and automatic actuation of the AFW System.

All AFW System flowrates indicate normal except Train 2 to C Steam Generator which indicates 50 gpm.

The crew suspects leakage in the AFW System.

Which ONE of the following identifies the location in which the ANPO will be initially directed to look for leaks in the AFW System based on the current system alignment?

The ANPO will be directed to look for leaks...

A. Downstream of the Train 2 AFW Flow Control Valve to the C Steam Generator.

B. Between the B AFW Pump and the Train 2 AFW Flow Control Valve to the C Steam Generator, ONLY.

C. Between the C AFW Pump and the Train 2 AFW Flow Control Valve to the C Steam Generator, ONLY.

D. Between the B and C AFW Pumps and the Train 2 AFW Flow Control Valve to the C Steam Generator.

Proposed Answer: B

Explanation (Optional):

A. Incorrect. This is not correct because the flow rates are indicative of a leak upstream of the flow instrument not downstream. This is plausible because there is a large portion of the system piping that is downstream of the flow instrument, and the operator may misinterpret the indications.

B. Correct. According to SD-i 17 (p 1 6; Rev 9/15/11), the three AFW Pumps discharge through their check valves to one of two redundant discharge headers. The administratively controlled, locked open and locked closed valve configuration aligns the pumps so that pump A discharges to the Train 1 feedwater header and pumps B & C normally discharge to the Train 2 feedwater header. However, the conditions state that the A AFW Pump is OOS. According to 3-NOP-075 (p14; RevS) Section 5.1, dual train redundancy is maintained in the AFW System by aligning the C AFW Pump to Train I in the event that A AFW Pump is OOS by aligning the C AFW Pump Discharge Valves per Attachment 5. This Attachment isolates the C AFW Pump from Train 2 (normal alignment) and aligns the C AFW Pump discharge to Train I. Therefore, the B AFW Pump ONLY is aligned to Train 2 under the current conditions. According to Drawing 5613-M-3075, each AFW flow transmitter is located just upstream of its associated train AFW Flow Control Valve. According to SD-I 17 (p18; Rev 9/15/Il), when the AFW system is in the standby mode (auto), the console HICs are set to a predetermined flow rate of 130 GPM as delineated in the plant operating procedures. Consequently, if flow is low to one of the three SGs in one train, the likely cause is an upstream leak robbing flow, or preventing it from passing by the flow transmitter. Since the B and the C AFW Pumps are isolated from each other under the stated conditions, the operator would be directed to check sumps and low lying areas between the B AFW Pump and the Train 2 Flow Control Valve to the C Steam Generator.

C. Incorrect. This is not correct because the C AFW Pump is not aligned to Train 2, it is aligned to Train 1. Consequently, a leak in Train 2, as indicated here, could not result from a leak in the discharge piping of the C AFW Pump. This is plausible because under normal conditions the C AFW Pump is aligned to Train 2, and the operator may incorrectly believe that with the A AFW Pump OOS, the B AFW Pump is aligned to Train I.

D. Incorrect. This is not correct because the C AFW Pump is not aligned to Train 2, it is aligned to Train 1. Consequently, a leak in Train 2, as indicated here, could not result from a leak in the discharge piping of the C AFW Pump. This is plausible because under normal conditions both the B and the C AFW Pumps are aligned to Train 2, and the operator may incorrectly believe that they are under the current conditions as well.

SD-1I7 (p16, 18; Rev 9/1 5/11) 3-NOP-075 (p I4; Rev 5)

Technical Reference(s): 5613-M-3075 Sht I (Rev 17) (Attach if not previously provided) 5613-M-3075 Sht 2 (Rev 13) 5613-M-3075 Sht 3 (Rev 4)

Proposed References to be provided to applicants during examination: None 6902123 Objective 3, 7, 10.a, and Learning Objective: 12.b (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam: NA Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 8 55.43 Components, capacity, and functions of emergency systems.

Comments:

The KA is matched because the operator must demonstrate ability to monitor automatic operation of the AFW, including recognition of leakage. There are no sump level indicators or alarms in the Control Room, consequently, to obtain operational validity the question must focus on the direction given to the AO who will attempt to look for leaks. It is expected that the RO will provide direction to the AO to look for leaks, consequently, the question meets the KA by requiring that the operator identify the direction expected to be given to the AO when looking for leaks.

The question is at the Comprehension/Analysis (2DR) cognitive level because the operator must recall bits of information (i.e. how the system is aligned under the present conditions), and then relate this information to itself (interpret the observed indications under the present conditions) to answer the question correctly.

Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 1 KIA# 062 A2.16 Importance Rating 2.5 Ability to (a) predict the impacts of the following malfunctions or operations on the ac distribution system; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Degraded system voltages Proposed Question: RO Question # 46 Plant conditions:

Both Units are operating at 100% power.

Subsequently:

Systems Operations notifies the station that electrical grid voltage oscillations are expected to occur and affect the station.

Switchyard voltage is expected to swing between 228 and 238KV starting within the next 15 minutes, and lasting for about 45 minutes.

Current Switchyard voltage has dropped to 229 Ky, below the lower administrative limit.

The crew enters 0-ONOP-004.6, Degraded Switchyard Voltage.

Which ONE of the following identifies the action required?

A. Trip the reactor B. Reduce Main Generator load until switchyard voltage is within limits C. Raise excitation on BOTH Unit 3 and Unit 4 Main Generator voltage regulators D. Declare BOTH Startup transformers inoperable Proposed Answer: D Explanation (Optional):

A. Incorrect. With a low voltage notification a reactor trip is not required, as in low frequency.

B. Incorrect. Plausible because reducing load would help with voltage if the voltage problem was local to the facility.

C. Incorrect. Raising excitation is plausible because grid voltage reductions are related to VARs, and raising excitation would raise the amount of VARs out on both units D. Correct. According to 0-ONOP-004.6 (p6; Rev 5/21/08) Step 4, the operator will be directed to Check Unit 3 or 4 in Mode 1-4 and Switchyard Voltage is below 233.0 kV or is forecast to be below 233.0 kV. If so, the operator will be directed to declare both SUTs inoperable and enterTS 3.8.1.1.

0-ONOP-004.6 (p6, Rev Technical Reference(s): 5/21/08) (Attach if not previously provided)

Proposed References to be provided to applicants during examinalion: None 6902957 Objectives 1, 2 and 3 Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

The KA is matched because the operator must demonstrate the Ability to (a) predict the impacts of degraded switchyard system voltages on the ac distribution system; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations.

The question is at the Comprehension/Analysis (3PEO) cognitive level because the operator must recall bits of information (i.e.switchyard voltage limits ONOP entry conditions), and then apply this information to a set of plant conditions to answer the question correctly.

Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 1 KIA# 063 K2.01 Importance Rating 2.9 Knowledge of bus power supplies to the following: Major dc loads Proposed Question: RO Question # 47 Which ONE of the following identifies the DC loads that are affected with a loss of 4D01?

A. EDG 4A, CV-4-2816 AFW Train 1 Feed Flow Control Valve B. EDG 4A, FCV-4-1 13B Blender to Charging Pump Suction C. EDG 4B, CV-4-2831 AFW Train 2 Feed Flow Control Valve D. EDG 4B, FCV-4-1 1 3B Blender to Charging Pump Suction Proposed Answer: C Explanation (Optional):

A. Incorrect. EDG 4A supplied by 4D23 B. Incorrect. Same as A C. Correct. See reference D. Incorrect. FCV-1 1 3B is supplied by 4D23 4-ONOP-003.4 (p4 & 16 Rev 8/8/05)

Technical Reference(s): 4-ONOP-003.5 (p4. & 13, Rev (Attach if not previously provided) 1/21/99 & 2/28/05)

Proposed References to be provided to applicants during examination: N Learning Objective: (As available)

Question Source: Bank #

Modified Bank # WTSI 88579 (Note changes or attach parent)

New Question History: Last NRC Exam: 2009 Turkey Point Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 8 55.43 Components, capacity, and functions of emergency systems.

Comments:

KA is matched because the item requires knowledge of loads from a DC bus Item is written at memory cognitive level because the applicant either knows which loads are supplied from the given DC bus or they do not

Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 1 KIA# 063 K3.01 Importance Rating 3.7 Knowledge of the effect that a loss or malfunction of the dc electrical system will have on the following: ED/G Proposed Question: RO Question # 48 Plant conditions:

Unit 3 is operating at 100% power.

A LOOP occurs with a simultaneous loss of 125 VDC Bus 3D01.

3-EOP-E-0, Reactor Trip or Safety Injection is in progress.

4KV bus 3A is de-energized.

Which ONE of the following identifies the status of the 3A EDG?

A. 3A EDG did NOT auto-start, but can be manually started from the Control Room.

B. 3A EDG did NOT auto-start and cannot be manually started from the Control Room.

C. 3A EDG auto-started but 3A EDG output breaker did NOT close due to loss of DC control power.

D. 3A EDG auto-started and 3A EDG has no output voltage due to loss of DC field flash power.

Proposed Answer: B Explanation (Optional):

A. Incorrect. This is incorrect because the 3A EDG cannot be manually started from the Control Room. This is plausible because may correctly believe that the auto start circuit has failed, but incorrectly believe that Black Start capability from the control room exists.

Black start capability does exist on the 4A EDG.

8. Correct. According to Lesson Plan 6902136 (p 1 03; Rev 9/15/11), the function of the air solenoid valve (CV-3-2023 NB or 2068 NB) is to pass air to begin the starting sequence to start the diesel engine. When an engine start signal is received, the solenoid valve is energized, allowing air from the tanks to pass through the solenoid

valve to the pinion gear end of the lower starting motor. Consequently, if the air start solenoid is not energized, the 3A EDG will NOT start. This is reflected in the ONOP for the loss of the 3D01 (3A) DC Bus. According to 3-ONOP-003.4 (p5; Rev 2, there will be a loss of DC power to EDG 3A. According to 3-BD-ONOP-003.4, the EDG 3A excitation switchgear, start circuit, sequence and lockout relay are de-energized, disabling EDG 3A, when 3D01 is de-energized.

C. Incorrect. This is incorrect because the 3A EDG cannot auto-start. This is plausible because the operator may incorrectly believe that the air start solenoids are de-energize to function; a typical design for ESF equipment required to fail safe. If the operator incorrectly believes that the EDG will auto start, they may incorrectly believe that the 3A EDG output breaker did NOT close due to loss of DC control power since control power is generally supplied via a DC source.

D. Incorrect. This is incorrect because the 3A EDG cannot auto-start. This is plausible because the operator may incorrectly believe that the air start solenoids are de-energize to function; a typical design for ESF equipment required to fail safe. If the operator incorrectly believes that the EDG will auto start, they may correctly believe that the 3A EDG has no output voltage due to loss of DC field flash power (powered from EDO1).

Lesson Plan 6902136 (plo 3

Rev 9/15/11) 3-ONOP-003.4 (p5 Rev 2)

Technical Reference(s): .

(Attach if not previously provided) 3-BD-ONOP-003.4 Rev 12/6/01)

Proposed References to be provided to applicants during examination: None 6902136 Objectives 5.g.4, 6 and Learning Objective: 14.a (As available)

Question Source: Bank # PTN 69021360652 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: NA Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X

10 CFR Part 55 Content: 55.41 7 55.43 Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Comments:

The KA is matched because the operator must demonstrate the Knowledge of the effect that a loss 3D01 (3A) will have on the 3A ED/G when it is required for operation.

The question is at the Comprehension/Analysis (2R1) cognitive level because the operator must recall bits of information and then apply this information to a set of plant conditions to answer the question correctly.

Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 1 KIA# 064 2.2.42 importance Rating 3.9 Equipment Control:: Ability to recognize system parameters that are entry-level conditions for Technical Specifications.

Proposed Question: RO Question # 49 Plant conditions:

Both units are at 100% power.

The ANPO reports the following parameters:

Component Parameter 4A Fuel Oil Storage Tank Level 33,600 gallons 4B Fuel Oil Storage Tank level 36,100 gallons 4A EDG Fuel Oil Day Tank level 270 gallons 4B EDG Fuel Oil Day Tank level 250 gallons Which ONE of the following identifies the operability of the Unit 4 Emergency Diesel Generators (EDG)?

A. Both the 4A and 4B EDGs are INOPERABLE B. Both the 4A and 4B EDGs are OPERABLE C. The 4A EDG is INOPERABLE due to Fuel Oil Storage Tank Level D. The 4B EDG is INOPERABLE due to Fuel Oil Day Tank level Proposed Answer: C Explanation (Optional):

A. Incorrect. All parameters on 4B EDG meet the TS LCO

B. Incorrect. 4A EDG FOST level is below the TS minimum C. Correct. See reference D. Incorrect. 4B EDG Day Tank level remains within limits required by TS 3.8.1.1 TS 3.8.1.1 Technical Reference(s): (Attach if not previously provided)

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

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 2 Facility operating limitations in the technical specifications and their bases.

Comments:

KA match because the question directly requires the applicant to determine which EDG is in an LCO action statement and the reason why Item is written to memory cognitive level because the applicant must know the setpoints for values that require TS LCO entry

Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 1 KIA# 073 A2.02 Importance Rating 2.7 Ability to (a) predict the impacts of the following malfunctions or operations on the PRM system; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Detector failure Proposed Question: RO Question # 50 Plant conditions:

A plant startup is in progress at Unit 3 in accordance with 3-GOP-301, Hot Standby to Power Operation.

Power is at 28%, stabilized for a Chemistry Hold.

Subsequently:

Annunciator H 1/6, PRMS CHANNEL FAILURE, alarms.

R-3-20, Reactor Coolant Letdown Monitor, amber FAIL light is lit.

The detector for R-3-20, Reactor Coolant Letdown Monitor, is failed downscale.

Which ONE of the following identifies the impact, if any, that this failure will have on the power increase, AND the procedure that must be entered?

A. Power level cannot be raised until R-3-20 is returned to service; 3-ONOP-067, Radioactive Effluent Release.

B. Power level cannot be raised until R-3-20 is returned to service; 3-ONOP-41 .4, Excessive Reactor Coolant System Activity.

C. There are no restrictions on power level; 3-ONOP-067, Radioactive Effluent Release.

D. There are no restrictions on power level; 3-ONOP-41 .4, Excessive Reactor Coolant System Activity.

Proposed Answer: C

Explanation (Optional):

A. Incorrect. 1 st part wrong, 2nd part correct. This is incorrect because there is no restriction on power level due to this failure. This is plausible because the operator may incorrectly believe that the plant power is restricted either by procedure, ODCM and or technical specification because of this failure. Several process radiation monitors are connected to the Technical Specifications or ODCM, and the operator may incorrectly believe that R-3-20 is as well. Furthermore, RCS Activity, the process that R-3-20 monitors, is governed by Technical Specification 3.4.8; and the operator may incorrectly believe that the Letdown Monitor is, as well.

B. Incorrect. s 1 t part wrong, 2nd part wrong. See explanations other options.

C. Correct. 1 st part correct, 2nd part correct. According to 3-GOP-301 (p22-23; Rev 12),

prerequisite 3.4.39, the Process Radiation Monitoring System must be in operation to conduct the startup. According to 3-NOP-067 (p 15; Rev 2), R-3-20 is placed in service by performing this procedure, and therefore is in service at the start of the power increase. However, there are no prerequisites listed in 3-OP-47 for having the Letdown Monitor in service, or restrictions placed on power ascent if the Letdown Monitor is removed from service. According to 3-NOP-067 (p20; Rev 2), a Note prior to Step 4.3.2.1 states that the Process Radiation Monitors are normally left in service, and that if a monitor is removed from service, it is done in accordance with the Clearance and Tagging procedure, rather than the NOP. According to 3-ARP-097.CR-H (p9; Rev 3),

Annunciator H116, PRMS CHANNEL FAILURE, will alarm if there is a loss of detector counts for three minutes. Consequently, under the stated conditions this alarm is expected. According to 3-ONOP-067 (p3; Rev 9/27/07) the purpose of the procedure is to provides instructions to be followed in the event of a radiation monitor alarm or malfunction of R-1 1, R-12, R-14, R-17AIB, R-18, and R-20 that could result in a radiation hazard or inadvertent release of radioactivity to the environment. To that end, according to Symptom or Entry Condition 2.3.2, the operator will enter this procedure if Annunciator H1/6 is LIT. Upon entry the operator will address Step 1, and proceed to Step 4 based on the RNO. At Step 4 the operator will proceed to Step 8 based on the RNO. At Step 8 the operator will determine that the detector has failed and notify I&C of the failure.

D. Incorrect. jst part correct, 2nd part wrong. This is incorrect because the entry conditions for ONOP-041 A are not met, and this procedure will not be entered. This is plausible because according to 3-ONOP-041 .4 (p4.; Rev 1) the purpose of this procedure is to provide instruction to be followed in the event of unusually high Reactor Coolant System (RCS) specific activity levels that may be caused by crud bursts, demineralizer resin exhaustion, or fuel element failures. It would be entered on a high alarm associated with R-3-20, not a failed detector. Lending plausibility is the fact that failed detector alarms associated with R-1 5 or R-1 9 would be handled by 3-ONOP-071.2, which also deals with actual SG tube leakage itself. The operator may incorrectly believe that this procedure deals with failed channel alarms for R-3-20 as well.

3-GOP-301 (pgs 15 22,23); Rev Technical Reference(s): . .

(Attach if not previously provided) 12)

3-NOP-067 (p15&20; Rev 2) 3-ARP-097.CR-H (p9; Rev 5) 3-ONOP-067 (p3, 7, 10, 13; Rev 1) 3-ONOP-041.4 (p4; Rev 1) 3-ONOP-071 .2 (p8; Rev 3)

Proposed References to be provided to applicants during examination: None 6918168 Objectives 4 and 7 Learning Objective: 6918242 Objectives 1, 2 and 5 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

The KA is matched because the operator must demonstrate the ability to (a) predict the impacts of a Detector failure on the PRM system. This is done by requiring the operator to identify that there are no operational power restrictions on the plant if R-3-20 is OOS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those a detector failure. This is accomplished by requiring that the operator identify the procedure that is to be used to direct the failed instrument be removed from service, which is not intuitive by the procedure titles (i.e. The operator must know the purpose and major actions of procedures to distinguish).

The question is at the Memory (1 F) cognitive level because the operator must recall bits of information, and then apply this information to a set of plant conditions to answer the question correctly.

Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 1 K/A# 076 A1.02 Importance Rating 2.6 Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the SWS controls including: Reactor and turbine building closed cooling water temperatures.

Proposed Question: RO Question # 51 Plant conditions:

Unit 4 is operating at 100% power with 4A and 4B ICW pumps running. 4C ICW pump is OOS for a shaft replacement. The following occurs:

1 413, ICWP NB/C MOTOR BRG HI TEMP, annunciator actuates

4A ICW pump upper bearing temperature is 195°F and slowly rising

4A ICW pump motor current is 50 amps and slowly rising

The crew enters 4-ONOP-019, INTAKE COOLING WATER MALFUNCTION, and stops the 4A ICW pump.

Total ICW flow is 20,500 gpm after the 4A ICW pump stops.

Which ONE of the following identifies the action required?

The crew will throttle 4-50-401, TPCW HX Outlet Combined ICW Isolation Valve _(1 )_ while maintaining TPCW HX outlet temperature < 110°F.

If attempts are unsuccessful to restore ICW system parameters the crew will (2)

A. (1)shut (2) reduce Unit Load.

B. (1)shut (2) place an additional TPCW Heat Exchanger in service.

C. (1)open (2) reduce Unit Load.

D. (1) open (2) place an additional TPCW Heat Exchanger in service.

Proposed Answer: A Explanation (Optional):

A. Correct. lAW ref 2.

B. Incorrect. Total ICW flow is too high (> 19,000 gpm). Operators should throttle shut on the TPCW HX Outlet causing an increase in TPCW outlet temperature.

C. Incorrect. Operators should reduce Unit Load, not line up another heat exchanger.

D. Incorrect. Operators should throttle the valve shut and should not place another heat exchanger in service

1. ARP-097.CR, CONTROL ROOM ANNUNCIATOR RESPONSE, pages 471-474, rev 04/11/04 Technical Reference(s): 2. ONOP-019, INTAKE (Attach if not previously provided)

COOLING WATER MALFUNCTION, page 7, rev 7/23/09 Proposed References to be provided to applicants during examination: No Learning Objective: (As available)

Question Source: Bank # X Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: 2005 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X

10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

KA is matched because the question asks what actions are required to maintain TPCW temperatures within limits during a malfunction of the service water (ICW) system Question is developed at the Comprehension cognitive level because the applicant must understand system interaction to answer correctly

Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 1 K/A# 076 2.4.31 Importance Rating 4.2 Emergency Procedures I Plan: Knowledge of annunciator alarms, indications, or response procedures.

Proposed Question: RO Question # 52 Plant conditions:

Unit 4 has just tripped due to an automatic Safety Injection actuation.

The crew has entered 4-EOP-E-0, Reactor Trip or Safety Injection.

The crew verifies that both the 4A and the 4B ICW Pumps are running.

Subsequently, the following alarm is received:

14/1 - ICWP A/B/C MOTOR OVERLOAD The 4B ICW Pump amps are observed to be 60 amps.

Which ONE of the following identifies the action that may be taken by the crew in accordance with alarm response procedures?

A. Start the 4C ICW Pump, and then stop the 4B ICW Pump.

B. Start the 4C ICW Pump, and operate with three ICW Pumps.

C. Stop the 4B ICW Pump and verify that the 4C ICW Pump automatically starts.

D. Dispatch an operator to evaluate the status of the 4B ICW Pump, and stop the pump if abnormal conditions are observed.

Proposed Answer: A Explanation (Optional):

A. Correct. According to SD-i 65 (p13; Rev 10/28/11), two ICW pumps are normally running. An overload condition will trigger the ICWP A/B/C MOTOR OVERLOAD alarm on annunciator I, window 4/1 at 52.5 amps. ICWP A/B/C TRIP annunciator I, window 4/2 will occur at 105 amps. According to SD-i65 (p13; Rev 10/28/11) a Safety Injection

Signal will send an auto start signal to ICW pumps on that unit via sequencing action.

A Sequencer will start the A ICW pump. B Sequencer will start the B ICW pump.

C ICW pump will auto-start if either the A or B pump has received a sequencer start signal and is racked out with its breaker open (depending on the alignment of D bus). According to 4-ARP-97.CR.I (p22; Rev 0), when this alarm occurs the operator is directed to confirm that it is valid by checking that the affected ICW Pump amps is 52.5 amps. It should be noted that a reading of amps greater than this value is an abnormal condition for this pump. The operator is then directed to start any standby pump, and shutdown the affected pump. According to 4-EOP-E-0, Attachment 3, (p25; Rev 4),

Step 4, the operator is directed to check that at least two ICW Pumps are running. If at least two ICW Pumps are NOT running the RNO directs the operator to start ICW Pumps to establish at least two running. On the safety Injection signal, the 4A and 4B lOW Pumps will continue to run. When the operator checks their status in E-0, they will be armed with the direction to establish at least two running. However, one of the running pumps will have an abnormal condition, as evidenced by the overload alarm.

Upon receiving this alarm, the operator will address the ARP, and upon verifying that the amperage on the 4B ICW Pump is abnormally high, will start the only standby pump (4C), and then stop the affected pump (4B).

B. Incorrect. This is incorrect because per the overload ARP, the 4B Pump must be shutdown. This is plausible because the overload condition has not reached the point at which the pump will automatically trip on overload, and the operator may incorrectly believe that because of the SIS, it is good practice to leave the affected pump running until this point is reached.

C. Incorrect. This is incorrect because per the overload ARP does not direct that the affected be pump be stopped, and the standby pump be allowed to auto start. This is plausible because the operator may correctly believe that the 4B lOW must be stopped, but incorrectly believe that the 4C pump will automatically start. It will not. The operator may incorrectly believe that it will start, and if so, may believe that this action will result in the fastest method establish two ICW Pumps as required by E-0.

D. Incorrect. This is incorrect because per the overload ARP, there is already sufficient indications available requiring that the 4B Pump be shutdown. This is plausible because according to 4-ARP-97.CR.I (p22; Rev 0), if abnormal conditions do not exist (i.e. amps are <52.5 amps, and the alarm occurred because of a spike) the operator will be directed to dispatch an operator to locally check the pump.

SD-165 (p13; Rev 10/28/11) 4-ARP-97.CR.I (p22; Rev 0)

Technical Reference(s): 4-EOP-E-0, Attachment 3, (p25; (Attach if not previously provided)

Rev 4)

Proposed References to be provided to applicants during examination: None

6902154 Objectives 5.a, 6.a, 7.a, Learning Objective: 10.a and 1O.b (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

The KA is matched because the operator must demonstrate the knowledge of annunciator alarms, indications, or response procedures associated with the Service Water System (ICW) in an Emergency Procedure usage situation.

The question is at the Comprehension/Analysis (3PEO) cognitive level because the operator must recall bits of information (i.e. What is abnormal ICWP amperage, what must be done if abnormal ICWP amperage), and then apply this information to a set of plant conditions to answer the question correctly.

Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 1 KJA# 078 K2.01 Importance Rating 2.7 Knowledge of bus power supplies to the following Instrument air compressor Proposed Question: RO Question # 53 Which ONE of the following identifies the power supply to the 3CM Instrument Air Compressor?

A. Load Center 3A B. Load Center 3C C. Load Center 3E D. Load Center 3G Proposed Answer: C Explanation (Optional):

A. Incorrect. This is plausible because it is a different 480V Load Center. However, it is a Vital Load Center, and the actual power supply is a Non-Vital Load Center. The operator may incorrectly believe that the 3CM IA Compressor is powered from a Vital LC because the Instrument Air Dryer is powered from Vital MCC-3A.

B. Incorrect. This is plausible because it is a different 480V Load Center. However, it is a Vital Load Center, and the actual power supply is a Non-Vital Load Center.

C. Correct. According to SD-155 (p12; Rev 11/18/09) Instrument Air Compressor 3CM is powered from Load Center 3E and Instrument Air Compressor 4CM is powered from Load Center 4E. According to the Breaker List (Rev 626) Instrument Air Compressor 3CM is powered from Load Center 3E (Breaker 34104) and Instrument Air Compressor 4CM is powered from Load Center4E (44104). According to 3-NOP-Ol 3.03, Attachment 6 (plo

Rev 12) Instrument Air Compressor 3CM is powered from Load 2

Center 3E (Breaker 34104).

D. Incorrect. This is plausible because it is a different 480V Load Center; and it is also Non-Vital. The operator may incorrectly believe that the 3CM IA Compressor is

powered from this LC because they may confuse this power supply with the power supply for a Service Air Compressor.

Breaker List (Rev 626)

Technical Reference(s): 3-NOP-Ol 3.03, Attachment 6 (Attach if not previously provided)

(p102; Rev 12)

Proposed References to be provided to applicants during examination: None 6902145 TEC Objective 5.a Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 4 55.43 Secondary coolant and auxiliary systems that affect the facility.

Comments:

The KA is matched because the operator must demonstrate Knowledge of bus power supplies to Instrument air compressor 3CM.

The question is at the Memory (1 F) cognitive level because the operator must recall bits of information to answer the question correctly.

Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 1 KIA# 103 K3.03 Importance Rating 3.7 Knowledge of the effect that a loss or malfunction of the containment system will have on the following: Loss of containment integrity under refueling operations.

Proposed Question: RO Question # 54 Plant conditions:

Unit 3 is in MODE 6.

Core reload is in progress.

RHR Pumps 3A and 3B are running.

Which ONE of the following, by itself, would require suspension of the refueling of the reactor?

A. Residual Heat Removal Pump 3B has tripped and will not restart.

B. Refueling cavity water level was found to be 56 feet, 11 inches.

C. The containment personnel air lock inner and outer doors were damaged during the movement of equipment and will not close D. A manual containment isolation valve was found stuck closed, and the upstream automatic containment isolation valve is inoperable.

Proposed Answer: C Explanation (Optional):

A. Incorrect. Only 1 RHR Pump required if cavity level is above 56 10 B. Incorrect. Cavity level is 1 inch above the minimum required.

C. Correct. If airlocks cannot be closed, then refueling activities must stop D. Incorrect. In this case, inoperability of these valves is in the safe position (closed or capable of being closed)

Technical Specifications 3.9.8 Technical Reference(s): refueling operations (Attach if not previously provided)

Proposed References to be provided to applicants during examination: N Learning Objective: (As available)

Question Source: Bank# WTSI 89141 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: 2009 Kewaunee Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 8 55.43 Components, capacity, and functions of emergency systems.

Comments:

KA is matched because the applicant must understand what loss of refueling integrity means and which failure results in suspension of refueling activities Question requires comprehension of refueling containment integrity to answer correctly because each of the choices may result in suspension of refueling activities under a different set of circumstances

Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 1 K/A# 103 A2.03 Importance Rating 3.5 Ability to (a) predict the impacts of the following malfunctions or operations on the containment system-and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations Phase A and B isolation Proposed Question: RO Question # 55 With Unit 4 operating at 100% power, a LOCA occurs which results in the automatic actuation of Phase A and Phase B Containment Isolation Signals.

Which ONE of the following identifies the impact that this will have on the Reactor Coolant Pump seal cooling, AND identifies an action that must be taken within 4-EOP-E-0 with respect to the RCPs?

A. Thermal Barrier CCW flow has been isolated, ONLY; Leave the RCPs running, reset SI, and re-establish seal injection flow.

B. Thermal Barrier CCW flow has been isolated, ONLY; The RCPs must be stopped.

C. Seal water injection flow path has been isolated AND Thermal Barrier CCW flow has been isolated; The RCPs must be stopped.

D. Seal water injection flow path has been isolated AND Thermal Barrier CCW flow has been isolated; Leave the RCPs running, reset SI, and re-establish seal injection flow.

Proposed Answer: B Explanation (Optional):

A. Incorrect. 1 st part correct, 2nd part wrong. This is incorrect because if Phase B has actuated the RCPs must be tripped. This is plausible because according to 4-EOP-E-0 (p13; Rev 4) Step 9c-g, if seal water return temperatures are < 235°F seal injection flow will be re-established the operator will reset SI, and re-establish seal injection flow.

However, the RCPs will have been stopped by this time.

B. Correct. 1 st part correct, 2nd part correct. According to Lesson Plan 6902108 (pg.

121; Rev 3/30/10)) Phase B isolation will automatically close MOV-626, 730 and 716B.

MOV-716B will isolate CCW flow to all three RCP Thermal Barriers and Bearing Oil Coolers (Upper and Lower). MOV..626 will isolate CCW flow from all three RCP Thermal Barriers. MOV-730 will isolate CCW flow from all three RCP Bearing Oil Coolers (Upper and Lower). Consequently, when a Phase B CIS occurs, the Thermal Barrier CCW flow will be isolated. According to Lesson Plan 6902108 (pg. 121; Rev 3/30/1 0) Phase A isolation will affect processes relating to the RCPs. For instance, Phase A CIS will automatically close MOV-1417 and 1418. These valves will isolate CCW flow to the normal Containment Coolers, and according to Lesson Plan 6902108 (pg. 121; Rev 3/30/1 0), At least one Normal Containment Cooler (NCC) is required to be running to support RCP operation. Additionally, Lesson Plan 6902108 (pg. 121; Rev 3/30/1 0) Phase A isolation will automatically close MOV-381 and 6386. These valves will isolate the common Seal Water Return line from all three RCPs. However, there is no automatic capability to isolate the seal water injection lines. Consequently, upon actuation of Phase A and B CIS, with respect to cooling of the RCP Seal, only the thermal barrier is isolated. According to 4-EOP-E-0 (pFoldout; Rev 4), Item #2, if Phase B CIS is actuated the operator is directed to trip all RCPs. Additionally, the body of the procedure will direct that the operator trip the RCPs under the stated conditions.

According to 4-EOP-E-0 (p 1 3; Rev 4) Step 9, the operator will be directed to check 8

that all RCP Thermal Barrier Alarms are OFF. Since CCW flow has been isolated on the Phase B CIS, the Lo RCP Thermal Barrier flow alarm will be lit. This will drive the crew to the RNO which will direct that the RCPs be tripped.

C. Incorrect. 1 st part wrong, 2nd part correct. This is incorrect because Seal Injection flow has not been isolated, it has been stopped (If seal water return temperatures are <

235°F seal injection flow will be re-established). This is plausible because the operator may incorrectly believe that automatic seal injection flow isolation does exist. It does not. Manual seal water injection isolation exists, and this action is taken under certain plant conditions such as high seal temperature with no injection flow. Additionally, the operator may confuse this with the automatic isolation of the Seal Return line which occurs on Phase A isolation.

D. Incorrect. 1 st part wrong, 2nd part wrong. This is incorrect because if Phase B has actuated the RCPs must be tripped, and because Seal Injection flow has not been isolated, it has been stopped. This is plausible because the operator may incorrectly believe that automatic seal injection flow isolation does exist. It does not. Manual seal water injection isolation exists, and this action is taken under certain plant conditions such as high seal temperature with no injection flow. Additionally, the operator may confuse this with the automatic isolation of the Seal Return line which occurs on Phase A isolation. Furthermore, it is plausible because according to 4-EOP-E-0 (p13; Rev 4)

Step 9c-g, if seal water return temperatures are < 235°F seal injection flow will be re established the operator will reset SI, and re-establish seal injection flow. However, the RCPs will have been stopped by this time.

Lesson Plan 6902108 (pg. 121; Rev 3/30/10) .

Technical Reference(s): .

(Attach if not previously provided) 4-EOP-E-0 (p8-13, and Foldout; Rev 4)

Proposed References to be provided to applicants during examination: None Learning Objective: 6902108 Objective 7, 8, 9 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam: NA Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Comments:

The KA is matched because the operator must demonstrate the ability to (a) predict the impacts of Phase A and B isolation on the containment system. This is accomplished by the operator demonstrating that it is understood that when Phase A and B CIS occur, the essential Containment processes are isolated, and that there are some lines penetrating the containment that cannot be automatically isolated on a Phase A and B CIS. And then, (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of Phase A and B isolation. This is accomplished by requiring that the operator identify the proper action to take with respect to the RCPs within E-0, which will be entered initially on an event causing an automatic Phase A and B isolation.

The question is at the Comprehension/Analysis (2R1) cognitive level because the operator must recall bits of information, and must recognize interaction between systems including consequences and implications, to answer the question correctly.

NOTE: This question could also be connected to 10CFR55.41(b)(10).

Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 2 KIA# 001 K6.13 Importance Rating 3.6 Knowledge of the effect of a loss or malfunction on the following CRDS components: Location and operation of RPIS Proposed Question: RO Question # 56 Plant conditions:

Unit 4 is operating at 100% power.

Annunciator RPIS POWER TROUBLE (F 4/6) alarms.

An operator is sent to determine RPI Inverter voltage.

RPI Inverter Voltage is reported as 105 VDC.

The RPI inverter voltage will be checked in the (1) . In accordance with the Alarm Response Procedure the RPI positions on console must be checked by (2)

A. (1) MCC Room (2) performing a flux map with 0-OSP-059.14, Rod Position Indication (RPI) Verification B. (1)MCCR00m (2) comparing against the Acceptance Criteria contained in 4-OSP-201 .1, RO Daily Logs C. (1)MGSetRoom (2) comparing against the Acceptance Criteria contained in 4-OSP-201 .1, RO Daily Logs D. (1)MGSetRoom (2) performing a flux map with 0-OSP-059.14, Rod Position Indication (RP!) Verification Proposed Answer: C Explanation (Optional):

A. Incorrect. MCC room is nearby the MG Set room and contains the RPI Breaker, but a flux map is not required for this condition B. Incorrect. See option A for location information

C. Correct. See reference D. Incorrect. Flux map not required but plausible because a flux map may be used to determine position of a stuck rod 6900106 (p45; Rev 2/16/10) 4-ARP-097.CR.F (p27; Rev 3) 4-NOP-028.0l (p4-5; Rev 0)

Technical Reference(s): Technical Specification LCO (Attach if not previously provided) 3.1.3.2, ACTION a.1 (p3/4 1-20; Amendment 233)

Proposed References to be provided to applicants during examination: None LP 6900106, Objectives 2.c, 4 and Learning Objective: 5.h (As available)

Question Source: Bank # PTN Bank Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

The KA is matched because the operator must demonstrate Knowledge of the effect of a malfunction (low voltage as indicated by the Trouble Alarm) on operation of RPIS power supply, as well as determine location of RPI components.

The question is at the Memory (1 P) cognitive level because the operator must recall bits of information (i.e. locations, how to check RPIs under alarm conditions) to answer the question

Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 2 KIA# 011 K2.O1 Importance Rating 3.1 Knowledge of bus power supplies to the following: Charging pumps Proposed Question: RD Question # 57 Unit 3 is in a normal electrical lineup.

The normal power supply to the 30 Charging Pump is Load Center (1) , normally powered from Load Center (1)

A. (1) 3H (2) 3C B. (1)3H (2)3D C. (1)3B (2)3C D. (1)3B (2)3D Proposed Answer: B Explanation (Optional):

A. Incorrect. 1 5t part correct, 2nd part correct. This is incorrect because Load Center 30 is the alternate, and non-preferred, source of power to Load Center 3H. It is plausible because it is one of two power supplies to Load Canter 3H that are automatically transferred on a low voltage condition; and the operator may incorrectly believe that Load Center 30, rather than 3D is the preferred source. This is strengthened by the fact that Charging Pump 3A is powered from Load Center 3A, and Charging Pump 3B is powered from Load Center 3B. By this convention one might expect the 30 Charging Pump to be normally powered from Load Center 30.

B. Correct. See reference C. Incorrect. See option A

D. Incorrect. See option A SD-013 (p83; Rev 9/20/11)

SD-140(p64and66 Rev Technical Reference(s): (Attach if not previously provided) 4/12/11)

Proposed References to be provided to applicants during examination: None 6902113 Objectives4.a, 6, 8.j, and Learning Objective: 10.h (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam: NA Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 8 55.43 Components, capacity, and functions of emergency systems.

Comments:

The KA is matched because the operator must demonstrate the Knowledge of bus power supplies to Charging pump 3C; specifically the train that the pump power supply normally receives its power, and the alternate source of power The question is at the Memory (1 F) cognitive level because the operator must recall bits of information to answer the question correctly.

Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 2 K!A# 014 K4.06 Importance Rating 3.4 Knowledge of RPIS design feature(s) and/or interlock(s) which provide for the following:

Individual and group misalignment Proposed Question: RO Question # 58 With the unit at 100% power, which ONE of the following will generate a B9/3 Shutdown Rod Off Top/Deviation alarm?

The position of any.

A. shutdown bank rod below 218 steps while control bank B is above 35 steps.

B. shutdown bank rod misaligned from a shutdown bank rod in a different bank by more than 12 steps.

C. control bank rod misaligned from the group step counter by more than 24 steps moving.

D. control bank rod misaligned from the group step counter by more than 12 steps stationary.

Proposed Answer: A Explanation (Optional):

A. Correct. According to Lesson Plan 6900106 (p35, Rev 2/1 6/1 0), the relative positions of each of the control rods and the absolute position of each of the shutdown rods are monitored by the rod deviation monitor. DC signals, proportional to each rod position, from the rod position indication system are compared to each other. Whenever a deviation from predetermined level exists, a bistable/relay circuit trips to actuate an annunciator and a local indicating light. The relative positions of the control rods are continuously monitored while the rods are in motion and at rest. If any rod in a bank deviates in actual position relative to another rod in that bank by more than 12 steps (7.5) at rest and 24 steps (15) in motion the SHUTDOWN RODS OFF TOP/ROD DEVIATION alarm (B 9/3) is triggered. These deviation limits ensure an acceptable power distribution. The shutdown rods are monitored by another drawer in the cabinet.

This drawer will generate the SHUTDOWN RODS OFF TOP/ROD DEVIATION alarm on annunciator B, window 9/3 when any shutdown rod drops below 218 steps and Control Bank B is greater than 35 steps. According to 4-ARP-097.CR.B (p55; Rev 2)

the SHUTDOWN ROD OFF TOP/DEVIATION Alarm, Any S/D rod below 218 steps and control bank B greater than 35 steps, OR, Deviation of 12(24 moving) steps between any two rods in the same bank.

B. Incorrect. This is incorrect because a difference between a shutdown rod in Bank A and a shutdown rod in Bank B of 12 steps will not cause the alarm. This is plausible because according to 4-ARP-.097.CR.B (p55; Rev 2) the SHUTDOWN ROD OFF TOP/DEVIATION Alarm, a deviation of 12 (24 moving) steps between any two rods in the same bank. The operator may incorrectly believe that the shutdown bank rod in A are compared to the shutdown bank rods in B.

C. Incorrect. This is incorrect because the Rod Deviation Monitor does not compare the Control Bank individual rod position to the step counters. This is plausible because the operator may incorrectly believe that the RDM does compare the individual rod positions to the step counters, and if so, 24 steps and rods moving are conditions associated with the generation of this alarm.

D. Incorrect. This is incorrect because the Rod Deviation Monitor does not compare the Control Bank individual rod position to the step counters. This is plausible because the operator may incorrectly believe that the RDM does compare the individual rod positions to the step counters, and if so, 12 steps and rods stationary are conditions associated with the generation of this alarm.

Lesson Plan 6900106 (p35,

Rev 2/16/10)

Technical Reference(s):

(Attach if not previously provided) 4-ARP-097.CR.B (p55; Rev 2)

Proposed References to be provided to applicants during examination: None

6900106 Objectives 2.f, 5.f, and 6.d Learning Objective: (As available)

Question Source: Bank # PTN 69022070212 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: NA Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis

10 CFR Part 55 Content: 55.41 7 55.43 Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Comments:

The KA is matched because the operator must demonstrate knowledge of RPIS design feature(s) (i.e. Rod Deviation Monitor outputs) and/or interlock(s) (i.e. Group B> 35 Steps enabling SD Bank OFF TOP Alarm) which provide for individual and group control rod misalignment.

The question is at the Memory (11) cognitive level because the operator must recall bits of information to answer the question correctly.

Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 2 KJA# 017 K5.01 Importance Rating 3.1 Knowledge of the operational implications of the following concepts as they apply to the ITM system: Temperature at which cladding and fuel melt Proposed Question: RO Question # 59 Which ONE of the following correctly completes the statements below?

According to 4-EOP-F-0, Critical Safety Function Status Trees, a Core Exit Thermocouple reading of (1) is the minimum temperature to indicate the onset of zirc water reaction and a potential fuel melt situation. The operator, if monitoring the Critical Safety Function Status Trees manually, will determine that this temperature exists by observing (2)

A. (1)700°F (2) at least five of the hottest Core Exit Thermocouples B. (1)700°F (2) the average of the five hottest Core Exit Thermocouples C. (1) 1200°F (2) the average of the five hottest Core Exit Thermocouples D. (1) 1200°F (2) at least five of the hottest Core Exit Thermocouples Proposed Answer: D Explanation (Optional):

A. Incorrect. 1st part wrong, 2nd part correct. This is incorrect because 700°F is too low to sufficiently indicate the onset of zirc water reaction and a potential fuel melt situation in accordance with 4-EOP-F-0. This is plausible because the status tree for Core Cooling does contain a 700°F threshold; and the operator may incorrectly believe that this is the threshold listed in 4-EOP-F-0 as the point at which the fuel cladding starts to melt at this point.

B. Incorrect. s 1 t part wrong, 2nd part wrong. See A and C.

C. Incorrect. 1 st part correct, 2nd part wrong. This is incorrect because when evaluating CETs to determine if an inadequate Core Cooling condition exists, the average of the five hottest CETs is not used. This is plausible because the operator may incorrectly believe that the average of the five hottest is used, rather than at least five of the hottest CETs.

D. Correct. 1 st part correct, 2nd part correct. According to BD-EOP-F-0 (p21; Rev 4/15/99) a CET temperature of 1200°F indicates the possible onset of Zirc-Water reaction (i.e. fuel cladding damage). According to 4-EOP-F-0, Enclosure 2 (p8; Rev 12/9/08), when the operator is evaluating the Core Cooling Critical Safety Status Tree, the operator is directed to Obtain core exit temperature using at least five of the hottest core exit thermocouples. If the CETs are >1200°F, the operator is directed to go to FR C.1 which deals specifically with an Inadequate Core Cooling Situation. A threshold of 700°F of this Status Tree will direct the operator to Degraded Core Cooling procedures.

BD-EOP-F-0 (p22; Rev 4/15/99) 4-EOP-F-0, Enclosure 2 (p8; Technical Reference(s): (Attach if not previously provided)

Rev 12/9/08)

Proposed References to be provided to applicants during examination: None 6910922 Objective 6 Learning Objective: 6910347 Objectives 1 and 2 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

The KA is matched because the operator must demonstrate Knowledge of the operational implications of the following concepts as they apply to the ITM system, specifically the temperature at which cladding and fuel melt. The is achieved by requiring the operator identify the beginning of core melt, the temperature at which fuel cladding starts to melt; and then identify the process used by the operator with respect to the ITM System, to evaluate the CETs.

The question is at the Memory (IF or 1 P) cognitive level because the operator must recall bits of information to answer the question correctly.

F

Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 2 K/A# 029 A1.03 Importance Rating 3.0 Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the Containment Purge System controls including: Containment pressure, temperature, and humidity Proposed Question: RO Question # 60 Given the following:

Unit 3 is in Mode 6.

A Core offload is in progress.

The Containment Equipment Hatch and Personnel hatch are closed.

Containment Purge is operating with Purge supply and Exhaust fans running.

A fuse blows for POV-3-2601, Containment Purge Supply Isolation.

Which ONE of the following identifies the concern due to this condition?

A. Spent Fuel Pool Level rises.

B. Containment Pressure lowers.

C. Containment Purge Supply Fan trips.

D. Containment Purge Exhaust Fan trips.

Proposed Answer: B Explanation (Optional):

A. Incorrect. This is incorrect because a decrease in Containment pressure will not cause SEP Level to rise. This is plausible, because it will have the opposite effect. According to 3-NOP-053 (p4; Rev 2) Precaution 2.1.2, If 3-12-031, SFP TRANSFER CANAL GATE VALVE, is open, the refueling cavity is filled with water, and refueling integrity is established; then a slow increase in containment pressure will occur and cause the SEP level to increase when the Containment Purge System is secured. The operator may confuse the concepts.

B. Correct. According to SD-029 (p16; Rev 2/23/12) the Containment Purge system

consists of two (2) purge supply fans and associated ductwork which supply outside air to containment. Two (2) purge exhaust fans pull containment air through roughing filters and discharge the air to the plant stack. One supply and exhaust fan is provided for each unit. During normal operation one Purge Supply and one Purge Exhaust is operating. If the supply Containment Isolation Valve were to fail closed, the Purge Supply Fan would be effectively isolated from the Containment, while the Purge Exhaust Fan continued to run. This would tend to lower Containment Pressure.

C. Incorrect. This is incorrect because Containment Purge supply fan does not trip under the stated conditions. This is plausible because the fan would be running without a suction source and the operator might incorrectly believe the fan will trip to prevent damage.

D. Incorrect. This is incorrect because this fan will not trip when the supply valve closes.

This is plausible because according to Drawing 5613-M-3053, the operation of the Exhaust Fan discharge damper is connected to the operation of the Exhaust Fan. The operator may incorrectly believe that the Exhaust Fan will trip if other dampers close as well.

SD-029 (p16; Rev 2/23/1 2)

Drawing 5614-M-3053 Technical Reference(s):

(Attach if not previously provided) 3-NOP-053 (p4; Rev 2)

Proposed References to be provided to applicants during examination: None 6902129 Objective 7 Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 9 55.43

Shielding, isolation, and containment design features, including access limitations.

Comments:

The KA is matched because the operator must demonstrate the ability to predict and/or monitor changes in parameters associated with operating the Containment Purge System controls including Containment pressure, temperature, and humidity. This is accomplished by creating a situation in which the Containment Purge System is aligned for normal operation, creating a malfunction, and requiring that the operator predict its effect on the Containment conditions.

The question is at the Comprehension/Analysis (2R1) cognitive level because the operator must recall bits of information (What is the normal configuration of the CPS during operation), and then relate back itself by predicting the effect of a system malfunction, to answer the question correctly.

Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 2 K/A# 041 A4.02 Importance Rating 2.7 Ability to manually operate and/or monitor in the control room: Cooldown valves Proposed Question: RO Question # 61 Plant conditions:

Unit 4 is cooling down in accordance with 4-GOP-305, Hot Standby to Cold Shutdown.

Cooldown is proceeding with Steam Dumps to the Condenser.

Taveis400°F.

The Steam Dump Mode Select Switch is in MANUAL.

Hagan Output is 40%.

Which ONE of the following identifies the mode of operation of the Steam Dump to Condenser Hagan Station, AND how many steam dump valves will indicate open?

The Hagan controller is in (1)

(2) steam dump valves indicate open.

A. (1)AUTO (2) Four B. (1)AUTO (2) Two C. (1)MANUAL (2) Four D. (1) MANUAL (2) Two Proposed Answer: D Explanation (Optional):

A. Incorrect. 1st part wrong, 2nd part correct. This is incorrect because the GOP directs that the operator place the Hagan in MAN for the Cooldown. This is plausible because

the Hagan has both a Man and AUTO mode of operation, and the controller can be operated in the Auto mode by adjusting the setpoint knob to a lower pressure throughout the cooldown.

B. Incorrect. 1 st part wrong, 2nd part wrong. See A and D.

C. Correct. 1 st part correct, 2nd part correct. According to 4-GOP-305 (pZ7-28; Rev II)

Step 5.3.4, the Steam Dump to Condenser Pressure Controller (Hagan) is placed in MAN, and the operator slowly opens the Steam Dump Valves. According to Lesson Plan 6902118 (p33; Rev 11/29/07) all three steam dump controllers (Hagan included) provide a 4-20 milliamps output. The first valve, CV-2827, opens at approximately 4-8 ma. The second valve, CV-2828, opens at approximately 8-12 ma. The third valve, CV-2829, opens at approximately 12-16 ma. The fourth valve, CV-2830, opens at approximately 16-20 ma. Actual ma range is shown on Figure 6. According to Lesson Plan 6902118 (p88; Rev 11/29/07) Figure 6, Valve 2827 will be fully open, and Valve 2828 will be approximately 70% open.

D. Incorrect. 1st part correct, 2nd part wrong. This is incorrect because the steam dump valves modulate open (i.e. operate sequentially), even when they are operated by the Plant Tavg or Load Reject controllers. This is plausible because the operator may incorrectly believe that with the Steam Dump Mode Select Switch in MANUAL, they operate together.

4-GOP-305 (p27-28; Rev 11)

Lesson Plan 6902118 (p33 and Technical Reference(s): (Attach if not previously provided) 88; Rev 11/29/07)

Proposed References to be provided to applicants during examination: None 6902118 Objective 4.b and 5.d Learning Objective: 6902408 Objective 4.b (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge

Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

The KA is matched because the operator must demonstrate the ability to manually operate and/or monitor the Steam Dump valves in the control room. This is accomplished by requiring the mode of operation required by procedure for the valve controller, and then, given a controller output, identify the expected valve positions.

The question is at the Comprehension/Analysis (2DR) cognitive level because the operator must recall bits of information, and then relate this information to itself in order to answer the question correctly.

Note: This question could be connected to 10CFR55.41(b)(7)

Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 2 K/A# 045 A2.12 Importance Rating 2.5 Ability to (a) predict the impacts of the following malfunctions or operation on the MT/G system; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Control rod insertion limits exceeded (stabilize secondary)

Proposed Question: RO Question # 62 Plant conditions:

Unit 3 is at 100% power, all systems in normal alignments.

A turbine runback occurs.

The unit is stabilized at 82% power.

Annunciator B 8/2, ROD BANK AIB/C/D EXTRA LO LIMIT is in alarm.

Control Bank D indicates 130 steps.

Which ONE of the following identifies the consequences of the alarm received, AND the actions required by the alarm response procedure to mitigate the condition?

A. The technical specification LCO for Rod Insertion Limits is exceeded; Commence emergency boration in accordance with 3-ONOP-46.1, Emergency Boration.

B. The technical specification LCO for Rod Insertion Limits is exceeded; Immediately borate equal to or greater than 16 gpm using 0-OP-046, CVCS - Boron Concentration Control.

C. The technical specification LCO for Rod Insertion Limits is NOT exceeded; Immediately borate equal to or greater than 16 gpm using 0-OP-046, CVCS - Boron Concentration Control.

D. The technical specification LCO for Rod Insertion Limits is NOT exceeded; Commence emergency boration in accordance with 3-ONOP-46.1, Emergency Boration.

Proposed Answer: B Explanation (Optional):

A. Incorrect. 1st part correct, 2nd part wrong. See B and D.

B. Correct. See reference C. Incorrect. 1st part wrong, 2nd part correct.

D. Incorrect. 1st part wrong, 2nd part wrong. This is incorrect because the normal boration, rather than the emergency boration, procedure is used. This is plausible because there are two procedures that govern the addition of boron to the RCS, and the operator may incorrectly believe that the emergency boration procedure is the one that must be used.

3-ARP-097.CR.B Alarm Response Procedure (p 47 rev 6) 3-ONOP-046.1 (p6-7; Rev 3 Technical Reference(s): Draft) (Attach if not previously provided)

Technical Specification LCO 3.1.3.6 (p3!4 1-26; Amendment 238 and 161)

Proposed References to be provided to applicants during examination: None

6910248 Objectives4and6 Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

The KA is matched because the operator must demonstrate the Ability to (a) predict the impacts of the control rod insertion limits exceeded while stabilizing the secondary plant on the MT/G system. This will be achieved by requiring the operator to identify the time limit beyond which the alarm condition must be corrected, to avoid further impact on the MT!G System; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of the alarm. This is accomplished by choosing the correct procedure to borate the rods out.

The question is at the Memory (1 P) cognitive level because the operator must recall bits of information, and then apply this information to a set of plant conditions to answer the question correctly.

Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 2 KIA# 056 2.1.28 Importance Rating 4.1 Conduct of Operations: Knowledge of the purpose and function of major system components and controls.

Proposed Question: RO Question # 63 Which ONE of the following describes the purpose and operation of CV-3-201 1, Low Pressure Heater Bypass Valve?

Automatically bypasses LP heaters on low feedwater pump suction pressure setpoint of...

A. 250 psig; automatically closes when feedwater suction pressure is restored.

B. 250 psig; must be manually closed when feedwater suction pressure is restored.

C. 210 psig; automatically closes when feedwater suction pressure is restored.

D. 210 psig; must be manually closed when feedwater suction pressure is restored.

Proposed Answer: B Explanation (Optional):

A. Incorrect. This is incorrect because the valve does not automatically close, it only opens automatically.

B. Correct.

C. Incorrect. According to SD-i 12 the Low Pressure Feedwater Heaters can be bypassed by flow through CV 2011 in the event of low suction pressure to the feedwater pumps. Control valve CV-4-20i 1 is actuated to open by pressure switch PS-4-20i I or PS-4-2014 when feedwater pump suction pressure drops to 250 psig. CV-3-201 I is also automatically opened by PS-3-1604 with a fast load reduction. The valve does not automatically close, but must be closed manually by the operator.

D. Incorrect. 2nd part correct. See C for first part. 210 psig is plausible because it is low FWP suction pressure trip

SD-I 12 (p31 and 53; Rev Technical Reference(s): 9/27/11) (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None 6902122 Objectives 7.f and 8.c Learning Objective: (As available)

Question Source: Bank# 69021220625 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 4 55.43 Secondary coolant and auxiliary systems that affect the facility.

Comments:

The KA is matched because the operator must demonstrate Knowledge of the purpose and function of major system components and controls, such as that of the LP Heater Bypass Valve.

The question is written to the memory cognitive level as it requires the applicant to recall a setpoint and know how the component is reclosed

Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 2 KIA# 072 A3.01 Importance Rating 2.9 Ability to monitor automatic operation of the ARM system, including: Changes in ventilation alignment Proposed Question: RO Question # 64 Which ONE of the following states how the high alarms on RAD-6643, Control Room Ventilation Radiation Monitor, and R-1 2, Containment Air Radiation Monitor, impact Control Room Ventilation Isolation and Containment Purge Isolation?

(1) RAD-6643 will actuate Control Room HVAC isolation...

(2) R-12 will actuate...

A. (1) and Containment Purge Isolation; (2) Control Room HVAC isolation and Containment Purge Isolation.

B. (1) ONLY; (2) Control Room HVAC isolation and Containment Purge Isolation.

C. (1) and Containment Purge Isolation; (2) Containment Purge Isolation, ONLY.

D. (1) ONLY; (2) Containment Purge Isolation, ONLY.

Proposed Answer: B Explanation (Optional):

A. Incorrect. 1st part wrong, 2nd part correct. This is incorrect because RAD-6643 will not actuate CVI. This is plausible because the operator may incorrectly believe that each instrument actuates both signals.

B. Correct. 1 st part correct, 2nd part correct. According to Lesson Plan 6902168 and Drawing 5610-T-L1 Sheet 11, Control Room Ventilation Air Intake Monitor RAI-6643 monitors the normal air intake to the Control Room, and will isolate the Control Room from outside air, and place it on recirculation, if the instrument detects 2mrem/hour.

According to Lesson Plan 6902168 and Drawing 5610-T-L1 Sheet 11, Containment

Radioactive Gas Monitor, ensures releases during containment purge are below limits, and will actuate CVI and isolate the Control Room from outside air, and place it on recirculation, if the instrument setpoint is reached.

C. Incorrect. 1 st part wrong, 2nd part wrong. This is incorrect because RAI-6643 will not actuate CVI, and because R-12 will actuate Control Room HVAC isolation. This is plausible because one of these two instruments provides isolation for both the CR and Containment, while the other provides isolation for only one of the two. This choice has reversed the two functions.

D. Incorrect. 1 st part correct, 2nd part wrong. This is incorrect because R-1 2 will actuate Control Room HVAC isolation. This is plausible because the operator may incorrectly believe that each instrument actuates the ventilation isolation for the area in which it provides monitoring only.

6902168 (p67, 69, and 108; Rev 2/26/10)

Technical Reference(s): .

(Attach f not previously provided)

Drawing 561 O-T-L1 Sheet 11 Proposed References to be provided to applicants during examination: None 6918168 Objective 4 Learning Objective: (As available)

Question Source: Bank # X Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: 2009 Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10CFRPart55Content: 55.41 11 55.43 Purpose and operation of radiation monitoring systems, including alarms and survey equipment.

Comments:

The KA is matched because the operator must demonstrate the Ability to monitor automatic operation of the ARM system, including changes in ventilation alignment. This is accomplished by requiring the operator to consider two radiation channes, one an area monitor, and identify the automatic ventilation signals that would be actuated by each should each instrument enter the alarm condition.

The question is at the Memory (ii) cognitive level because the operator must recall bits of information to answer the question correctly.

Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 2 K/A# 075 K1.01 Importance Rating 2.5 Knowledge of the physical connections and/or cause-effect relationships between the circulating water system and the following systems: SWS Proposed Question: RO Question # 65 Plant conditions:

Unit3ismode5.

Unit 4 is operating at 28% power during a downpower due to the inability to maintain all Intake Screens clean.

Subsequently, the Shift Manager has evaluated the effectiveness of the Intake Screens and decides that they cannot be maintained clean enough to support the Intake Cooling Water and Circulating Water Systems at Unit 4.

In accordance with 4-ONOP-Ol 1, Screenwash System/Intake Malfunction, which ONE of the following describes the action that should be taken at Unit 4, AND states the reason for this action?

A. Trip the Turbine ONLY, and stop all four Circulating Water Pumps; Maintain operability of the ICW System.

B. Trip the Reactor and Turbine, and stop all four Circulating Water Pumps; Maintain operability of the ICW System.

C. Trip the Turbine ONLY, and stop all four Circulating Water Pumps; Prevent a loss of Main Condenser vacuum D. Trip the Reactor and Turbine, and stop all four Circulating Water Pumps; Prevent a loss of Main Condenser vacuum Proposed Answer: B Explanation (Optional):

A. Incorrect. 1 st part wrong, 2nd part correct. This is incorrect because the reactor must be tripped as well. This is plausible because according to SD-063 anytime the turbine is

automatically or manually tripped above the P-7 permissive setpoint, the reactor is tripped. According to SD-063, Table 4, P-7 enables the AT Power Reactor Trips when power level is> 10%. Since power level is at 8%, the Turbine Trip will not trip the reactor. The operator may incorrectly believe that all that is required at this power level is to trip the Turbine.

B. Correct. 1 st part correct, 2nd part correct. According to 4-ONOP-011, Item #6, IF the Intake Screens can NOT be maintained clean enough to support operation of the Intake Cooling Water System and Circulating Water System, THEN the Reactor and Turbine shall be tripped AND all four Circulating Water Pumps shall be stopped to maintain the Intake Cooling Water System.

C. Incorrect. s 1t part wrong, 2nd part wrong. See A and D.

D. Incorrect. 1 st part correct, 2nd part wrong. This is incorrect because the reason for taking the action is to maintain the ICW System operable which according to SD-i 65 has several safety or quality related functions. This is plausible because the operator may incorrectly believe that the action s taken mainly to reduce the flow through the screens to ensure enough Circ Water flow to prevent loss of vacuum. While a large reduction will take place, the action is taken so that the Circulating Water Pumps will not rob the limited amount of water within the intake, at the expense of this water not being available to support the safety related functions of the ICW System.

4-ONOP-Oli (Foldout; Rev 1)

SD-i65 (p6; Rev 10/28/li)

Technical Reference(s): SD-063 (p52; Rev 9/10/11) (Attach if not previously provided)

SD-063, Table 4 (p103; Rev 9/10/li)

Proposed References to be provided to applicants during examination: None 6918152 Objectives ii and i2.b Learning Objective: 6902281 Objectives 4 and 6 (As available)

Question Source: Bank #

Modified Bank # 71021520305 (Note changes or attach parent)

New Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

The KA is matched because the operator must demonstrate Knowledge of the cause-effect relationships between the circulating water system and the ICW System, or SWS. This is done by requiring the operator to indicate why the specific action of stopping all four CWPs is taken under plant conditions in which the intake screens cannot be kept clean (i.e. action taken to ensure that ICW System remains operable).

The question is at the Memory (1 P) cognitive level because the operator must recall bits of information to answer the question correctly.

NOTE: The Question may be classified as a new question rather than a Significantly Modified Question.

NOTE: The question is significantly modified because the question now includes specific conditions, and modified answers. According to NUREG-1 021, ES-401 Section D.2.f, paragraph 1, bullet 4; to be considered a significantly modified question, at least one pertinent condition in the stem and at least one distractor must be changed from the original bank question. Changing the conditions in the stem such that one of the three distractors in the original question becomes the correct answer would also be considered a significant modification.

Examination Outline Cross-reference: Level RO SRO Tier# 3 Group# 1 K!A# G1 2.1.14 importance Rating 3.1 Conduct of Operations: Knowledge of criteria or conditions that require plant-wide announcements, such as pump starts, reactor trips, mode changes, etc.

Proposed Question: RO Question # 66 Plant conditions:

Unit 3 is operating at 100% power.

A Steam Generator Tube Leak is in progress on the 3A Steam Generator.

The crew is implementing 3-ONOP-071.2, Steam Generator Tube Leakage.

The crew is preparing to conduct a load reduction.

Which ONE of the following identifies a plant wide announcement that must be made for the SG tube leakage, AND the load reduction?

For the Steam Generator Tube Leak the crew must notify plant personnel of A. any potential hazardous effluent releases; Additionally, notify plant personnel of the load reduction prior to initiating it.

B. the entry into 3-ONOP-071 .2; Additionally, notify plant personnel of the load reduction prior to initiating it.

C. the entry into 3-ONOP-071 .2; There is no requirement to notify personnel of the load reduction.

D. any potential hazardous effluent releases; There is no requirement to notify personnel of the load reduction.

Proposed Answer: A Explanation (Optional):

A. Correct. 1 st part correct, 2nd part correct. According to 3-ONOP-071.2 Step 14, prior to commencing the load reduction the operator is directed to notify plant personnel of the load reduction. This requirement is reiterated in 3-GOP-100. According to 3-GOP-100 Step 3, the operator is directed to notify plant personnel of the load reduction, prior

to the unit load reduction is initiated in the next step. There are no other requirements in the procedure to make any additional page announcements.

B. Incorrect. jst part wrong, 2nd part correct. This is incorrect because the identification of the procedure that has been entered is not needed to be communicated to plant personnel. This is plausible because the operator may incorrectly believe that this information must be communicated.

C. Incorrect. 5 1t part wrong, 2nd part wrong. See B and D.

D. Incorrect. 1 st part correct, 2nd part wrong. This is incorrect because the only requirement for a page announcement is prior to initiating the load reduction. There is no requirement to continue announcing that the load reduction is occurring or on-going.

This is plausible because the operator may incorrectly believe that this information must be communicated.

3-ONOP-071.2 (p13; Rev 7/26/12)

Technical Reference(s): 3-GOP-i 00 (p5; Rev 8/7/12) (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None 6910236 Objectives 4 and 6 Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam: NA Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

The KA is matched because the operator must demonstrate Knowledge of criteria or conditions that require plant-wide announcements, such as those required during a SGTL and prior to initiating a fast load reduction.

The question is at the Memory (1 P) cognitive level because the operator must recall bits of information (3-ONOP-071 .2 Foldout Page announcements, load reduction announcements in ONOP and GOP for fast load reduction) to answer the question correctly.

Examination Outline Cross-reference: Level RO SRO Tier# 3 Group# 1 KIA# G1 2.1.38 Importance Rating 3.7 Conduct of Operations: Knowledge of the stations requirements for verbal communications when implementing procedures.

Proposed Question: RO Question # 67 In accordance with O-ADM-21 1, Emergency and Off-Normal Operating Procedure Usage, which ONE of the following identifies a procedure that when entered from 3-EOP-E-O, Reactor Trip or Safety Injection requires that a crew brief be conducted?

A. 3-EOP-ECA-O.O, Loss of All AC B. 3-EOP-FR-S.1, Response to Nuclear Power Generation/ATWS C. 3-EOP-E-3 Steam Generator Tube Rupture D. 3-EOP-E-2, Faulted Steam Generator Isolation Proposed Answer: D Explanation (Optional):

A. Incorrect. This is incorrect because the transition brief is prohibited. According to 0-ADM-21 1 crew briefings should NOT be performed during critical evolutions that may delay event mitigation and cause further plant degradation such as transitioning from E 0 to ECA-0.0. The first priority of the crew should involve the restoration of power to a Safeguards Bus. This is plausible because ECA-0.0 is a procedure to which a transition can be made from E-0, and the operator may incorrectly believe that a crew brief is required.

B. Incorrect. This is incorrect because the transition brief is prohibited. According to 0-ADM-21 1 crew briefings should NOT be performed during critical evolutions that may delay event mitigation and cause further plant degradation such as transitioning from E 0 to FR-S.1 until the plant is stabilized. The first priority of the crew should be to stabilize the plant after the trip. This is plausible because FR-S.1 is a procedure to which a transition can be made from E-0, and the operator may incorrectly believe that a crew brief is required.

C. Incorrect. This is incorrect because the transition brief is prohibited. According to 0-ADM-21 1 crew briefings should NOT be performed during critical evolutions that may delay event mitigation and cause further plant degradation such as transitioning from E o to E-3. The first priority of the crew should involve actions to prevent ruptured SG release to the environment. This is plausible because E-3 is a procedure to which a transition can be made from E-O, and the operator may incorrectly believe that a crew brief is required.

D. Correct. According to ODI-CO-028, the number and timing of crew briefs will be situational, depending on the event in progress and the crews performance as a team.

Even though a transient is in progress, careful selection of the proper conditions and time will permit a crew brief. Crew Briefs should be held periodically during the event or transient in situations such as EOP TRANSITIONS, just before a significant series of high level steps (prior to interrupting flow to the core in ES-i .3), and when significant plant/equipment status that has been reported to a crew member and should be communicated to the crew. Crew briefs should not be performed after E-0, Step 4 if an SI has occurred or is required. The brief should occur at the transition from E-0.

According to 0-ADM-2i 1, during the mitigation of an off normal or emergency event, it is imperative to conduct a crew briefing to ensure that the Control Room Team is aware of the plant status and event mitigating strategy. Briefings should be performed when operator actions to mitigate plant transients are NOT in progress. Briefings should NOT be performed prior to the initial review of CSFSTs is completed. According to 3-EOP-E-0, the monitoring of the CSFSTs is done prior to the Transition to E-2, and there will be no actions in progress on the transition. A Transition brief is required because the actions necessary to mitigate and stabilize the plant were taken in 3-EOP-E-0, Reactor Trip or Safety Injection.

ODI-CO-028 (p4; Rev 7/i /08)

Technical Reference(s): 0-ADM-2i 1 (p12; Rev 9/20/il) (Attach if not previously provided) 3-EOP-E-0 (p16; Rev 8/7/12)

Proposed References to be provided to applicants during examination: None

6902052 Objectives 3 and 4 Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam: NA

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

The KA is matched because the operator must demonstrate knowledge of when to conduct crew briefs when implementing off-normal and emergency operating procedures. The goal of the crew brief is to ensure the crew is aware of plant status and the mitigation strategy for the event.

The question is at Memory (1 F) cognitive level because the operator must recall bits of information, and then apply this information to a set of plant conditions to answer the question correctly.

Examination Outline Cross-reference: Level RO SRO Tier# 3 Group# 2 K/A# G2 2.2.22 Importance Rating 4.0 Equipment Control: Knowledge of limiting conditions for operations and safety limits.

Proposed Question: RO Question # 68 During hydrostatic testing of the RCS in Mode 5, RCS pressure is increased to a point exceeding the RCS Pressure Safety Limit.

Which ONE of the following states the RCS Pressure Safety limit setpoint, and the MAXIMUM time allowed in accordance with Technical Specifications to reduce RCS pressure below the safety limit?

A. 2735 PSIG; 5 minutes B. 2735 PSIG; 60 minutes C. 2750 PSIG; 5 minutes D. 2750 PSIG; 60 minutes Proposed Answer: A Explanation (Optional):

A. Correct B. Incorrect. Modes 1-3 time limit is 60 minutes C. Incorrect. PSIA instead of PSIG D. Incorrect. Wrong units and time TS 2.1.2 (p2-i, Amendment Technical Reference(s): 247 and 243) (Attach if not previously provided)

Proposed References to be provided to applicants during examination: N Learning Objective: LP 6902517 Obj 6, 7, 8 (As available)

Question Source: Bank# WTSI 97195 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: 2007 Callaway Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 2 Facility operating limitations in the technical specifications and their bases.

Comments:

This item matches the KA because it evaluates knowledge of the RCS pressure safety limit.

Item was written at the memory cognitive level because the applicant must know the TS safety limit and actions required if it is exceeded

Examination Outline Cross-reference: Level RO SRO Tier# 3 Group# 2 KIA# G2 2.2.25 Importance Rating 3.2 Equipment Control: Knowledge of bases in technical specifications for limiting conditions for operations and safety limits.

Proposed Question: RO Question # 69 Plant conditions:

Unit 3 is at 100% power.

An RCS leak has developed.

RCS pressure is degrading to a reactor trip setpoint.

Which ONE of the following describes reactor trip instrumentation designed to protect against a small break LOCA (SBLOCA) and the core power distribution limit it protects?

A. CT Delta T; DNBR protection B. CT Delta T; Enthalpy Rise Hot Channel Factor protection C. OP Delta T; DNBR protection D. OP Delta T; Enthalpy Rise Hot Channel Factor protection Proposed Answer: A Explanation (Optional):

A. Correct. OTDT provides DNBR protection, as it has an input from RCS pressure.

As pressure lowers, DNBR lowers. OTDT setpoint will lower as pressure lowers.

B. Incorrect because Hot channel factors are protected by control rod positions, reactor power rate of change limitations, and total core power limitations.

C. ncorrect because OPDT protects against core power limitations, it does not have an RCS pressure input. Protection for DNB is correct, however.

D. Incorrect because OPDT does not provide protection in this event. Plausible because OPDT trips do provide protection against exceeding maximum hot channel factors UFSAR (p. 7.2-1; rev 16 10/99)

Technical Reference(s): SD-063 (p. 44-47; rev 9/10/11) (Attach if not previously provided)

Proposed References to be provided to applicants during examination: N Learning Objective: (As available)

Question Source: Bank # WTSI 97975 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: 2007 Harris Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 2 Facility operating limitations in the technical specifications and their bases.

Comments:

Matches KA because item evaluates knowledge of the TS basis for a limiting safety system setting required by technical specifications Item was developed at a comprehension cognitive level because the applicant must determine that PZR pressure is input to one parameter protection channel and not the other, then determine basis from the channel selected

Examination Outline Cross-reference: Level RO SRO Tier# 3 Group# 3 K!A# G3 2.3.13 Importance Rating 3.4 Radiation Control: Knowledge of Radiological Safety Procedures pertaining to licensed operator duties, such as response to radiation monitor alarms, containment entry requirements, fuel handling responsibilities, access to locked high radiation areas, aligning filters, etc.

Proposed Question: RO Question # 70 There is a full core flux map in progress. A containment entry to the 58 of Containment is required to backseat a MS Steam Flow Detector Isolation valve due to packing leakage.

Which ONE of the following describes the conditions necessary to enter containment in accordance with 0-ADM-009, CONTAINMENT ENTRIES WHEN CONTAINMENT INTEGRITY IS ESTABLISHED?

A. Containment Entry is allowed after flux mapping is complete, the In-Core detectors are fully inserted into their storage location and RP has signed onto the ECO.

B. Containment will be posted as a locked high rad area when core flux mapping is in progress, entry into containment is allowed.

C. The In-Core detector area in containment is posted as a locked high radiation area during the flux mapping and entry into containment is allowed.

D. No additional precautions are necessary to access the 58 of Containment during a core flux map.

Proposed Answer: A Explanation (Optional):

A. Correct. See reference B. Incorrect. Containment is not posted as a locked high radiation area during flux mapping. Access is not allowed when flux mapping is in progress.

C. Incorrect. In-Core Detector area is not posted during flux mapping and entry will not be allowed until flux mapping is complete

D. Incorrect. When flux mapping is in progress, containment entry is not allowed in accordance with reference.

0-ADM-009 (p. 20 Rev 13)

Technical Reference(s): .

(Attach if not previously provided)

Proposed References to be provided to applicants during examination: None N/A Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam: NA Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 12 55.43 Radiological safety principles and procedures.

Comments:

KA is matched because the item tests applicants knowledge of containment entry requirements as they pertain to activities causing high radiation environment.

Item developed at memory cognitive level because the applicant must recall procedural precaution for the conditions presented

Examination Outline Cross-reference: Level RO SRO Tier# 3 Group# 3 KJA# G3 2.3.14 Importance Rating 3.4 Radiation Control: Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or emergency conditions or activities.

Proposed Question: RO Question # 71 Plant conditions:

A Steam Generator Tube Rupture has occurred on SG 3B.

The crew has performed all actions of 3-EOP-E-3, Steam Generator Tube Rupture, up to the step to commence depressurization of the RCS.

All equipment is functioning as designed.

Which ONE of the following describes the status of 3B SG Steam Dump to Atmospheric Valve, and the reason for the status?

A. CLOSED with controller in Manual; prevent radioactive release to atmosphere B. CLOSED with controller in Manual; ensures minimum RCS subcooling will be maintained when RCS depressurization is initiated C. Set at 1060 psig with controller in AUTO; prevent uncontrolled radioactive release due to SG safety valve lifting D. Set at 1060 psig with controller in AUTO; ensures minimum RCS subcooling will be maintained when RCS depressurization is initiated Proposed Answer: C Explanation (Optional):

A. Incorrect. Plausible because it is logical to maintain valve closed but controller will not be in manual. Reason is correct B. Incorrect. Same reason as option A, and additionally, reason is plausible because if the ARV stuck open on a ruptured SG, the depressurization would also cause depressurization of the RCS. This would result in loss of RCS subcooling

C. Correct. Controller is in AUTO which allows valve to open as required if pressure rises.

This prevents safety valves from lifting and potentially becoming stuck open, causing radioactive release D. Incorrect. Correct for status of valve, but reason is incorrect. Plausible because valve would be placed in manual and closed if it stuck open below 1050 psig, but this is not the reason that the valve is placed in AUTO. The remainder of the steps for SG isolation are correct for this reason 3-EOP-E-3( p9. Rev 4)

Technical Reference(s): .

(Attach if not previously provided)

BD-EOP-E-3 (p.18; Rev 3)

Proposed References to be provided to applicants during examination: N Learning Objective: (As available)

Question Source: Bank # WTSI 98894 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: 2010 Ginna Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

KA is matched because it evaluates applicants knowledge of a situation where radiation is a concern and is directly affected by operator actions Question is written at memory cognitive level because it evaluates the applicants knowledge of how a procedure is applied for a specific situation/event

Examination Outline Cross-reference: Level RO SRO Tier# 3 Group# 3 KIA# G3 2.3.7 Importance Rating 3.5 Radiation Control: Ability to comply with radiation work permit requirements during normal or abnormal conditions.

Proposed Question: RO Question # 72 Plant conditions:

Unit 3 is in Mode 3.

A containment entry is planned.

You are to perform seal table work.

Radiation level in the area is 250 mr/hr.

The working area contamination level is 12,000 dpmIlOO cm .

2

No waivers are allowed by the HP Supervisor.

Which ONE of the following correctly completes the statements below?

The Containment entry will be made using a (1) . If there are no direct reading dosimeters available, the individuals performing the work must (2)

A. (1) Job Specific RWP (2) wear a respirator B. (1) Job Specific RWP (2) be accompanied by an RP Technician C. (1) General RWP (2) wear a respirator D. (1) General RWP (2) be accompanied by an RP Technician Proposed Answer: B Explanation (Optional):

A. Incorrect. 1st part correct, 2nd part wrong. This is incorrect because the entry requirements for an HRA are not met by wearing a respirator. This is plausible because

the contamination levels are high, and the operator may incorrectly believe this will require the use of a respirator.

B. Correct. 1 st part correct, 2nd part correct. According to 0-ADM-600 Step 4.6.l.B, General RWPs should be used for routine work which does NOT involve work in high dose rate areas, airborne radioactivity areas (except when posting is due to noble gases), areas contaminated in excess of 10,000 dpm!100 cm 2 or jobs involving complex radiological conditions. Since the area is in a High Radiation Area (>100 mrem/hour) and an area where contamination levels are in excess of 10,000 dpm/100

, a Job-Specific RWP is required. This is confirmed by Step 4.6.1.0 which states 2

cm Job specific RWPs should be issued for a work activity outside the scope of a general RWP where specific radiological concerns need to be addressed. Secondly, according to 0-ADM-600 Step 4.9.2.A.3, Any individual or group of individuals permitted to enter a HRA shall be provided with or accompanied by one or more of the following: (1) A radiation monitoring device (i.e. Survey Meter per preceding NOTE) which continuously indicates the radiation dose rate in the area, (2) A radiation monitoring device which continuously integrates the radiation dose rate in the area and alarms when a preset integrated dose is received, (i.e. a direct reading dosimeter per the preceding NOTE, which none of are available) or (3) An individual qualified in Radiation Protection procedures with a radiation dose rate monitoring device, who is responsible for providing positive control over the activities within the area and shall perform periodic radiation surveillance at the frequency specified by RP supervision on the RWP.

C. Incorrect. 1st part wrong, 2nd part wrong. See A and B.

D. Correct. jst part wrong, 2nd part correct. This is incorrect because the entry cannot be made using a General RWP. This is plausible because it is identified as routine work, and the operator may be unaware of the requirements for using a Job-Specific RWP.

0-AD M-600 (p32 & 39 Rev 4)

Technical Reference(s): .

(Attach if not previously provided)

Proposed References to be provided to applicants during examination: None 6902970 Objective 2 Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X

Question History: Last NRC Exam: NA Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 11 55.43 Purpose and operation of radiation monitoring systems, including alarms and survey equipment.

Comments:

The KA is matched because the operator must demonstrate the Ability to comply with radiation work permit requirements during normal or abnormal conditions by identifying the type of RWP that is needed, given a set of radiological conditions, and identifying what additional requirement must be met to enter an HRA, under a specific set of conditions.

The question is at the Comprehension/Analysis (2Rl!3PEO) cognitive level because the operator must recall bits of information (Requirements for General RWP Use, Job-Specific Use, and HRA entry requirements), and then relate this to itself through the identification of procedural requirements, and! or apply this information to a set of plant conditions to answer the question correctly.

Examination Outline Cross-reference: Level RO SRO Tier# 3 Group# 4 K!A # G4 2.4.20 Importance Rating 3.8 Emergency Procedures! Plan: Knowledge of operational implications of LOP warnings, cautions, and notes.

Proposed Question: RO Question # 73 Plant conditions:

The crew is performing a cooldown in 3-EOP-ES-1 .2, Post LOCA Cooldown and Depressurization.

RCS pressure is 1500 psig and stable.

RCS subcooling is 25°F and increasing slowly.

Containment pressure is 2.5 psig and rising slowly.

MOV-3-749A CCW TO RHR HX A could not be opened.

Both RHR pumps have been stopped.

Subsequently, the LOCA increases in size RCS pressure equalizes with containment pressure, which reaches 27 psig.

Which ONE of the following describes the required operation of 3A RHR pump?

The 3A RHR pump....

A. will be manually restarted to provide low-head SI flow.

B. will not be restarted since cooling is not available to RHR Heat Exchanger 3A.

C. will auto start when the SI signal is received due to the High Containment Pressure.

D. will not be restarted because it is not needed under the present plant conditions if the B RHR Pump can be started.

Proposed Answer: A Explanation (Optional):

A. Correct. According to 3-EOP-E-i, a Caution prior to Step 14 states that High-Head SI flow and RCS Subcooling are required to be monitored. If either High-Head SI flow increases or RCS Subcooling decreases in an uncontrolled manner, the RHR pumps must be manually restarted to supply water to the RCS. In Step 14, the crew would have stopped the HHSI Pumps, and then in Step 20, the crew would have made the transition to the current procedure (i.e. ES-i .2). According to BD-EOP-E-i, the basis of this Caution is to alert the operator that if RCS pressure should decrease in an uncontrolled manner (as indicated by an increase in HHSI flow or a reduction in Subcooling), to less than the shutoff head of the RHR pumps, they must be manually restarted since the SI signal has been reset.

Except for relatively large LOCAs, the RCS pressure should remain greater than the shutoff head pressure of the RHR pumps until later in the recovery following a controlled cooldown and depressurization. To avoid damage to the RHR pumps, instructions are provided to stop these pumps early in the recovery if RCS pressure is greater than their shutoff head. However, if RCS pressure subsequently decreases in an uncontrolled manner to less than the RHR pump shutoff head, then the pumps will have to be restarted manually since no automatic signal is available. This is a generic caution that applies throughout the EOP network.

B. Incorrect. This is incorrect because a Caution prior to step 14 in 3-EOP-E-i indicates that they should be started. This is plausible because the operator may incorrectly believe that they do not have cooling, and they need it. The plant is operating in the injection Mode of Safety Injection in 3-EOP-ES-1 .2. The RHR pumps were secured either on the foldout page criteria or stopped in 3-EOP-E-i, Loss of Reactor or Secondary Coolant, step 14. The 3A RHR pump is taking suction from the RWST which is approximately 100°F and CCW is isolated to both RHR Heat Exchangers. Only when the crew transitions to 3-EOP-ES-1 .3, Transfer to Cold Leg Recirculation are their requirements to start only the RHR pumps with CCW available to the RHR Heat Exchangers.

C. Incorrect. This is incorrect because these pumps will NOT start automatically. This is plausible because the operator may incorrectly believe that they will. Even though the SI signal reset has occurred in 3-EOP-E-0, Attachment 3 or in 3-EOP-E-1, step 6. Any additional SI signal is prevented from actuating the sequencer because the original SI signal has not cleared (only Reset).

D. Incorrect. This is incorrect because a Caution prior to step 14 in 3-EOP-E-1 indicates that both RHR Pumps should be started. This is plausible because the operator may incorrectly believe that only one RHR pump needs to be started under the current conditions.

SD-063 (p54-57; rev 9/10/il)

Technical Reference(s): 561 0-T-Ll (sheets 11 and I 2A) (Attach if not previously provided)

Proposed References to be provided to applicants during examination: N

Learning Objective: 6902163 Objectives 2, 8 (As available)

Question Source: Bank # WTSI 70240 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: 2009 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

The KA is matched because the operator must demonstrate the knowledge/ability to apply the Caution in the EOP when the RHR pumps must be restarted after they are stopped.

The question is at the Comprehension/Analysis (2R1) cognitive level because the operator must recall bits of information (When in E-1 Series of procedures, after RHR Pumps have been stopped, what conditions would require that they be manually restarted), and then apply this information to a set of plant conditions to answer the question correctly.

Examination Outline Cross-reference: Level RO SRO Tier# 3 Group# 4 KIA# G4 2.4.41 Importance Rating 2.9 Emergency Procedures I Plan: Knowledge of the emergency action level thresholds and classifications.

Proposed Question: RO Question # 74 An Alert was declared.

In accordance with the Turkey Point Emergency Plan, which one of the following is the emergency action level threshold that was met for this event?

Events are in process or have occurred which...

A. involve actual or likely major failures of plant functions needed for protection of the public B. involve actual or imminent substantial core degradation or melting with potential for loss of containment integrity C. involve an actual or potential substantial degradation of the level of safety of the plant D. indicate a potential degradation of the level of safety of the plant Proposed Answer: C Explanation (Optional):

A. Incorrect; this is the definition for site are emergency. Plausible; defines one of the EAL classifications B. Incorrect; this is the definition for general emergency. Plausible; defines one of the EAL classifications C. Correct per above reference D. Incorrect; this is the definition for a UE. Plausible; defines one of the EAL classifications

PTN Radiological Emergency Plan- PTN EP sections 3.1 Technical Reference(s): - .

(Attach if not previously provided) 3.4 REV-55 Proposed References to be provided to applicants during examination: N Learning Objective: (As available)

Question Source: Bank #

Modified Bank # WTSI 99803 (Note changes or attach parent)

New Question History: Last NRC Exam: 2009 Turkey Point Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

Question matches KA because it evaluates applicants knowledge of EAL thresholds.

Question is written at memory cognitive level because it asks for direct knowledge of the EAL threshold by definition

Examination Outline Cross-reference: Level RO SRO Tier# 3 Group# 4 KIA# G4 2.4.42 Importance Rating 2.6 Emergency Procedures! Plan: Knowledge of emergency response facilities.

Proposed Question: RO Question # 75 Plant conditions:

An emergency event is in progress which requires activation of the Emergency Response Organization.

A Site Area Emergency has been declared.

o Emergency Response Facilities have been declared operational.

Which ONE of the following identifies the location that you will direct the Off-Shift SNPOs to report?

A. Control Room B. Operations Support Center C. Technical Support Center D. Site Assembly Area for evacuation Proposed Answer: B Explanation (Optional):

A. Incorrect. Plausible because it is logical that Operations personnel would report to the control room.

B. Correct. See reference C. Incorrect. SNPOs do not report to TSC but plausible because they will provide technical support when E-Plan is activated.

D. Incorrect. Site Assembly Area is for non-essential personnel, which would not include qualified SNPOs

Turkey Point Plant Radiological Emergency Plan section 2.4.4 Technical Reference(s): page 2-30 (Rev 55) (Attach if not previously provided)

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

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam: NA Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

The KA is matched because the operator must demonstrate knowledge of emergency response facilities including the OSC.

The question is at the Memory (1 P) cognitive level because the operator must recall information about where to assemble

ML13156A030

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