Text

2012-10-Draft Written Exam
	ML13127A260
Person / Time
Site:	Waterford
Issue date:	10/01/2012
From:	Vincent Gaddy; Operations Branch IV
To:	; Entergy Nuclear Operations
	LAURA HURLEY
References
	50-382/OL-12
	Download: ML13127A260 (202)
	v • d • e

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # CE/E02 EK1.3 Importance Rating 3.0 K/A Statement Knowledge of the operational implications of the following concepts as they apply to the (Reactor Trip Recovery): Annunciators and conditions indicating signals, and remedial actions associated with the (Reactor Trip Recovery).

Proposed Question: RO 1 Rev: 0 Given the following:

Feedwater Pump A is running with FWCS 1 and FWCS 2 in RTO with both SG levels

> 27.4 % NR and slowly rising

Reactor Coolant Pumps 1A, 1B, and 2A were secured due to Pressurizer Spray issues

OP-902-001, Reactor Trip Recovery Procedure is being implemented S/U XFMR B 86 Trip/Trouble Alarm comes in and the 86STB lockout is actuated. The operational implication is that the ______(1)______ Heat Removal Safety Function will not be satisfied and implementation of __________(2)__________ Recovery procedure will be necessary.

(1) (2)

A. Core OP-902-006, Loss of Main Feedwater B. Core OP-902-003, Loss of Offsite Power/

Loss of Forced Flow C. RCS OP-902-006, Loss of Main Feedwater D. RCS OP-902-003, Loss of Offsite Power/

Loss of Forced Flow Revision 0 Facility: Waterford 3 Page 1 of 202

2012 NRC Exam RO Written Exam Worksheet Proposed Answer: B Explanation: (Optional)

A. Incorrect. The Safety Function is correct. The procedure required is incorrect.

Feedwater pump A support equipment is available based on loss of power only affecting the B train and the Feedwater pump would remain in service.

B. CORRECT: The Safety Function is correct. The only running RCP (RCP 2B) loses power in this scenario and natural circulation does not support a loop delta t of < 13 degrees as required by OP-902-001, Reactor Trip Recovery Procedure. The procedure is correct. With no RCPs running, the procedure that covers this is OP-902-003, Loss of Offsite Power/Loss of Forced Flow Recovery Procedure.

C. Incorrect. The Safety Function is incorrect. The procedure required is incorrect. The only running RCP (RCP 2B) loses power in this scenario and natural circulation does not support a loop delta t of < 13 degrees as required by OP-902-001, Reactor Trip Recovery Procedure. The RCS heat removal safety function is met.

D. Incorrect. The Safety Function is incorrect. The procedure required is correct.

OP-902-001, Reactor Trip Recovery Procedure Technical Reference(s): Rev.012 (Attach if not previously provided) OP-902-009, Standard Appendices, Rev. 306 (including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PPE01, Obj. 16 (As available)

WLP-OPS-PPE05, Obj. 5 Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 5,10 55.43 Comments:

Revision 0 Facility: Waterford 3 Page 2 of 202

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 000008 AA2.30 Importance Rating 4.3 K/A Statement Ability to determine and interpret the following as they apply to the Pressurizer Vapor Space Accident: Inadequate core cooling.

Proposed Question: RO 2 Rev: 0 Given:

The plant has experienced an inadvertent opening of a Pressurizer relief valve.

The crew has entered OP-902-002, Loss of Coolant Accident Recovery Procedure.

One hour into the event, the following indications exist:

Pressurizer level is 100% and stable Pressurizer pressure is 1450 PSIA and stable Representative CET temperature is 595°F and slowly rising Reactor Vessel Plenum level indicates all levels voided High Pressure Safety Injection (HPSI) Pump B tripped on startup HPSI Cold Leg flows are 0 gpm Core cooling is ____(1)____ because ________(2)________.

(1) (2)

A. adequate pressurizer level is above HPSI throttle criteria B. adequate HPSI flow curve is met C. inadequate CET temperatures indicate superheat conditions D. inadequate only one HPSI pump is running Revision 0 Facility: Waterford 3 Page 3 of 202

2012 NRC Exam RO Written Exam Worksheet Proposed Answer: C Explanation: (Optional)

A. Incorrect. Core cooling is inadequate because CET temperatures are indicating superheat using the steam tables. Pressurizer level above HPSI throttle criteria is not part of the acceptance criteria for core heat removal.

B. Incorrect. Core cooling is inadequate because CET temperatures are indicating superheat using the steam tables. HPSI flow curves are met at 0 gpm per the table at 1450 PSIA.

C. CORRECT: Core cooling is inadequate because CET temperatures are indicating superheat using the steam tables. Some core uncovery and superheat conditions can exist within 30 minutes of the event, but the initial conditions state we are one hour into the event.

D. Incorrect. One HPSI Pump being available is not the reason for the low flow condition High pressure in the RCS is. A cooldown and depressurization of the RCS would correct this situation.

Technical Reference(s): Steam tables (Attach if not previously provided) OP-902-009 revision 307 (including version/revision number) OP-902-002 revision 14 Proposed references to be provided to applicants during examination: Steam Tables Learning Objective: WLP-OPS-PPE02 Obj. 8 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 6,8 55.43 Comments:

Revision 0 Facility: Waterford 3 Page 4 of 202

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 000009 G2.1.19 Importance Rating 3.9 K/A Statement Ability to use plant computers to evaluate system or component status. (Small Break LOCA)

Proposed Question: RO 3 Rev: 0 Given:

A Small Break LOCA occurred one hour ago

Pressurizer level is 0%

The RCS is in a saturated condition Reactor Vessel Level can be monitored using QSPDS ____(1)____. If Safety Injection flow requirements are met, the minimum reactor vessel plenum level to remain in the optimum recovery procedure is ___(2)___%.

(1) (2)

A. 1 only 20 B. 1 only 80 C. 1 or 2 20 D. 1 or 2 80 Revision 0 Facility: Waterford 3 Page 5 of 202

2012 NRC Exam RO Written Exam Worksheet Proposed Answer: C Explanation: (Optional)

A. Incorrect. Wrong monitoring capability. Correct plenum level requirement.

B. Incorrect. Wrong monitoring capability. Wrong plenum level requirement.

C. CORRECT: Vessel level can be monitored using both QSPDS 1 and 2. The Operability of the indication is verified using OP-903-013, Monthly Channel Checks.

OP-902-002, requires > 20% plenum level when coupled solely with Safety Injection flow requirements. The higher requirement is coupled with having PZR level and operating in a subcooled condition. It would be expected that the higher plenum level could be maintained with those conditions, and thus the tighter restriction. 0%

PZR level was given in the stem.

D. Incorrect. Wrong plenum level requirement.

Technical Reference(s): OP-902-002 Rev. 14 (Attach if not previously provided) OP-903-013 Rev. 16 (including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OP-PPE02 Obj. 16 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7,10 55.43 Comments:

Revision 0 Facility: Waterford 3 Page 6 of 202

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 000015/17 AK1.01 Importance Rating 4.4 K/A Statement Knowledge of the operational implications of the following concepts as they apply to Reactor Coolant Pump Malfunctions (Loss of RC Flow): Natural circulation in a nuclear reactor power plant.

Proposed Question: RO 4 Rev: 0 Given:

The plant is at 100% power.

A loss of offsite power has occurred.

The crew has entered OP-902-003, Loss of Offsite Power/Loss of Forced Circulation Recovery Procedure.

Verification of an RCS temperature response to a steam flow or feed flow change during these conditions cannot be accomplished until approximately ____(1)____ minutes following the action. RCS subcooling will be verified using __(2)__ subcooling monitors.

(1) (2)

A. 20 to 30 CET B. 5 to 15 CET C. 20 to 30 Thot D. 5 to 15 Thot Revision 0 Facility: Waterford 3 Page 7 of 202

2012 NRC Exam RO Written Exam Worksheet Proposed Answer: B Explanation: (Optional)

A. Incorrect. Per OI-038-000, Verification of an RCS temperature response to a plant change during natural circulation cannot be accomplished until approximately 5-15 minutes following the action.

B. CORRECT: Per OI-038-000, Verification of an RCS temperature response to a plant change during natural circulation cannot be accomplished until approximately 5-15 minutes following the action. Guidance also provided in OI-038-000 that while on natural circulation, CETs will be used to verify subcooling.

C. Incorrect. Per OI-038-000, Verification of an RCS temperature response to a plant change during natural circulation cannot be accomplished until approximately 5-15 minutes following the action. Guidance also provided in OI-038-000 that while on natural circulation, CETs will be used to verify subcooling.

D. Incorrect. Guidance provided in OI-038-000 that while on natural circulation, CETs will be used to verify subcooling.

Technical Reference(s): OI-038-000 revision 5 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PPE05 obj. 3 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 5, 8 55.43 Comments:

Revision 0 Facility: Waterford 3 Page 8 of 202

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 000022 G2.4.1 Importance Rating 4.6 K/A Statement Knowledge of EOP entry conditions and immediate action steps. (Loss of Rx Coolant Makeup)

Proposed Question: RO 5 Rev: 0 Given:

The plant is at 100% power.

A loss of all Charging Pumps has occurred due to gas intrusion.

The crew has entered OP-901-112, Charging or Letdown Malfunction.

The ATC reports that Pressurizer level is 45% and slowly dropping.

Tave and Tref are matched.

Due to the stated conditions, the crew will ____(1)____. Per OP-901-112, a

______(2)______ will be aligned to restore Pressurizer level.

(1) (2)

A. commence a rapid Low Pressure Safety Injection Pump plant downpower B. trip the reactor High Pressure Safety Injection Pump C. commence a rapid High Pressure Safety Injection Pump plant downpower D. trip the reactor Low Pressure Safety Injection Pump Revision 0 Facility: Waterford 3 Page 9 of 202

2012 NRC Exam RO Written Exam Worksheet Proposed Answer: B Explanation: (Optional)

A. Incorrect. OP-901-112 directs a reactor trip (not a rapid plant downpower) if minimum pressurizer level is not met. After the reactor is tripped OP-901-112 directs the crew to depressurize the RCS and start a HPSI pump to restore pressurizer level.

B. CORRECT: The minimum pressurizer level at 100% power is 46.7%. The given level is 45%. OP-901-112 directs a reactor trip if minimum pressurizer level is not met. After the reactor is tripped OP-901-112 directs the crew to depressurize the RCS and start a HPSI pump to restore pressurizer level.

C. Incorrect. OP-901-112 directs a reactor trip (not a rapid plant downpower) if minimum pressurizer level is not met.

D. Incorrect. After the reactor is tripped OP-901-112 directs the crew to depressurize the RCS and start a HPSI pump to restore pressurizer level.

Technical Reference(s): OP-901-112 revision 4 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: OP-901-112 Attachment 1 Learning Objective: WLP-OPS-PP010 obj. 3 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7, 10 55.43 Comments:

Revision 0 Facility: Waterford 3 Page 10 of 202

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 000025 AA1.03 Importance Rating 3.4 K/A Statement Ability to operate and / or monitor the following as they apply to the Loss of Residual Heat Removal System: LPI pumps Proposed Question: RO 6 Rev: 0 Given:

The plant is in Mode 5

The RCS was drained to 14.5 feet MSL.

Low Pressure Safety Injection (LPSI) pumps were secured due to RCS leakage

High Pressure Safety Injection (HPSI) Pump B was started in accordance with OP-901-131, Shutdown Cooling Malfunction

The RCS level has been raised and is being maintained at 16 feet To restore Shutdown Cooling per OP-901-131, the crew should vent and start LPSI Pump ____(1)_____ because HPSI Pump B is injecting to hot leg _______(2)______.

(1) (2)

A. A 1 B. A 2 C. B 1 D. B 2 Revision 0 Facility: Waterford 3 Page 11 of 202

2012 NRC Exam RO Written Exam Worksheet Proposed Answer: B Explanation: (Optional)

A. Incorrect. HPSI Pump B injects to Hot leg 2 (not Hot leg 1).

B. CORRECT: OP-901-131, section E1, step 5 directs the crew to vent the suction piping of the LPSI Pump that will take suction on the Hot leg with the operating HPSI Pump. In this case, HPSI Pump B was started to fill the RCS which injects to Hot leg

2. The crew will start LPSI Pump A since its suction is from Hot leg 2.

C. Incorrect. HPSI Pump B was started to fill the RCS which injects to Hot leg 2. The crew will start LPSI Pump A since its suction is from Hot leg 2.

D. Incorrect. HPSI Pump B was started to fill the RCS which injects to Hot leg 2. The crew will start LPSI Pump A since its suction is from Hot leg 2. HPSI Pump B injects to Hot leg 2 (not Hot leg 1).

Technical Reference(s): OP-901-131 rev. 302 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-REQ21 obj. 2 (As available)

Question Source: Bank # X 08281 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 3,10 55.43 Comments:

Revision 0 Facility: Waterford 3 Page 12 of 202

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 000026 AK3.02 Importance Rating 3.6 K/A Statement Knowledge of the reasons for the following responses as they apply to the Loss of Component Cooling Water: The automatic actions (alignments) within the CCWS resulting from the actuation of the ESFAS Proposed Question: RO 7 Rev: 0 Given the following:

Plant has tripped from 100% power.

RCS pressure is 1680 PSIA and slowly lowering.

Containment pressure is 16.8 PSIA and slowly rising.

CCW Trains A and B were in a normal Mode 1 alignment prior to the trip.

Based upon the current conditions, which describes the configuration of the CCW System and the reason for this configuration?

A. CC-200 A and 727, CCW A Header Isolation valves, received a CLOSE signal.

CCW Train B is supplying cooling to the Reactor Coolant Pumps with these conditions.

B. CC-200 A and 727, CCW A Header Isolation valves, received a CLOSE signal.

CCW Train B is supplying cooling to the Fuel Pool Heat Exchanger with these conditions.

C. CC-200 A and 727, CCW A Header Isolation valves, remain OPEN. CCW Train A provides cooling to the Reactor Coolant Pumps with these conditions.

D. CC-200 A and 727, CCW A Header Isolation valves, remain OPEN. CCW Train A provides cooling to the Fuel Pool Heat Exchanger with these conditions.

Revision 0 Facility: Waterford 3 Page 13 of 202

2012 NRC Exam RO Written Exam Worksheet Proposed Answer: B Explanation: (Optional)

A. Incorrect. CC-200A and CC-727 remain open, these two valves close on a CSAS. In this condition only a SIAS is present. Therefore the CCW AB loop is being supplied from the A train of CCW (not the B train).

B. Incorrect. CC-200A and CC-727 remain open, these two valves close on a CSAS.

But, the FPHX TCV closeed due to a SIAS C. CORRECT: CC-200A and CC-727 remain open, these two valves close on a CSAS.

In this condition only a SIAS is present. Therefore the CCW AB loop is being supplied from the A train of CCW D. Incorrect. CC-200A and CC-727 remain open, these two valves close on a CSAS.

But, the FPHX TCV closed due to a SIAS Technical Reference(s): OP-902-009 revision 307 appendix 4.

(Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-CC00 obj. 5 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Comments:

Revision 0 Facility: Waterford 3 Page 14 of 202

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 000027 AK2.03 Importance Rating 2.6 K/A Statement Knowledge of the interrelations between the Pressurizer Pressure Control Malfunctions and the following: Controllers and positioners Proposed Question: RO 8 Rev: 0 Given:

The plant is at 100% power.

The AUTOMATIC function of Pressurizer Pressure controller (RC-IPIC-0100) fails, causing RCS pressure to rise.

The crew has entered OP-901-120, Pressurizer Pressure Control Malfunction.

The Pressurizer Pressure controller (RC-IPIC-0100) has been taken to manual.

If no further actions were taken prior to Pressurizer Pressure reaching 2275 psia, the pressurizer spray valves would ____(1)____. To lower RCS pressure, the ATC operator must _____(2)_____ the output of the pressurizer pressure controller.

(1) (2)

A. automatically open raise B. automatically open lower C. require manual opening raise D. require manual opening lower Revision 0 Facility: Waterford 3 Page 15 of 202

2012 NRC Exam RO Written Exam Worksheet Proposed Answer: C Explanation: (Optional)

A. Incorrect. The spray valve controller process is fed by the pressurizer pressure controller output (which is failed low). Therefore, the spray valve controller will not auto open and will require manual operation.

B. Incorrect. The spray valve controller process is fed by the pressurizer pressure controller output (which is failed low). Therefore, the spray valve controller will not auto open and will require manual operation. The output of the pressurizer pressure controller must be raised to lower RCS pressure.

C. CORRECT: The pressurizer pressure controller failing low will cause pressurizer pressure to rise. The spray valve controller process is fed by the pressurizer pressure controller output (which is failed low). Therefore, the spray valve controller will not auto open and will require manual operation. The output of the pressurizer pressure controller must be raised to lower RCS pressure.

D. Incorrect. The output of the pressurizer pressure controller must be raised to lower RCS pressure.

Technical Reference(s): OP-901-120 revision 301 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PP010 obj. 3 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 5, 7 55.43 Comments:

Revision 0 Facility: Waterford 3 Page 16 of 202

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 000029 EA1.12 Importance Rating 4.1 K/A Statement Ability to operate and monitor the following as they apply to a ATWS: M/G set power supply and reactor trip breakers Proposed Question: RO 9 Rev: 0 Given:

The plant was at 100% power Main Feedwater Isolation Valve 1 fails closed The CRS directs the BOP to trip the reactor The reactor trip pushbuttons fail to trip the reactor at CP-2 The ATC operator should manually initiate Diverse Reactor Trip which opens the

____(1)_____. If this method for tripping the reactor also fails, the ATC operator should open the _______(2)______ feeder breakers.

(1) (2)

A. reactor trip breakers 32A and 32B B. MG set load contactors 32A and 32B C. MG set load contactors 31A and 31B D. reactor trip breakers 31A and 31B Revision 0 Facility: Waterford 3 Page 17 of 202

2012 NRC Exam RO Written Exam Worksheet Proposed Answer: B Explanation: (Optional)

A. Incorrect. Diverse Reactor Trip (DRTS) pushbuttons trip the reactor by opening MG set output breakers (Reactor Trip breakers remain closed).

B. CORRECT: Diverse Reactor Trip (DRTS) pushbuttons trip the reactor by opening MG set output breakers (Reactor Trip breakers remain closed). If DRTS fails to trip the Reactor, OP-902-000 directs the operator to remove power to the 32 busses by opening the feeder breaker to them.

C. Incorrect. If DRTS fails to trip the Reactor, OP-902-000 directs the operator to remove power to the 32 busses by opening the feeder breaker to them.

D. Incorrect. Diverse Reactor Trip (DRTS) pushbuttons trip the reactor by opening MG set output breakers (Reactor Trip breakers remain closed). If DRTS fails to trip the Reactor, OP-902-000 directs the operator to remove power to the 32 busses by opening the feeder breaker to them.

Technical Reference(s): OP-902-000 revision 13 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PPE01 obj. 9 (As available)

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

New Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 6,10 55.43 Comments:

Revision 0 Facility: Waterford 3 Page 18 of 202

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 000038 G2.2.42 Importance Rating 3.9 K/A Statement Ability to recognize system parameters that are entry-level conditions for Technical Specifications. (Steam Gen. Tube Rupture)

Proposed Question: RO 10 Rev: 0 Given:

A tube rupture event is in progress involving both Steam Generators

Prior to the trip Main Feedwater Reg Valve B, FW-173B failed closed

SG 1 level is 40% NR and rising at 1%/min

SG 2 level is 17% NR and rising at 2%/min

Both Startup Feedwater Reg Valves are ~ 17% open

Chemistry has reported the following activity samples:

o SG 1 Dose Equivalent Iodine - 0.11 ci/gm o SG 2 Dose Equivalent Iodine - 0.20 ci/gm Steam Generator _____(1)_____ exceeded LCO entry conditions per Technical Specification 3.7.1.4, Secondary Activity. After the initial cooldown the crew should isolate Steam Generator ____(2)____.

(1) (2)

A. 2 has 1 B. 1 and 2 have 1 C. 2 has 2 D. 1 and 2 have 2 Revision 0 Facility: Waterford 3 Page 19 of 202

2012 NRC Exam RO Written Exam Worksheet Proposed Answer: D Explanation: (Optional)

A. Incorrect. Wrong TS evaluation. Wrong Steam Generator.

B. Incorrect. Correct TS evaluation. Wrong Steam Generator.

C. Incorrect. Wrong TS evaluation. Correct Steam Generator.

D. CORRECT: TS 3.7.1.4 requires the Secondary activity to be < 0.1 uci/gm DEQ I-131: both SGs are greater than the limit. OP-902-007 requires isolation of the most affected SG. Criteria include highest activity level and one SG rising faster with essentially the same feed rate. These parameters are mentioned in the stem. A malfunction was included in the stem which would make the most affected SG level actually be lower than the least affect SG. If the candidate selects the generator just on level they will pick another one of the distractors.

Technical Reference(s): TS 3.7.1.4 Amendment 0 (Attach if not previously provided) OP-902-007, Rev 13 (including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: OP-902-007 Obj. (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 Comments:

Revision 0 Facility: Waterford 3 Page 20 of 202

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # CE/E05 EA2.2 Importance Rating 3.4 K/A Statement Adherence to appropriate procedures and operation within the limitations in the facility*s license and amendments. (Steam Line Rupture - Excessive Heat Removal)

Proposed Question: RO 11 Rev: 0 Given:

An Excess Steam Demand event has occurred due to an unisolable Main Steam line break on Steam Generator 1.

The crew has entered OP-902-004, Excess Steam Demand Recovery Procedure.

The crew has completed actions for stabilizing RCS temperature.

The ATC will then stabilize RCS pressure ____(1)____ HPSI shutoff head and will adjust RCS pressure such that the differential pressure across the steam generator tubes does not exceed the design differential pressure of _____(2)________.

(1) (2)

A. above 1400 psid B. below 1400 psid C. below 1600 psid D. above 1600 psid Revision 0 Facility: Waterford 3 Page 21 of 202

2012 NRC Exam RO Written Exam Worksheet Proposed Answer: D Explanation: (Optional)

A. Incorrect. The design maximum delta P across the SG tubes is 1600 psid.

B. Incorrect. The operator ensures RCS pressure is above (not below) HPSI shutoff head to minimize the possibility of the RCS becoming water solid due to continued flow from the HPSI pumps. The design maximum delta P across the SG tubes is 1600 psid.

C. Incorrect. The operator ensures RCS pressure is above (not below) HPSI shutoff head to minimize the possibility of the RCS becoming water solid due to continued flow from the HPSI pumps.

D. CORRECT: The operator ensures RCS pressure is above HPSI shutoff head to minimize the possibility of the RCS becoming water solid due to continued flow from the HPSI pumps. The design maximum delta P across the SG tubes is 1600 psid.

Technical Reference(s): OP-902-004 Revision 13 (Attach if not previously provided) TGOP-902-004 (including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PPE04 obj. 4 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Comments:

Revision 0 Facility: Waterford 3 Page 22 of 202

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # CE/E06 EK2.1 Importance Rating 3.3 K/A Statement Knowledge of the interrelations between the (Loss of Feedwater) and the following: Components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Proposed Question: RO 12 Rev: 0 Given:

The reactor has tripped due to a loss of Main Feedwater.

The crew has entered OP-902-006, Loss of Main Feedwater Recovery Procedure.

Condensate Storage Pool (CSP) level is 30% and dropping with CMU-141, CSP LCV Bypass, Open.

When CSP level reaches ____(1)_____ %, the crew should commence alignment of Emergency Feedwater Pump suctions to _______(2)______ train of Auxiliary Component Cooling Water (ACCW).

(1) (2)

A. 11 one operating train of ACCW B. 11 their respective train of ACCW C. 25 one operating train of ACCW D. 25 their respective train of ACCW Revision 0 Facility: Waterford 3 Page 23 of 202

2012 NRC Exam RO Written Exam Worksheet Proposed Answer: C Explanation: (Optional)

A. Incorrect. 11% CSP level is the level at which EFW pump suction transfer should be completed to prevent from cavitating the EFW pump. 25% level in the CSP is when the procedure directs the transfer to take place.

B. Incorrect. 11% CSP level is the level at which EFW pump suction transfer should be completed to prevent from cavitating the EFW pump. 25% level in the CSP is when the procedure directs the transfer to take place. Any operating train of ACCW is aligned to EFW.

C. CORRECT: Step 30 of OP-902-006 directs the crew to transfer EFW pump suction to one operating train of ACCW when CSP level is less than 25%. Any operating train of ACCW is aligned to EFW.

D. Any operating train of ACCW is aligned to EFW.

Technical Reference(s): OP-902-006 revision 13 (Attach if not previously provided) OP-902-009 revision 307 (including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PPE06 obj. 9 (As available)

Question Source: Bank # X 5662-A Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Comments:

Revision 0 Facility: Waterford 3 Page 24 of 202

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 000055 EK3.02 Importance Rating 4.3 K/A Statement Knowledge of the reasons for the following responses as the apply to the Station Blackout: Actions contained in EOP for loss of offsite and onsite power Proposed Question: RO 13 Rev: 0 Given:

Off-site power has been lost

Emergency Diesel Generator A is danger tagged out

Emergency Diesel Generator B trips on overspeed

The crew is performing actions contained in OP-902-005, Station Blackout Recovery Procedure

All CEAs are inserted The CRS has directed the BOP to commence a cooldown using Atmospheric Dump Valves. The cooldown must be stopped before Tcold reaches ____(1)____ to ensure that

_____(2)________.

(1) (2)

A. 400°F Shutdown Margin requirements are met B. 520°F Shutdown Margin requirements are met C. 400°F RCS subcooled margin requirements are not exceeded D. 520°F RCS subcooled margin requirements are not exceeded Revision 0 Facility: Waterford 3 Page 25 of 202

2012 NRC Exam RO Written Exam Worksheet Proposed Answer: A Explanation: (Optional)

A. CORRECT: Shutdown Margin is maintained while above 400°F as long as all CEAs are inserted. Step 19 of OP-902-005 directs stopping the cooldown prior to Tc lowering to less than 400°F.

B. Incorrect. Shutdown Margin is maintained while above 400°F (not 520°F) as long as all CEAs are inserted. 520°F is a trigger temperature in the EOPs for a rapid cooldown.

C. Incorrect. The temperature limit of 400°F is to ensure shutdown margin requirements are met. The limit is not to ensure proper subcooling.

D. Incorrect. Shutdown Margin is maintained while above 400°F (not 520°F) as long as all CEAs are inserted.

Technical Reference(s): OP-902-005 rev. 15 (Attach if not previously provided) TG-OP-902-005 revision 304 (including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PPE05 obj. 7 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Comments:

Revision 0 Facility: Waterford 3 Page 26 of 202

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 000056 AA2.46 Importance Rating 4.2 K/A Statement Ability to determine and interpret the following as they apply to the Loss of Offsite Power: That the ED/Gs have started automatically and that the bus tie breakers are closed Proposed Question: RO 14 Rev: 0 Given:

The plant is at 100% power

A Loss of offsite power occurs The BOP is performing verification of the Maintenance of Vital Auxiliaries and observes the following for EDG B.

EDG B is running.

EDG B Frequency is 59.8 Hertz.

EDG B output breaker is open.

EDG B voltage is 3910 AC volts.

Which of the following should the BOP perform?

A. Raise EDG B voltage and verify EDG B output breaker closes.

B. Manually close EDG B output breaker.

C. Raise EDG B frequency and verify EDG B output breaker closes.

D. Direct the NAO to close EDG B output breaker locally.

Revision 0 Facility: Waterford 3 Page 27 of 202

2012 NRC Exam RO Written Exam Worksheet Proposed Answer: A Explanation: (Optional)

A. CORRECT: Per OP-902-000, the contingency action for an EDG output breaker not closing is to verify EDG voltage 3920-4350 VAC. If the voltage is not in the band, the operator is expected to manually adjust voltage to within the band and the EDG output breaker should then close. (This step is an immediate action)

B. Incorrect. If EDG B output breaker did not close in auto, then the operator should verify voltage in band prior to manually closing the EDG output breaker.

C. Incorrect. EDG frequency can not be adjusted when the EDG is running in Emergency mode.

D. Incorrect. Closing the EDG output breaker locally is an immediate action step but is performed only after EDG voltage is verified to be within band.

Technical Reference(s): OP-902-000 revision 13 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PPE01 obj. 10 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7,10 55.43 Comments:

Revision 0 Facility: Waterford 3 Page 28 of 202

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 000057 AA2.17 Importance Rating 3.1 K/A Statement Ability to determine and interpret the following as they apply to the Loss of Vital AC Instrument Bus: System and component status, using local or remote controls Proposed Question: RO 15 Rev: 0 Given:

The plant is at 100% power.

A loss of Instrument SUPS MA has just occurred.

The ATC will report to the CRS that Reactor Trip Breakers 1, 5, ________(1)_______

are open. Due to these conditions, the reactor will _____(2)_____.

(1) (2)

A. 2 and 6 trip B. 4 and 8 trip C. 2 and 6 remain at 100% power D. 4 and 8 remain at 100% power Revision 0 Facility: Waterford 3 Page 29 of 202

2012 NRC Exam RO Written Exam Worksheet Proposed Answer: C Explanation: (Optional)

A. Incorrect. On a loss of SUPS MA, reactor trip breakers 1, 2, 5 and 6 open. The reactor trip breaker alignment is such that the reactor will not trip.

B. Incorrect. On a loss of SUPS MA, reactor trip breakers 1, 2, 5 and 6 open. The reactor trip breaker alignment is such that the reactor will not trip. Reactor trip breakers 1, 4, 5 and 8 opening would cause a reactor trip.

C. CORRECT: On a loss of SUPS MA, reactor trip breakers 1, 2, 5 and 6 open. The reactor trip breaker alignment is such that the reactor will not trip.

D. Incorrect. On a loss of SUPS MA, reactor trip breakers 1, 2, 5 and 6 open. The reactor trip breaker alignment is such that the reactor will not trip. Reactor trip breakers 1, 4, 5 and 8 opening would cause a reactor trip.

Technical Reference(s): OP-901-312 revision 306 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PP030 obj. 3 (As available)

Question Source: Bank # X 2010 SRO Exam Q53 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam 2010 SRO Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 Comments:

Revision 0 Facility: Waterford 3 Page 30 of 202

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 000058 AA1.02 Importance Rating 3.1 K/A Statement Ability to operate and / or monitor the following as they apply to the Loss of DC Power: Static inverter dc input breaker, frequency meter, ac output breaker, and ground fault detector Proposed Question: RO 16 Rev: 0 The ATC reports that indication has been lost on the following CP-4 components:

CVC-101, Letdown to Regen HX from RCS Loop 2B CVC-109, Letdown HX Inlet Header Isolation CVC-510, Borated Water to VCT Header Isolation All 3 components indicate closed on the PMC.

Based on these indications, which listed SUPS AC output breaker has tripped open?

A. SA B. SB C. SAB D. SMD Revision 0 Facility: Waterford 3 Page 31 of 202

2012 NRC Exam RO Written Exam Worksheet Proposed Answer: A Explanation: (Optional)

A. CORRECT: All valves listed are powered from SUPS SA. A loss of SUPS SA power supply will cause the CVC valves to reposition to their fail positions and indication will be lost because the indication is powered from the same source.

B. Incorrect: A loss of SUPS SB does not affect these valves.

C. Incorrect: A loss of SUPS SAB does not affect these valves.

A. Incorrect: A loss of SUPS SMD does not affect these valves.

OP-901-312 revision 306, Section B1, Page 4; Technical Reference(s): Section B2, Pages 40, 41 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PPO30 Obj 4 (As available)

Question Source: Bank # X 2010 SRO Exam Q21 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam 2010 SRO Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Comments:

Revision 0 Facility: Waterford 3 Page 32 of 202

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 000065 AK3.04 Importance Rating 3.0 K/A Statement Knowledge of the reasons for the following responses as they apply to the Loss of Instrument Air: Cross-over to backup air supplies.

Proposed Question: RO 17 Rev: 0 Given:

The plant has been manually tripped due to a loss of instrument air.

The crew has entered OP-901-511, Instrument Air Malfunction and OP-902-001, Reactor Trip Recovery.

The Essential Air System will be aligned to service ____(1)____ during a loss of instrument air. Essential Air system is required to ensure _____(2)_____ is maintained.

(1) (2)

A. after 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months containment integrity B. after 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months ECCS functionality C. immediately containment integrity D. immediately ECCS functionality Revision 0 Facility: Waterford 3 Page 33 of 202

2012 NRC Exam RO Written Exam Worksheet Proposed Answer: A Explanation: (Optional)

A. CORRECT: OP-901-511 step 18 directs the crew to align essential air to instrument air 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months after instrument air is lost. The reason for aligning essential air is to provide air to selected containment isolation valves that fail open on loss of instrument air.

B. Incorrect. The reason for aligning essential air is to provide air to selected containment isolation valves that fail open on loss of instrument air.

C. Incorrect. OP-901-511 step 18 directs the crew to align essential air to instrument air 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months after instrument air is lost. (local accumulators have a 10 hour1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months capacity).

D. Incorrect. OP-901-511 step 18 directs the crew to align essential air to instrument air 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months after instrument air is lost. The reason for aligning essential air is to provide air to selected containment isolation valves that fail open on loss of instrument air.

Technical Reference(s): OP-901-511 revision 9 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-AIR00 Obj. 4 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7,10 55.43 Comments:

Revision 0 Facility: Waterford 3 Page 34 of 202

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 000077 AK1.02 Importance Rating 3.3 K/A Statement Knowledge of the operational implications of the following concepts as they apply to Generator Voltage and Electric Grid Disturbances: Over-excitation Proposed Question: RO 18 Rev: 0 Given:

Reactor power is 100%.

The BOP operator has been directed to raise Main Generator voltage to address grid instability issues.

Reactive load (MVAR) is administratively limited to __(1)__ MVAR out. Excessive over-excitation is a major concern due to increased chances of _______(2)______.

(1) (2)

A. 75 pole slippage B. 75 high rotor field winding temperature C. 400 pole slippage D. 400 high rotor field winding temperature Revision 0 Facility: Waterford 3 Page 35 of 202

2012 NRC Exam RO Written Exam Worksheet Proposed Answer: D Explanation: (Optional)

A. Incorrect. OP-010-004 provides the administrative limits for outgoing MVARS. This limit is 400 MVARs OUT. (75 MVARs IN is the limit for incoming MVARs per OP-010-004). Pole slippage is a major concern when excessive under-excitation conditions exist.

B. Incorrect. OP-010-004 provides the administrative limits for outgoing MVARS. This limit is 400 MVARs OUT. (75 MVARs IN is the limit for incoming MVARs per OP-010-004).

C. Incorrect. Pole slippage is a major concern when excessive under-excitation conditions exist.

D. CORRECT: OP-010-004 provides the administrative limits for outgoing MVARS.

This limit is 400 MVARs OUT. Excessive over-excitation is a major concern due to increased rotor winding temperature.

Technical Reference(s): OP-010-004 Revision 315 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-GEN00 Obj. 6 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 5,10 55.43 Comments:

Revision 0 Facility: Waterford 3 Page 36 of 202

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 2 K/A # 000001 AA1.01 Importance Rating 3.5 K/A Statement Ability to operate and / or monitor the following as they apply to the Continuous Rod Withdrawal: Bank select switch Proposed Question: RO 19 Rev: 0 Given:

The plant is at 5% power.

Group P CEAs are 85 inches withdrawn.

All other CEAs are at the Upper Electrical Limit.

The ATC withdraws Group P rods 3 steps and releases the CEA Manual Shim switch.

CEA Group P CEA continued stepping out.

Which of the following is performed FIRST per OP-901-102, CEA or CEDMCS Malfunction?

A. Momentarily place the Manual Shim Switch to insert and then release.

B. Place the Group Select Switch to the Reg Group 6 position.

C. Place the CEDMCS Mode Select switch to OFF.

D. Trip the Reactor and verify all CEAs insert.

Revision 0 Facility: Waterford 3 Page 37 of 202

2012 NRC Exam RO Written Exam Worksheet Proposed Answer: C Explanation: (Optional)

A. Incorrect. Placing the CEA manual shim switch to insert and releasing it is not a step in the off-normal for continuous motion of a group of CEAs: however, it does hold the possibility of correcting the situation by exercising the switch contacts.

B. Incorrect. The Mode Select switch to OFF is the method used to attempt to stop rod motion. Selecting a different group of CEAs is not a step in the procedure for continuous movement of CEAs: however selecting a group of CEAs that is fully withdrawn does hold the possibility of correcting the situation.

C. CORRECT: Placing the CEMCS mode select switch to OFF is the first step directed in the continuous movement of CEA group section of the off-normal. This step is performed in an attempt to stop the CEA group movement.

D. Incorrect. Tripping the Reactor is performed when placing the CEDMCS mode select switch in OFF does not stop rod movement but is not the first step performed and not the preferred sequence.

Technical Reference(s): OP-901-102, Rev 5 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PP010 obj. 4 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 Comments:

Revision 0 Facility: Waterford 3 Page 38 of 202

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 2 K/A # 000024 AK2.01 Importance Rating 2.7 K/A Statement Knowledge of the interrelations between Emergency Boration and the following:

Valves Proposed Question: RO 20 Rev: 0 Given:

Emergency Diesel Generator A is Danger Tagged out.

The plant has experienced a Steam line leak in MODE 3 resulting in an uncontrolled cooldown.

A loss of power to the 3A safety bus has occurred.

The CRS directs the ATC operator to commence Emergency Boration due to the uncontrolled cooldown.

Which of the following will be the emergency boration flow path aligned by the ATC?

A. Boric Acid Makeup (BAM) Tank B via BAM Pump B and BAM-133, Emergency Boration Valve.

B. Refueling Water Storage Pool (RWSP) via CVC-507, RWSP to Charging Pump Suction Isolation.

C. BAM Tanks via BAM-143, Direct Boration Valve.

D. BAM Tanks via BAM-113 A and B, Boric Acid Gravity Feed Valves.

Revision 0 Facility: Waterford 3 Page 39 of 202

2012 NRC Exam RO Written Exam Worksheet Proposed Answer: D Explanation: (Optional)

A. Incorrect. The loss of the 3A bus resulted in a loss of both Boric Acid Makeup (BAM) pumps (both BAM pumps are A train powered). Emergency boration must be aligned from the BAMTs via BAM-113A and B (both gravity feed valves are B train powered).

B. Incorrect. Emergency boration from the RWSP is only aligned when the BAMTs are

< 15%.

C. Incorrect. BAM-143 is the direct boration valve from the BAMTs. BAM-143 is not an optional emergency boration flowpath per OP-901-103, Emergency Boration.

D. CORRECT: The loss of the 3A bus resulted in a loss of the BAM pumps. Emergency boration must be aligned from the BAMTs via BAM-113A and B (both gravity feed valves are B train powered).

Technical Reference(s): OP-901-103 Revision 2.

(Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-CVC00 Obj. 3 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 6,10 55.43 Comments:

Revision 0 Facility: Waterford 3 Page 40 of 202

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 2 K/A # 000028 AK3.02 Importance Rating 2.9 K/A Statement Knowledge of the reasons for the following responses as they apply to the Pressurizer Level Control Malfunction: Relationships between PZR pressure increase and reactor makeup/letdown imbalance Proposed Question: RO 21 Rev: 0 Given:

The plant is operating at 100% power.

Charging Pumps A and B are running.

The Pressurizer Level Control CHANNEL SELECT switch is selected to Channel X.

RC-ILT-0110X has failed low.

Assuming no operator actions are taken, Pressurizer pressure will INITIALLY:

A. rise due to pressurizer backup heaters energizing.

B. rise due to backup charging pumps starting.

C. lower due to backup charging pump securing.

D. lower due to pressurizer spray valve opening.

Revision 0 Facility: Waterford 3 Page 41 of 202

2012 NRC Exam RO Written Exam Worksheet Proposed Answer: B Explanation: (Optional)

A. Incorrect. The pressurizer backup heaters would energize if the pressurizer level instrument failed high.

B. CORRECT: Pressurizer pressure will initially rise due to the Pressurizer level instrument failing low. The PLCS will see a low level in the pressurizer and start backup charging pumps. This will cause pzr level to rise with a corresponding rise in pressure.

C. Incorrect. Pressurizer pressure will initially rise due to the Pressurizer level instrument failing low. The PLCS will see a low level in the pressurizer and start backup charging pumps. The backup charging pump would secure if the level instrument failed high.

D. Incorrect. Pressurizer pressure will lower. The pressurizer spray valves would eventually open in this PLCS failure once Pzr pressure rose to 2275 PSIA.

Technical Reference(s): OP-901-110 revision 6 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PP010 obj. 3 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7, 10 55.43 Comments:

Revision 0 Facility: Waterford 3 Page 42 of 202

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 2 K/A # 000036 AK1.03 Importance Rating 4.0 K/A Statement Knowledge of the operational implications of the following concepts as they apply to Fuel Handling Incidents: Indications of approaching criticality Proposed Question: RO 22 Rev: 0 Given:

The plant is in a refueling outage Core reload is in progress The refuel SRO notifies the control room that an irradiated fuel assembly has been dropped over the core Startup channel counts have risen from 10 CPS to 50 CPS.

Startup Rate is 0.0 DPM and stable Indications available to the ATC operator to monitor for criticality are ____(1)____. The conditions given indicate the reactor is _____(2)_____.

(1) (2)

A. startup channels, and boron dilution NOT critical alarms B. startup and control channels NOT critical C. startup channels, and boron dilution critical alarms D. startup and control channels critical Revision 0 Facility: Waterford 3 Page 43 of 202

2012 NRC Exam RO Written Exam Worksheet Proposed Answer: A Explanation: (Optional)

A. CORRECT: Startup channels and boron dilution alarms are available to the operators at this power level. Startup channel counts not steadily rising and SUR not positive and sustained is indications the reactor is not critical.

B. Incorrect. The control channels are not available to the operators at this power level.

C. Incorrect. Startup channel counts not steadily rising and SUR not positive and sustained is indications the reactor is not critical.

D. Incorrect. The control channels are not available to the operators at this power level.

Startup channel counts not steadily rising and SUR not positive and sustained is indications the reactor is not critical.

Technical Reference(s): OP-010-003 revision 324 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-ENI00 obj. 4 (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 1, 2 55.43 Comments:

Revision 0 Facility: Waterford 3 Page 44 of 202

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 2 K/A # 000051 AA2.02 Importance Rating 3.9 K/A Statement Ability to determine and interpret the following as they apply to the Loss of Condenser Vacuum: Conditions requiring reactor and/or turbine trip.

Proposed Question: RO 23 Rev: 0 Given:

The plant was initially at 100% power Condenser vacuum is 20.5 INHG and continuing to lower slowly.

The crew is implementing OP-901-220, Loss of Condenser Vacuum and OP-901-212, Rapid Plant Down power.

Which of the following is the appropriate action for this event?

A. Continue the rapid plant downpower until vacuum recovers.

B. Trip the Main Turbine and then verify Reactor Power Cutback.

C. Trip the Reactor and then verify the Main Turbine trips.

D. Trip the Main Turbine and then trip the reactor.

Revision 0 Facility: Waterford 3 Page 45 of 202

2012 NRC Exam RO Written Exam Worksheet Proposed Answer: C Explanation: (Optional)

A. Incorrect. OP-901-220 directs the crew to perform a rapid plant power reduction at 25 Hg condenser vacuum. If vacuum continues to lower during the rapid plant down power further direction is to trip the Reactor if condenser vacuum approaches 20 Hg.

B. Incorrect. The main turbine would trip at 20 Hg condenser vacuum and cause a reactor power cutback but OP-901-220 directs the crew to trip the reactor and then verify a turbine trip.

C. CORRECT: OP-901-220 directs the crew to trip the reactor and verify the turbine is tripped if Condenser vacuum has not stabilized and is approaching the Main Turbine Trip value of 20 Hg.

D. Tripping the Main Turbine first could challenge the integrity of the condenser by supplying high energy steam directly to the condenser via Steam Bypass valves Technical Reference(s): OP-901-220 revision 302 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PP020 obj. 3 (As available)

Question Source: Bank # X 03965a 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 10 55.43 Comments:

Revision 0 Facility: Waterford 3 Page 46 of 202

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 2 K/A # 000067 G2.4.11 Importance Rating 4.0 K/A Statement Knowledge of abnormal condition procedures. (Plant Fire On-site)

Proposed Question: RO 24 Rev: 0 OP-901-503, Isolation Panel Fire, directs both performing a ________(1)_______ and initiating a _____(2)_____ as part of the mitigating strategy.

(1) (2)

A. manual reactor trip Safety Injection Actuation Signal B. manual reactor trip Main Steam Isolation Signal C. rapid plant shutdown Safety Injection Actuation Signal D. rapid plant shutdown Main Steam Isolation Signal Revision 0 Facility: Waterford 3 Page 47 of 202

2012 NRC Exam RO Written Exam Worksheet Proposed Answer: B Explanation: (Optional)

A. Incorrect. OP-901-503 directs the crew to manually initiate a MSIS to prevent spurious actuations in the secondary from affecting the plant. A SIAS is not manually actuated from an isolation panel fire.

B. CORRECT: OP-901-503 directs the crew to trip the reactor on an isolation panel fire. OP-901-503 also directs the crew to manually initiate a MSIS to prevent spurious actuations in the secondary from affecting the plant. Both actions are 10 minute time critical steps.

C. Incorrect. OP-901-503 directs the crew to trip the reactor on an isolation panel fire.

OP-901-503 directs the crew to manually initiate a MSIS to prevent spurious actuations in the secondary from affecting the plant. A SIAS is not manually actuated from an isolation panel fire.

D. Incorrect. OP-901-503 directs the crew to trip the reactor on an isolation panel fire.

Technical Reference(s): OP-901-503 revision 307 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PP050 obj. 3 (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 Comments:

Revision 0 Facility: Waterford 3 Page 48 of 202

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 2 K/A # 000068 AK3.12 Importance Rating 4.1 K/A Statement Knowledge of the reasons for the following responses as they apply to the Control Room Evacuation: Required sequence of actions for emergency evacuation of control room Proposed Question: RO 25 Rev: 0 Given The control room has been evacuated in accordance with OP-901-502, Evacuation of Control Room and Subsequent Plant Shutdown due to a fire at CP-8.

Controls have been established at LCP-43.

Which of the following indicates the method used to maintain pressurizer level 33% to 60% at LCP-43 by operating:

A. Charging Pump B manually.

B. Pressurizer level controller in automatic.

C. Charging Pump A manually.

D. Pressurizer level controller in manual.

Revision 0 Facility: Waterford 3 Page 49 of 202

2012 NRC Exam RO Written Exam Worksheet Proposed Answer: A Explanation: (Optional)

A. CORRECT: Charging Pump B is cycled to maintain pressurizer level during a Control Room Evacuation with fire to make up for CBO IAW section E1 of OP-901-502.

B. Incorrect. Pressurizer level is maintained with the Pressurizer level controller in Auto when the control room is evacuated without a fire (in this instance, there is a fire in the control room). Letdown is isolated during immediate actions for a control evacuation with fire.

C. Incorrect. Charging Pump A will have no power when the control room is evacuated with a fire.

D. Incorrect. Letdown is isolated during immediate actions for a control evacuation with fire. Adjusting the pressurizer level control valve will have no effect.

Technical Reference(s): OP-901-502 revision 21 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PPO51 obj. 14 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 6, 10 55.43 Comments:

Revision 0 Facility: Waterford 3 Page 50 of 202

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 2 K/A # 00074 EA1.01 Importance Rating 4.2 K/A Statement Ability to operate and monitor the following as they apply to a Inadequate Core Cooling: RCS Water Inventory.

Proposed Question: RO 26 Rev: 0 Which of the following parameters will provide a void indication on QSPDS 1 with the Reactor Vessel Level Monitoring System?

A. A Core Exit Thermocouple indicates temperature of 725 °F.

B. An Unheated Junction Thermocouple indicates temperature of 725 °F.

C. A Heated versus Unheated Junction Thermocouple differential temperature of 50

°F.

D. A Core Exit Thermocouple versus Unheated Junction Thermocouple differential temperature of 50 °F.

Revision 0 Facility: Waterford 3 Page 51 of 202

2012 NRC Exam RO Written Exam Worksheet Proposed Answer: B Explanation: (Optional)

A. Incorrect: A temperature of > 700 °F will cause a void indication, but not as supplied from a Core Exit Thermocouple.

B. CORRECT: A temperature of > 700 °F will cause a void indication, and this is the correct input component.

C. Incorrect: A Heated versus Unheated Junction Thermocouple differential temperature can generate a void indication, but the value that causes this indication is 200 °F. 50 °F is the value that a work request is required when performing the surveillance for the Reactor Vessel Level Monitoring System, OP-903-013, Monthly Channel Checks.

D. Incorrect: A Heated versus Unheated Junction Thermocouple differential temperature can generate a void indication, but the value that causes this indication is 200 °F. This distractor is comparing the Core Exit Thermocouple and Unheated Junction Thermocouple vice the Heated versus Unheated Junction Thermocouple.

Technical Reference(s): SD-QSP Rev. 5 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-QSP00 Obj. 3 (As available)

Question Source: Bank # 6014-A Modified Bank # X (Note changes or attach parent)

New Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 Comments:

Revision 0 Facility: Waterford 3 Page 52 of 202

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 2 K/A # CE/A13 AK1.3 Importance Rating 3.1 K/A Statement Knowledge of the operational implications of the following concepts as they apply to the (Natural Circulation Operations): Annunciators and conditions indicating signals, and remedial actions associated with the (Natural Circulation Operations).

Proposed Question: RO 27 Rev: 0 Given:

A reactor trip occurred due to a loss of offsite power 20 minutes ago

Pressurizer Pressure is 2250 PSIA and slowly rising

Tcold Loop 1A and 1B are 545°F and stable

Thot Loop1 is 568°F and stable

Tcold Loop 2A and 2B are 560°F and slowly rising

Thot Loop 2 is 565°F and stable

Representative CET temperature is 570°F and stable The minimum criteria for verifying single phase natural circulation is____(1)____. The crew should adjust feeding and steaming of Steam Generator __(2)__.

(1) (2)

A. NOT met 1 B. met 1 C. NOT met 2 D. met 2 Revision 0 Facility: Waterford 3 Page 53 of 202

2012 NRC Exam RO Written Exam Worksheet Proposed Answer: D Explanation: (Optional)

A. Incorrect. Wrong conclusion concerning natural circ, the minimum criteria is met Wrong SG, To get natural circulation to meet criteria in both loops Steam Generator 2 heat removal must be adjusted.

B. Incorrect. Correct conclusion concerning natural circ, the minimum criteria is met Wrong SG, To get natural circulation to meet criteria in both loops Steam Generator 2 heat removal must be adjusted.

C. Incorrect. Wrong conclusion concerning natural circ, the minimum criteria is met Correct SG.

D. CORRECT: Natural circulation is verified if EITHER loop. To get natural circulation to meet criteria in both loops Steam Generator 2 heat removal must be adjusted.

Technical Reference(s): OP-902-003 revision 7 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: Steam Tables Learning Objective: WLP-OPS-PPE03 Objective 5 (As available)

Question Source: Bank # 08321 Modified Bank # X (Note changes or attach parent)

New Question History: Last NRC Exam 2008 NRC Exam # 63 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 2,5 55.43 Comments:

Revision 0 Facility: Waterford 3 Page 54 of 202

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 003 K3.02 Importance Rating 3.5 K/A Statement K3.02 Knowledge of the effect that a loss or malfunction of the RCPS will have on the following: S/G Proposed Question: RO 28 Rev: 0 Given:

The plant is at 100% power RCP 1A shaft shears.

All actuations occurred as required RCP Amps would indicate ____(1)____ than normal and a reactor trip would occur due to _____(2)_____.

(1) (2)

A. higher low steam generator flow B. higher low DNBR C. lower low DNBR D. lower low steam generator flow Revision 0 Facility: Waterford 3 Page 55 of 202

2012 NRC Exam RO Written Exam Worksheet Proposed Answer: D Explanation: (Optional)

A. Incorrect. The first portion of the question is incorrect because lower amps would be indicated on the CP-2 indication for RCP due to a decreased load on the RCP. The second portion of the question is correct because the reactor would trip on low SG flow. This is the purpose of the low SG flow trip.

B. Incorrect. The first portion of the question is incorrect because lower amps would be indicated on the CP-2 indication for RCP due to a decreased load on the RCP. The second portion of the question is also incorrect because CPCs (low DNBR) would not trip the Reactor because CPCs detects flow from the RCP speed sensors which is not affected by a sheared shaft.

C. Incorrect. The first portion of the question is correct because lower amps would be indicated on the CP-2 indication for RCP due to a decreased load on the RCP. The second portion of the question is incorrect because CPCs (low DNBR) would not trip the Reactor because CPCs detects flow from the RCP speed sensors, which is not affected by a sheared shaft.

D. CORRECT: The first portion of the question is correct because lower amps would be indicated on the CP-2 indication for RCP due to a decreased load on the RCP. The second portion of the question is correct because the reactor would trip on low SG flow. This is the purpose of the low SG flow trip.

Technical Reference(s): TS 2.2.1 Bases (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-CPC00 obj. 2 (As available)

WLP-OPS-RCP00 obj. 7 Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 2 55.43 Comments:

Revision 0 Facility: Waterford 3 Page 56 of 202

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 004 K5.30 Importance Rating 3.8 K/A Statement K5.30 Knowledge of the operational implications of the following concepts as they apply to the CVCS: Relationship between temperature and pressure in CVCS components during solid plant operation Proposed Question: RO 29 Rev: 0 Given:

Solid Plant operations are in progress

CCW flow is reduced to the in-service SDC Heat Exchanger

The Letdown Back Pressure Controller is operating in Auto RCS pressure will ____(1)____ until the Letdown Back Pressure Control Valve throttles

_____(2)____.

(1) (2)

A. lower closed B. rise closed C. lower open D. rise open Revision 0 Facility: Waterford 3 Page 57 of 202

2012 NRC Exam RO Written Exam Worksheet Proposed Answer: D Explanation: (Optional)

A. Incorrect. Lowering CCW flow will cause temperature and RCS pressure to rise. To restore RCS pressure to setpoint the letdown backpressure valve will be opened (not closed) to raise letdown flow to lower RCS pressure to setpoint.

B. Incorrect. Lowering CCW flow will cause temperature and RCS pressure to rise. To restore RCS pressure to setpoint the letdown backpressure valve will be opened (not closed) to raise letdown flow to lower RCS pressure to setpoint.

C. Incorrect. Lowering CCW flow will cause temperature and RCS pressure to rise. To restore RCS pressure to setpoint the letdown backpressure valve will be opened to raise letdown flow to lower RCS pressure to setpoint.

D. CORRECT: Lowering CCW flow will cause temperature and RCS pressure to rise.

To restore RCS pressure to setpoint the letdown backpressure valve will open to raise letdown flow to lower RCS pressure to setpoint.

Technical Reference(s): OP-010-005 revision 317 (Attach if not previously provided) SD-PLC (including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-CVC00 objective 3 (As available)

Question Source: Bank # X 08405a 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 5 55.43 Comments:

Revision 0 Facility: Waterford 3 Page 58 of 202

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 005 K2.01 Importance Rating 3.0 K/A Statement K2.01 Knowledge of bus power supplies to the following: RHR pumps Proposed Question: RO 30 Rev: 0 Given:

The plant has experienced a Loss of Coolant Accident (LOCA).

The crew has implemented OP-902-002, LOCA Recovery Procedure.

A Recirculation Actuation Signal (RAS) has just occurred.

Low Pressure Safety Injection (LPSI) Pump A failed to automatically or manually trip from the control room.

LPSI flow from the SI sump must be secured within _____(1)_____ of the SI Pumps recirc Isolation valves closing. An operator should be dispatched to switchgear

____(2)____ to locally trip LPSI pump A.

(1) (2)

A. 5 minutes 3A B. 5 minutes 31A C. 2 minutes 3A D. 2 minutes 31A Revision 0 Facility: Waterford 3 Page 59 of 202

2012 NRC Exam RO Written Exam Worksheet Proposed Answer: A Explanation: (Optional)

A. CORRECT: LPSI pump A should auto trip on a RAS. If the LPSI pump did not auto trip and cannot be tripped using the C/S on CP-8, then OP-902-002 step 42 directs the LPSI pump to be tripped locally (3A bus). This action must be completed within 5 minutes of the SI pumps recirc isol valves closed.

B. Incorrect. LPSI Pump A is powered from the 3A bus, not the 31A bus. The time limit of 5 minutes is correct.

C. Incorrect. The power supply is correct for LPSI pump A but the time limit for tripping LPSI Pump A is 5 minutes. The time limit of 2 minutes is the limit for closing the SI recircs upon an RAS.

D. Incorrect. LPSI Pump A is powered from the 3A bus. The time limit for tripping LPSI Pump A is 5 minutes. The time limit of 2 minutes is the limit for closing the SI recircs upon an RAS.

Technical Reference(s): OP-902-002 revision 14 (Attach if not previously provided) TGOP-902-002 revision 13 (including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PPE02 obj. 19 (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 8 55.43 Comments:

Revision 0 Facility: Waterford 3 Page 60 of 202

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 006 K4.11 Importance Rating 3.9 K/A Statement K4.11 Knowledge of ECCS design feature(s) and/or interlock(s) which provide for the following: Reset of SIS Proposed Question: RO 31 Rev: 0 Given:

The plant has experienced an Excess Steam Demand event.

The crew has implemented OP-902-004, Excess Steam Demand Recovery Procedure.

Shutdown Cooling (SDC) Entry conditions are being met and the CRS has directed the ATC to reset the Safety Injection and Containment Isolation signal in accordance with OP-902-009 Attachment 5-A: SIAS and CIAS Pressurizer Pressure Reset Procedure.

CP-10 channel A and CP-10 channel B SIAS and CIAS Pressurizer Pressure has already been reset.

To complete the reset of the safety injection and containment isolation signals, the BOP will:

A. Bypass CP-10 Channel C and Channel D trips, reset the Initiation relays and then reset the Actuation relays for SIAS and CIAS.

B. Bypass CP-10 Channel C and Channel D trips, reset the Actuation relays and then reset the Initiation relays for SIAS and CIAS.

C. Bypass CP-10 Channel C trip, reset the Initiation relays and then reset the Actuation relays for SIAS and CIAS.

D. Bypass CP-10 Channel C trip, reset the Actuation relays and then reset the Initiation relays for SIAS and CIAS.

Revision 0 Facility: Waterford 3 Page 61 of 202

2012 NRC Exam RO Written Exam Worksheet Proposed Answer: C Explanation: (Optional)

A. Incorrect. Only CP-10 Channel C is bypassed before the initiation and actuation relays are reset.

B. Incorrect. Only CP-10 Channel C is bypassed before the initiation and actuation relays are reset. Also, the initiation relays must be reset before the actuation relays can be reset.

C. CORRECT: OP-902-009 Attachment 5-A: SIAS and CIAS Pressurizer Pressure Reset Procedure directs the crew to reset PPS Channel A and B, then bypass Channel C. At this time, the initiation relays are reset and then the actuation relays are reset.

D. Incorrect. The initiation relays must be reset before the actuation relays can be reset.

Technical Reference(s): OP-902-009 Rev. 307 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PPS00 obj. 1 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7,10 55.43 Comments:

Revision 0 Facility: Waterford 3 Page 62 of 202

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 006 A2.11 Importance Rating 4.0 K/A Statement A2.11 Ability to (a) predict the impacts of the following malfunctions or operations on the ECCS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Rupture of ECCS header Proposed Question: RO 32 Rev: 0 Given:

A LOCA has occurred and the crew has diagnosed into OP-902-002, Loss of Coolant Accident.

RCS pressure - 600 psia.

HPSI pumps A and B are running and all Safety Injection Flow Control valves are open.

Cold leg injection header flow rates are as follows:

1A-190 gpm 1B-160 gpm 2A-175 gpm 2B-0 gpm

Fifteen minutes into the event, RVLMS indicates core uncovery and CETs are showing indications of superheat.

Based on this information, the crew should ____(1)____ because

________(2)________.

(1) (2)

A. Remain in OP-902-002, Core uncovery and superheated conditions LOCA Recovery could exist for up to 30 minutes.

B. Remain in OP-902-002, the SI flow curves are being met.

LOCA Recovery C. Exit to OP-902-008, Core uncovery and superheated conditions Functional Recovery should not be expected at this time.

D. Exit to OP-902-008, the SI flow curves are not being met.

Functional Recovery Revision 0 Facility: Waterford 3 Page 63 of 202

2012 NRC Exam RO Written Exam Worksheet Proposed Answer: D Explanation: (Optional)

A. Incorrect. Core uncovery and superheat conditions may be expected for up to 30 minutes as indicated by the note in the SFSC, but to stay in OP-902-002, the minimum SI flows must be met. In this case, they are not.

B. Incorrect. SI flows are not being met. The HPSI flow curve is for individual lines not trains as is the LPSI flow curve.

C. Incorrect. The reason for exiting to OP-902-008 is not because plenum level is not being met. Core uncovery can be expected for up to 30 minutes and stay in this optimal but SI flow rates must be met.

D. CORRECT: The crew must exit to OP-902-008 because Core uncovery and superheat conditions may be expected for up to 30 minutes as indicated by the note in the SFSC, but to stay in OP-902-002, the minimum SI flows must be met. In this case, they are not.

Technical Reference(s): OP-902-002 rev 14, OP-902-009 rev 305 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: OP-902-009 Appendix 2, Att. 2-E Learning Objective: WLP-OPS-PPE02 obj. 7 (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 8, 10 55.43 Comments:

Revision 0 Facility: Waterford 3 Page 64 of 202

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 007 A1.02 Importance Rating 2.7 K/A Statement A1.02 Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the PRTS controls including:

Maintaining quench tank pressure Proposed Question: RO 33 Rev: 0 Given:

A Loss of Coolant Accident is in progress.

The crew has implemented OP-902-002, LOCA Recovery Procedure.

The crew is depressurizing the RCS to Shutdown Cooling (SDC) entry conditions.

Voiding in the Reactor Vessel Head is inhibiting the depressurization of the RCS.

The CRS directs the BOP to vent the Reactor vessel head to the Quench Tank in accordance with Appendix 11 of OP-902-009, Void Elimination.

Venting the vessel head using the RCS vent valves will ____(1)____ in the vessel head.

Venting of the reactor vessel head should be stopped before reaching ____(2)____ in the quench tank to prevent rupturing the quench tank rupture disc.

(1) (2)

A. remove the non- 120 psig condensible gasses B. collapse the steam 150 psig bubble C. remove the non-condensible gasses 150 psig D. collapse the steam 120 psig bubble Revision 0 Facility: Waterford 3 Page 65 of 202

2012 NRC Exam RO Written Exam Worksheet Proposed Answer: A Explanation: (Optional)

A. CORRECT: Venting the reactor head using RCS vent valves is performed when non-condensable gas removal is desired. Venting the head will not collapse a steam bubble. The setpoint for the Quench Tank rupture disc is 124 psig.

B. Incorrect. Venting the reactor head using RCS vent valves is performed when non-condensable gas removal is desired. Venting the head will not collapse a steam bubble. The setpoint for the Quench Tank rupture disc is 124 psig.

C. Incorrect. The setpoint for the Quench Tank rupture disc is 124 psig.

D. Incorrect. Venting the reactor head using RCS vent valves is performed when non-condensable gas removal is desired. Venting the head will not collapse a steam bubble.

Technical Reference(s): OP-902-002 revision 14 (Attach if not previously provided) OP-902-009 Appendix 11 revision 307 (including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-RCS00 obj. 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 Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 3, 8 55.43 Comments:

Revision 0 Facility: Waterford 3 Page 66 of 202

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 008 K1.05 Importance Rating 3.0 K/A Statement K1.05 Knowledge of the physical connections and/or cause-effect relationships between the CCWS and the following systems: Sources of makeup water Proposed Question: RO 34 Rev: 0 The normal source of makeup to the Auxiliary Component Cooling Water Cooling Tower is the _______(1)_______. The normal source of makeup water to the Component Cooling Water Surge Tank is the _______(2)_________.

(1) (2)

A. CCW Makeup Pumps Condensate Storage Tank Pumps B. CCW Makeup Pumps Hotwell Transfer Pump C. Condensate Transfer Pump Hotwell Transfer Pump D. Condensate Transfer Pump Condensate Storage Tank Pumps Revision 0 Facility: Waterford 3 Page 67 of 202

2012 NRC Exam RO Written Exam Worksheet Proposed Answer: D Explanation: (Optional)

A. Incorrect: CCW makeup Pumps are for large water loss to the CCW Surge Tank.

B. Incorrect: The Hotwell Transfer Pump is a standby pump in the CMU system, but not capable of pumping to the elevation of the CCW Surge Tank.

C. Incorrect. The Hotwell Transfer Pump is a standby pump in the CMU system, but not capable of pumping to the elevation of the CCW Surge Tank.

D. CORRECT: The Condensate Transfer Pump provides water to both ACCW basins and the CMU pressurized header provides water for the CCW Surge Tank.

Technical Reference(s): SD-CMU Fig. 2 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-CC00 Obj. 7 (As available)

Question Source: Bank # X 2009 RO Exam Q34 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam 2009 RO Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 Comments:

Revision 0 Facility: Waterford 3 Page 68 of 202

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 008 G2.2.12 Importance Rating 3.7 K/A Statement 2.2.12 Knowledge of surveillance procedures. Component Cooling Water Proposed Question: RO 35 Rev: 0 Given:

The plant is operating at 100% power.

The crew is performing OP-903-118, Primary Quarterly IST Valve Tests for Component Cooling Water valves.

The BOP closes CC-200A, CCW Header A to AB Supply Isolation, for a stroke time.

CC-963A, Shutdown Heat Exchanger A CCW Flow Control, will open for 5 seconds and then ____(1)____. CC-963A, Shutdown Heat Exchanger A CCW Flow Control, opens to

_____(2)_____.

(1) (2)

A. must be manually closed maintain minimum flow requirements through the SDC Heat Exchanger B. must be manually closed reduce pressure transients on the CCW system C. will auto close maintain minimum flow requirements through the SDC Heat Exchanger D. will auto close reduce pressure transients on the CCW system Revision 0 Facility: Waterford 3 Page 69 of 202

2012 NRC Exam RO Written Exam Worksheet Proposed Answer: D Explanation: (Optional)

A. Incorrect. CC-963A will automatically close after 5 seconds if opened due to CC-200A going closed. This interlock exists to reduce pressure transients in the CCW system (not to maintain minimum flow requirements).

B. Incorrect. CC-963A will automatically close after 5 seconds if opened due to CC-200A going closed.

C. Incorrect. CC-963A will automatically close after 5 seconds if opened due to CC-200A going closed. This interlock exists to reduce pressure transients in the CCW system (not to maintain minimum flow requirements).

D. CORRECT: CC-963A will automatically close after 5 seconds if opened due to CC-200A going closed. This interlock exists to reduce pressure transients in the CCW system.

Technical Reference(s): OP-903-118 revision 27 (Attach if not previously provided) SD-CC revision 17 (including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-CC00 obj. 3 (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 Comments:

Revision 0 Facility: Waterford 3 Page 70 of 202

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 010 K6.01 Importance Rating 2.7 K/A Statement K6.01 Knowledge of the effect of a loss or malfunction of the following will have on the PZR PCS: Pressure detection systems Proposed Question: RO 36 Rev: 0 The plant is at 100% power:

If the controlling Pressurizer pressure instrument RC-IPI-0100 X fails high, with no operator actions, the first reactor trip signal would be generated by a:

A. CPC RCS Pressure Aux Trip LO signal.

B. CPC RCS Pressure Aux Trip HI signal.

C. RPS Pressurizer Pressure LO signal.

D. RPS Pressurizer Pressure HI signal.

Revision 0 Facility: Waterford 3 Page 71 of 202

2012 NRC Exam RO Written Exam Worksheet Proposed Answer: A Explanation: (Optional)

A. CORRECT. The Aux trip would occur first because 1) RCS pressure would be lowering vice increasing; and 2) the aux trip occurs at 1860 PSIA vice 1684 PSIA for the Pressurizer Pressure LO trip.

B. Incorrect. Channel failing high causes spray valves to open and heaters to trip, reducing RCS pressure.

C. Incorrect. Can cause a trip on low RCS pressure, but setpoint is lower than CPC trip, so CPCs will trip unit first.

D. Channel failing high causes spray valves to open and heaters to trip, reducing RCS pressure.

TS 2.0 Bases, DNBR-Low Technical Reference(s): SD-CPC, Pages 35 & 37 (Attach if not previously provided) OP-901-120, revision 301 Page 4, Step 2 (including version/revision number) OP-009-007, rev 15 Page 20, Steps 9.5 & 9.6 Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PPS Obj. 7 (As available)

Question Source: Bank # X 2008 RO Exam Q8 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam 2008 RO Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Comments:

Revision 0 Facility: Waterford 3 Page 72 of 202

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 012 A4.03 Importance Rating 3.6 K/A Statement A4.03 Ability to manually operate and/or monitor in the control room: Channel blocks and bypasses Proposed Question: RO 37 Rev: 0 A reactor trip has occurred, which of the following Operating Bypasses will automatically change state with no operator action?

A. Steam Generator High Level Trip B. Steam Generator Flow Low Trip C. Log Power High Trip D. CPC Trip Revision 0 Facility: Waterford 3 Page 73 of 202

2012 NRC Exam RO Written Exam Worksheet Proposed Answer: C Explanation: (Optional)

A. Incorrect: Normally the keyswitch is OFF. For this to automatically change state would require a switch to be mispositioned.

B. Incorrect. Normally the keyswitch is OFF. For this to automatically change state would require a switch to be mispositioned.

C. CORRECT: The High Log Power Trip requires no operator action post trip to change state. Operator action would only be required during a startup to activate the bypass.

D. Incorrect. Normally the keyswitch is OFF. For this to automatically change state would require a switch to be mispositioned.

Technical Reference(s): OP-009-007 (Attach if not previously provided) SD-PPS (including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PPS00 obj. 4 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Comments:

Revision 0 Facility: Waterford 3 Page 74 of 202

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 013 K3.01 Importance Rating 4.4 K/A Statement K3.01 Knowledge of the effect that a loss or malfunction of the ESFAS will have on the following: Fuel Proposed Question: RO 38 Rev: 0 Given:

HPSI Pump A is tagged out for impeller replacement and HPSI pump AB is NOT aligned to replace HPSI Pump A A LOCA has occurred RCS Pressure stabilizes at 750 psia Charging Pumps A and B are running HPSI pump B failed to start due to a faulty relay IF these conditions continue with NO operator action, which of the following describes the effect on the fuel assemblies?

A. Fuel failure will occur. Minimum safety function requirements are NOT met and the break is too small for other ECCS equipment to provide core cooling.

B. Fuel failure will occur. Minimum safety function requirements are met but the break is too large for other ECCS to provide core cooling.

C. Fuel failure will NOT occur. Safety Injection Tanks will provide core reflood.

D. Fuel failure will NOT occur. Steam Generators will provide core cooling.

Revision 0 Facility: Waterford 3 Page 75 of 202

2012 NRC Exam RO Written Exam Worksheet Proposed Answer: A Explanation: (Optional)

A. CORRECT: The parameters given are for a break size that does not allow core reflood by either the SITs or LPSI Pumps. RCS pressure is well above LPSI shutoff head. Therefore fuel failure will occur and minimum safety function requirements are not met because there is no HPSI pumps running.

B. Incorrect. Fuel failure will occur The parameters given are for a break size that does not allow core reflood by either the SITs or LPSI Pumps.

C. Incorrect. Fuel failure will occur Safety Injection Tanks will no dump until 600-650 psia.

D. Incorrect. The Steam Generators will not provide any cooling without additional operator action. Additionally, without sufficient makeup capability the vessel fluid will eventually be lost to boiloff.

Technical Reference(s): OP-902-002 revision 14 (Attach if not previously provided) TG-OP-902-002 (including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PPE02 obj. 16 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 2 55.43 Comments:

Revision 0 Facility: Waterford 3 Page 76 of 202

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 013 A3.01 Importance Rating 3.7 K/A Statement A3.01 Ability to monitor automatic operation of the ESFAS including: Input channels and logic.

Proposed Question: RO 39 Rev: 0 Given:

Pressurizer Pressure is 1650 psia and slowly lowering

Containment Pressure is 16.7 and slowly rising

Steam Generator 1 Pressure is 700 psia and lowering

Steam Generator 2 Pressure is 680 psia and lowering At this time the following ESFAS signal(s) should have been generated:

A. SIAS only B. SIAS and CIAS only C. SIAS, CIAS, and MSIS only D. SIAS, CIAS, MSIS, and CSAS Revision 0 Facility: Waterford 3 Page 77 of 202

2012 NRC Exam RO Written Exam Worksheet Proposed Answer: B Explanation: (Optional)

A. Incorrect. SIAS and CIAS will be generated together at PZR pressure of 1684 psia.

B. CORRECT: PZR Pressure at 1684 psia will generate an SIAS and CIAS.

C. Incorrect. MSIS will not be generated until one SG pressure drops below 666 psia or Containment pressure rises to 17.1 psia.

D. Incorrect. MSIS will not be generated until one SG pressure drops below 666 psia or Containment pressure rises to 17.1 psia. CSAS will not be generated until Containment pressure rises to 17.7 psia.

Technical Reference(s): TS 2.2.1 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PPS00 obj. 3 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 4,7 55.43 Comments:

Revision 0 Facility: Waterford 3 Page 78 of 202

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 022 K4.04 Importance Rating 2.8 K/A Statement K4.04 Knowledge of CCS design feature(s) and/or interlock(s) which provide for the following: Cooling of control rod drive motors Proposed Question: RO 40 Rev: 0 Given the following:

The plant was at 100% power.

Containment Fan Coolers A, B, & D are in operation.

Containment Fan Cooler C is in standby.

CEDM Fans A and D are in operation.

Due to an RCS leak, the reactor was manually tripped and SI was manually actuated.

Which describes the response during the event?

A. Containment Fan Coolers A, B, C, & D will operate in slow speed. CEDM Fans A and D will trip.

B. Containment Fan Coolers A, B, & D will operate in slow speed. CEDM Fans A and D will trip.

C. Containment Fan Coolers A, B, C, & D will operate in slow speed. CEDM Fans A and D will continue to operate.

D. Containment Fan Coolers A, B, D will operate in slow speed. CEDM Fans A and D will continue to operate.

Revision 0 Facility: Waterford 3 Page 79 of 202

2012 NRC Exam RO Written Exam Worksheet Proposed Answer: A Explanation: (Optional)

A. CORRECT: On a SIAS, all four Containment Fan Coolers start in slow speed and any running CEDM fans will trip.

B. Incorrect. On a SIAS, all four Containment Fan Coolers start in slow speed and any running CEDM fans will trip.

C. Incorrect. On a SIAS, all four Containment Fan Coolers start in slow speed and any running CEDM fans will trip.

D. Incorrect. On a SIAS, all four Containment Fan Coolers start in slow speed and any running CEDM fans will trip.

Technical Reference(s): OP-008-004, Revision 007 (Attach if not previously provided) OP-008-003, Revision 301 (including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-CCS00 obj. 1 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 4 55.43 Comments:

Revision 0 Facility: Waterford 3 Page 80 of 202

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 026 K1.01 Importance Rating 4.2 K/A Statement K1.01 Knowledge of the physical connections and/or cause effect relationships between the CSS and the following systems: ECCS Proposed Question: RO 41 Rev: 0 Given the following:

A LOCA has occurred.

All systems are operating as designed.

RCS pressure is 700 psia.

Containment pressure is 18 psia.

RWSP level is 9% and lowering.

NO operator actions have been taken.

Which ONE of the following describes the alignment of the Containment Spray System?

A. SIS Sump Outlet valves, SI-602A and B open RWSP Outlet Isolation valves, SI-106A and B open CS Pump Recirculation valves, SI-120A and B closed B. SIS Sump Outlet valves, SI-602A and B closed RWSP Outlet Isolation valves, SI-106A and B open CS Pump Recirculation valves, SI -120A and B open C. SIS Sump Outlet valves, SI-602A and B open RWSP Outlet Isolation valves, SI-106A and B open CS Pump Recirculation valves, SI -120A and B open D. SIS Sump Outlet valves, SI-602A and B closed RWSP Outlet Isolation valves, SI-106A and B open CS Pump Recirculation valves, SI -120A and B closed Revision 0 Facility: Waterford 3 Page 81 of 202

2012 NRC Exam RO Written Exam Worksheet Proposed Answer: C Explanation: (Optional)

A. Incorrect. Recirc valves remain open, and must be manually closed by the operator.

B. Incorrect. SI-602A and B open on a RAS.

C. CORRECT: Less than 10% RWSP, RAS actuates. SIS Sump suction valves (SI-106A and B, SI-602A and B realign; recirc valves remain open and must be manually closed by the operator.

D. Incorrect. SI-602A and B open on a RAS. SI-120A and B remain open after an RAS and must be manually reclosed.

Technical Reference(s): OP-902-002, Revision 14 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PPE02 obj. 19 (As available)

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

New Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 Comments:

Revision 0 Facility: Waterford 3 Page 82 of 202

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 039 A4.04 Importance Rating 3.8 K/A Statement A4.04 Ability to manually operate and/or monitor in the control room: Emergency feedwater pump turbines Proposed Question: RO 42 Rev: 0 Given:

The plant has experienced a Loss of Offsite Power from MODE 3 with both diesels running loaded.

5 minutes into the event EFAS1 and EFAS2 initiates and EFW Pump AB trips on ELECTRICAL overspeed.

The BOP has closed both MS 401A and MS 401B, EFW Pump AB Turb Stm Supply Valves The EFW Pump AB steam line has been depressurized.

Select the correct response for resetting MS-416, EFW Pump AB Stop Valve for these conditions.

A. MS-416 will drive shut automatically. When closed the valve operator re-latches.

MS-416 will be re-opened from CP-8.

B. An NAO must manually close MS-416. When closed the valve operator re-latches. MS-416 must be opened locally by the NAO.

C. MS-416 motor will drive shut automatically. The NAO must reset the local linkage. MS-416 will be reopened from CP-8.

D. An NAO must manually close MS-416. The NAO must reset the local linkage.

MS-416 must be opened locally by the NAO.

Revision 0 Facility: Waterford 3 Page 83 of 202

2012 NRC Exam RO Written Exam Worksheet Proposed Answer: A Explanation: (Optional)

A. CORRECT: AB 311 re-energizes when the EDG output breaker for the side the AB busses are tied to closes. Operation of MS-416 can be performed in this case from CP-8.

B. Incorrect. MS-416 has power and an AO is not needed to operate the valve.

C. Incorrect. The linkage would only need to be reset for a mechanical overspeed.

D. MS-416 has power and an AO is not needed to operate the valve. The trip linkage for MS-416 is only required to be reset after a mechanical overspeed.

Technical Reference(s): OP-009-003, Revision 304 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-EFW00 obj. 6 (As available)

WLP-OPS-PPE05 obj. 7 Question Source: Bank # 08501 Modified Bank # X (Note changes or attach parent)

New Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 4, 7, 10 55.43 Comments:

Revision 0 Facility: Waterford 3 Page 84 of 202

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 059 K3.02 Importance Rating 3.6 K/A Statement K3.02 Knowledge of the effect that a loss or malfunction of the MFW will have on the following: AFW system Proposed Question: RO 43 Rev: 0 Given:

The plant has experienced an Excess Steam Demand.

The crew has entered OP-902-004, Excess Steam Demand Recovery Procedure.

Steam Generator 1 pressure is 400 psia.

Steam Generator 2 pressure is 550 psia.

Steam Generator 1 level is 5% NR and dropping.

Steam Generator 2 level is 7% NR and dropping.

Containment Pressure is 15.2 psia and slowly rising.

Based on these conditions, the BOP operator should verify:

A. EFAS 2 initiated.

B. EFAS 1 and EFAS 2 initiated.

C. EFAS 1 initiated.

D. Neither EFAS 1 nor EFAS 2 initiated.

Revision 0 Facility: Waterford 3 Page 85 of 202

2012 NRC Exam RO Written Exam Worksheet Proposed Answer: A Explanation: (Optional)

A. CORRECT: EFAS 2 will have initiated even though S/G #2 pressure is less than 666 psia because S/G #2 pressure is greater than 123 psid above S/G #1 pressure.

EFAS 1 will not be initiated because S/G #1 pressure is less than 666 psia and is not 123 psid greater than S/G #2 pressure. Both S/G levels are less than the EFAS initiation criteria of 27.4% NR.

B. Incorrect. EFAS 1 will not be initiated because S/G #1 pressure is less than 666 psia and is not 123 psid greater than S/G #2 pressure.

C. Incorrect. EFAS 1 will not be initiated because S/G #1 pressure is less than 666 psia and is not 123 psid greater than S/G #2 pressure.

D. Incorrect. EFAS 2 will have initiated even though S/G #2 pressure is less than 666 psia because S/G #2 pressure is greater than 123 psid above S/G #1 pressure.

EFAS 1 will not be initiated because S/G #1 pressure is less than 666 psia and is not 123 psid greater than S/G #2 pressure. Both S/G levels are less than the EFAS initiation criteria of 27.4% NR.

Technical Reference(s): OP-009-003, Revision 304 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-EFW00 obj. 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 Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 4, 7 55.43 Comments:

Revision 0 Facility: Waterford 3 Page 86 of 202

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 059 A2.11 Importance Rating 3.0 K/A Statement A2.11 Ability to (a) predict the impacts of the following malfunctions or operations on the MFW; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Failure of feedwater control system Proposed Question: RO 44 Rev: 0 Given:

The plant is at 100% power.

SG ILR 1111 Steam Generator 1 Downcomer Level fails high.

The crew has entered OP-901-201, Steam Generator Level Control Malfunction.

Based on these conditions, the contingency actions for the BOP in the event of a Reactor trip would be to:

A. Close the MFRV.

Close the Startup FRV to13-21% open.

Reduce MFP A speed to minimum.

B. Close the MFRV to 13-21% open.

Close the Startup FRV.

Reduce MFP A speed to minimum.

C. Close the MFRV.

Close the Startup FRV to 13-21% open.

D. Close the MFRV to 13-21% open Close the Startup FRV.

Revision 0 Facility: Waterford 3 Page 87 of 202

2012 NRC Exam RO Written Exam Worksheet Proposed Answer: C Explanation: (Optional)

A. Incorrect. No actions are required for the Main Feed Pump speed controller because this controller will swap to auto for 5 seconds to reduce MFP A speed to minimum.

B. Incorrect. The Startup FRV must be closed to 13-21% open (not the MFRV). No actions are required for the Main Feed Pump speed controller because this controller will swap to auto for 5 seconds to reduce MFP A speed to minimum.

C. CORRECT: SG ILR 1111 Steam Generator 1 Downcomer Level failing high will cause the MFRV, Startup FRV and the Main Feed Pump speed controller for SG1 to swap to manual. On a Reactor trip, the contingency action for the BOP will be to close the MFRV, close the Startup FRV to 13-21% open. No actions are required for the Main Feed Pump speed controller because this controller will swap to auto for 5 seconds to reduce MFP A speed to minimum.

D. Incorrect. The Startup FRV must be closed to 13-21% open (not the MFRV)

Technical Reference(s): OP-902-000, Standard Post Trip Actions (Attach if not previously provided) Revision 13 (including version/revision number) OP-901-201, Revision 005 Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PPO20 obj. 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 4, 7, 10 55.43 Comments:

Revision 0 Facility: Waterford 3 Page 88 of 202

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 061 K5.02 Importance Rating 3.2 K/A Statement K5.02 Knowledge of the operational implications of the following concepts as the apply to the AFW: Decay heat sources and magnitude Proposed Question: RO 45 Rev: 0 Given:

The plant is operating at 100% power.

Emergency Feedwater Pump A is danger tagged out.

CRS directed a reactor trip due to a loss of both Main Feedwater Pumps.

The crew has entered OP-902-006, Loss of Main Feedwater Recovery.

EFW Pump AB tripped following announcement of entry into OP-902-006.

If the current plant conditions existed for 30 minutes, then the appropriate guidance in OP-902-006 for these conditions would be to ____(1)____. This action is required

____(2)____.

(1) (2)

A. reduce the number of conserve Condensate Storage Pool running Reactor inventory Coolant Pumps to two B. reduce the number of to lower the heat generation rate to less running Reactor than the EFW system heat removal Coolant Pumps to two capacity C. trip all running Reactor conserve Condensate Storage Pool Coolant Pumps inventory D. trip all running Reactor to lower the heat generation rate to less Coolant Pumps than the EFW system heat removal capacity Revision 0 Facility: Waterford 3 Page 89 of 202

2012 NRC Exam RO Written Exam Worksheet Proposed Answer: D Explanation: (Optional)

A. Incorrect. OP-902-006 directs the crew to trip all 4 RCPs within 30 minutes of the time that the capacity of only one motor driven EFW Pump is available.

B. Incorrect. OP-902-006 directs the crew to trip all 4 RCPs within 30 minutes of the time that the capacity of only one motor driven EFW Pump is available.

C. Incorrect. OP-902-006 directs the crew to trip all 4 RCPs within 30 minutes of the time that the capacity of only one motor driven EFW Pump is available. Two motor driven EFW pumps would be enough capacity to keep 2 RCPs running, therefore all 4 RCPs are not secured whenever EFW Pump AB is unavailable.

D. CORRECT: OP-902-006 directs the crew to trip all 4 RCPs within 30 minutes of the time that the capacity of only one motor driven EFW Pump is available. The capacity of one motor driven EFW pump is not enough to make up for decay heat sources and Reactor Coolant Pump heat input.

Technical Reference(s): OP-902-006, Revision 13 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PPE06 obj. 4 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 5,10 55.43 Comments:

Revision 0 Facility: Waterford 3 Page 90 of 202

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 061 A3.03 Importance Rating 3.5 K/A Statement A3.03 Ability to monitor automatic operation of the AFW, including: AFW S/G level control on automatic start Proposed Question: RO 46 Rev: 0 Given:

SG 1 pressure 980 PSIA.

SG 2 pressure 960 PSIA.

SG 1 level 51% WR level (decreasing)

SG 2 level 38% WR level (decreasing)

SG 2 EFW flow transmitter failed high Pressurizer pressure 1900 PSIA Containment pressure 15.3 PSIA Choose the proper EFW system response for SG 2:

A. SG 2 Primary FCV open to preset valve position and SG 2 Backup FCV open to 400 gpm flow value.

B. SG 2 Primary FCV and SG 2 Backup FCV are closed.

C. SG 2 Primary FCV is closed and SG 2 backup FCV is open to 175 gpm.

D. SG 2 Primary FCV open to preset valve position and SG 2 Backup FCV is closed.

Revision 0 Facility: Waterford 3 Page 91 of 202

2012 NRC Exam RO Written Exam Worksheet Proposed Answer: D Explanation: (Optional)

A. Incorrect. This would be the correct answer if the EFW flow transmitter was not failed high. At 45% WR, the Backup FCV would have opened to 400 gpm, but with the EFW flow transmitter failed high, the Backup FCV will remain closed.

B. Incorrect. At 55% WR level, the Primary FCV opens to a preset valve position. The EFW flow transmitter failing high has no effect on the Primary FCV at this level.

C. Incorrect. At 55% WR level, the Primary FCV opens to a preset valve position. The EFW flow transmitter failing high has no effect on the Primary FCV at this level. The backup FCV opening to 175 gpm would occur if the Primary FCV did not open and SG level is less than 55% WR.

D. CORRECT: At 55% WR level, the Primary FCV opens to a preset valve position.

The EFW flow transmitter failing high has no effect on the Primary FCV at this level.

At 45% WR, the Backup FCV would have opened to 400 gpm, but with the EFW flow transmitter failed high, the Backup FCV will remain closed.

Technical Reference(s): SD-EFW (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-EFW00 obj. 11 (As available)

Question Source: Bank # X 1786-A 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, 8 55.43 Comments:

Revision 0 Facility: Waterford 3 Page 92 of 202

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 062 K2.01 Importance Rating 3.0 K/A Statement K2.01 Knowledge of bus power supplies to the following: Major system loads Proposed Question: RO 47 Rev: 0 The Auxiliary Feedwater pump (AFW Pump) is powered from which of the following:

A. 1A bus B. 2A bus C. 1B bus D. 2B bus Revision 0 Facility: Waterford 3 Page 93 of 202

2012 NRC Exam RO Written Exam Worksheet Proposed Answer: C Explanation: (Optional)

A. Incorrect. The Auxiliary Feedwater (AFW) Pump is a 6.9 KV pump, but is powered from the 1B bus.

B. Incorrect. The Auxiliary Feedwater (AFW) Pump is a 6.9 KV (not 4.16 KV) motor powered from the 1B bus.

C. CORRECT: The Auxiliary Feedwater (AFW) Pump is a 6.9 KV motor powered from the 1B bus.

D. Incorrect. The Auxiliary Feedwater (AFW) Pump is a 6.9 KV (not 4.16 KV) motor powered from the 1B bus.

Technical Reference(s): OP-003-035, Revision 301 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-ED00 obj. 1 (As available)

Question Source: Bank # X 6755a Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 4 55.43 Comments:

Revision 0 Facility: Waterford 3 Page 94 of 202

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 062 A2.12 Importance Rating 3.2 K/A Statement A2.12 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: Restoration of power to a system with a fault on it Proposed Question: RO 48 Rev: 0 Given:

A loss of offsite power occurred.

The crew has entered OP-902-003, Loss of Offsite Power/Loss of Forced Circulation Recovery.

All indications for this condition were normal until:

Auxiliary Component Cooling Water Pump B tripped at the 17 second load block.

All EDG B Sequencer load block lights have extinguished.

EDG B Sequencer LOCKOUT light is LIT.

For the above indications, the ____(1)____ are de-energized and necessary actions to restore power to all Train B safety components requires racking down the ACCW Pump B breaker, then placing the Sequencer control switch on CP-1 to ____(2)____.

(1) (2)

A. all Train B sequencer RESET only loads B. all Train B sequencer RESET and then TEST loads C. Train B loads after RESET only block 17 D. Train B loads after RESET and then TEST block 17 Revision 0 Facility: Waterford 3 Page 95 of 202

2012 NRC Exam RO Written Exam Worksheet Proposed Answer: A Explanation: (Optional)

A. CORRECT: All Train B sequencer loads are deenergized on a Sequencer Lockout condition, not just the loads after the lockout. To restore power, the fault must be removed and the sequencer must be taken to Reset.

B. Incorrect. To restore power, the fault must be removed and the sequencer must be taken to Reset.

C. Incorrect. All Train B sequencer loads are deenergized on a Sequencer Lockout condition, not just the loads after the lockout.

D. Incorrect. To restore power, the fault must be removed and the sequencer must be taken to Reset.

Technical Reference(s): OP-901-311, Revision 308 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PPO30 obj. 3 (As available)

Question Source: Bank # X 08930 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 8 55.43 Comments:

Revision 0 Facility: Waterford 3 Page 96 of 202

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 063 K4.02 Importance Rating 2.9 K/A Statement K4.02 Knowledge of DC electrical system design feature(s) and/

or interlock(s) which provide for the following: Breaker interlocks, permissives, bypasses and cross-ties Proposed Question: RO 49 Rev: 0 Given:

The plant is at 100% power.

A loss of the 125 Volt TGB-DC bus has occurred.

The crew has entered OP-901-313, Loss of a 125 Volt DC Bus.

For this event, the backup over current protection for Reactor Coolant Pumps 1A and 2A are ____(1)____. All other remote manual control and automatic protection for Switchgear 1A including its associated connected components are

________(2)________.

(1) (2)

A. disabled disabled B. disabled available C. available available D. available disabled Revision 0 Facility: Waterford 3 Page 97 of 202

2012 NRC Exam RO Written Exam Worksheet Proposed Answer: D Explanation: (Optional)

A. Incorrect. Backup overcurrent protection for RCPs is the feeder breaker to the 1A(B) opening if the RCP breaker does not open within 4 seconds. The feeder breaker control power is from the A and B DC bus, which is still available.

B. Incorrect. Backup overcurrent protection for RCPs is the feeder breaker to the 1A(B) opening if the RCP breaker does not open within 4 seconds. The feeder breaker control power is from the A and B DC bus, which is still available. TGB-DC bus supplies the control power to the 1A(B) bus, which is not available.

C. Incorrect. TGB-DC bus supplies the control power to the 1A(B) bus, which is not available.

D. CORRECT: Backup overcurrent protection for RCPs is the feeder breaker to the 1A(B) opening if the RCP breaker does not open within 4 seconds. The feeder breaker control power is from the A and B DC bus, which is still available. TGB-DC bus supplies the control power to the 1A(B) bus, which is not available.

Technical Reference(s): OP-901-313, Revision 303 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-ED00 obj. 2 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 4,6 55.43 Comments:

Revision 0 Facility: Waterford 3 Page 98 of 202

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 064 K6.07 Importance Rating 2.7 K/A Statement K6.07 Knowledge of the effect of a loss or malfunction of the following will have on the ED/G system: Air receivers Proposed Question: RO 50 Rev: 0 Given:

EDG A air receiver A1 and associated air compressor are out of service with the air receiver isolated and vented to atmosphere

A relief valve lifts and local air receiver A2 pressure drops from 240 psig to 175 psig EDG A is ____(1)____. If a loss of offsite power were to occur the EDG would attempt to crank ________(2)________.

(1) (2)

A. inoperable until the A2 receiver is depleted B. Operable but until the A2 receiver is depleted degraded C. inoperable for a maximum of 25 seconds D. Operable but for a maximum of 25 seconds degraded Revision 0 Facility: Waterford 3 Page 99 of 202

2012 NRC Exam RO Written Exam Worksheet Proposed Answer: A Explanation: (Optional)

A. CORRECT: Per OP-100-014 the EDG must have at least one air receiver with a pressure of 182.5 psig to be considered operable but degraded and > 237.5 psig to be considered operable. Per SD-EDG the diesel will continue to crank until the air receiver is depleted on an emergency start if the EDG does not reach normal speed.

B. Incorrect. Wrong evaluation. The minimum pressure for operable but degraded is 179 psig per OP-100-014. Correct cranking behavior.

C. Incorrect. Correct evaluation. Wrong cranking behavior.

D. Incorrect. Wrong evaluation. . The minimum pressure for operable but degraded is 179 psig per OP-100-014. Wrong cranking behavior.

Technical Reference(s): OP-100-014, Revision 317 (Attach if not previously provided) SD-EDG (including version/revision number)

Proposed references to be provided to applicants during examination: OP-100-014 pg 20 Learning Objective: WLP-OPS-EDG00 Obj. 2,7 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 8,10 55.43 Comments:

Revision 0 Facility: Waterford 3 Page 100 of 202

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 064 A4.01 Importance Rating 4.0 K/A Statement A4.01 Ability to manually operate and/or monitor in the control room: Local and remote operation of the ED/G Proposed Question: RO 51 Rev: 0 Given:

A surveillance run is being performed for EDG A per OP-903-068, Emergency Diesel Generator and Subgroup Relay Operability Verification.

EDG A is operating in Local Control mode.

EDG A is paralleled with the grid on the 3A bus.

The speed governor control is taken to raise at the local control panel. EDG A speed will

____(1)____ and EDG A load will ________(2)________.

(1) (2)

A. increase remain the same B. remain the same increase C. increase increase D. remain the same remain the same Revision 0 Facility: Waterford 3 Page 101 of 202

2012 NRC Exam RO Written Exam Worksheet Proposed Answer: B Explanation: (Optional)

A. Incorrect. IF EDG A is paralleled to the grid, raising the speed adjust will raise load (MW) and EDG A speed will remain constant.

B. CORRECT: EDG A being in local control does not change the operation of the speed controller or indications locally or at CP-1 in the control room. IF EDG A is paralleled to the grid, raising the speed adjust will raise load (MW) and EDG A speed will remain constant.

C. Incorrect. IF EDG A is paralleled to the grid, raising the speed adjust will raise load (MW) and EDG A speed will remain constant.

D. Incorrect. IF EDG A is paralleled to the grid, raising the speed adjust will raise load (MW) and EDG A speed will remain constant. Neither EDG A load nor speed would rise if EDG A was running in Emergency Mode (Isochronous).

Technical Reference(s): OP-009-002, Revision 318 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-EDG00 obj. 8 (As available)

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

New Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 8 55.43 Comments:

Revision 0 Facility: Waterford 3 Page 102 of 202

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 073 K1.01 Importance Rating 3.6 K/A Statement K1.01 Knowledge of the physical connections and/or cause effect relationships between the PRM system and the following systems: Those systems served by PRMs Proposed Question: RO 52 Rev: 0 Given:

The plant is at 100% power.

DRY COOLING TOWER SUMP 1 ACTIVITY HI annunciator is received.

High Activity reading on DCT Sump 1 radiation monitor, PRM-IRE-6775, indicated on the RM-11.

Which of the following AUTOMATIC actions are expected as a result of these indications?

A. DCT 1 Sump Pumps align to the Waste Tanks.

B. DCT 1 Sump Pumps flow path to Circ Water isolates.

C. DCT 1 Sump Pumps trip.

D. DCT 1 Sump Pumps to the storm drains isolate.

Revision 0 Facility: Waterford 3 Page 103 of 202

2012 NRC Exam RO Written Exam Worksheet Proposed Answer: C Explanation: (Optional)

A. Incorrect. The automatic actions on a DCT Sump 1 rad monitor high alarm is that both DCT sump pumps for the respective rad monitor will trip. The DCT sumps are aligned to the waste tanks on high radiation. But, this action is performed manually.

B. Incorrect. The automatic actions on a DCT Sump 1 rad monitor high alarm is that both DCT sump pumps for the respective rad monitor will trip. The DCT Sump Pumps are normally aligned to the Circ Water System.

C. CORRECT: The automatic actions on a DCT Sump 1 rad monitor high alarm is that both DCT sump pumps for the respective rad monitor will trip.

D. Incorrect. The automatic actions on a DCT Sump 1 rad monitor high alarm is that both DCT sump pumps for the respective rad monitor will trip. The storm drains is an alternate flow path for the DCT Sump Pumps.

Technical Reference(s): OP-901-414, Revision 301 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-SP00 obj. 6 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 11 55.43 Comments:

Revision 0 Facility: Waterford 3 Page 104 of 202

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 076 A3.02 Importance Rating 3.7 K/A Statement A3.02 Ability to monitor automatic operation of the SWS, including: Emergency heat loads Proposed Question: RO 53 Rev: 0 Auxiliary Component Cooling water (ACCW) flow to the Essential Chillers will be aligned when Component Cooling Water (CCW) heat exchanger temperature reaches

___(1)____ . The ACCW valves open ________(2)________.

(1) (2)

A. 102°F after the CCW valves close B. 95°F first, then the CCW valves close C. 95°F after the CCW valves close D. 102°F first, then the CCW valves close Revision 0 Facility: Waterford 3 Page 105 of 202

2012 NRC Exam RO Written Exam Worksheet Proposed Answer: A Explanation: (Optional)

A. CORRECT: Auxiliary Component Cooling water (ACCW) flow to the Essential Chillers will be automatically be aligned when Component Cooling Water (CCW) heat exchanger outlet temperature reaches 102°F. The ACCW valves must see the CCW valves closed before the ACCW valves will open.

B. Incorrect. Auxiliary Component Cooling water (ACCW) flow to the Essential Chillers will be automatically be aligned when Component Cooling Water (CCW) heat exchanger outlet temperature reaches 102°F (95°F is the reset value that will swap back to CCW). The ACCW valves must see the CCW valves closed before the ACCW valves will open.

C. Incorrect. Auxiliary Component Cooling water (ACCW) flow to the Essential Chillers will be automatically be aligned when Component Cooling Water (CCW) heat exchanger outlet temperature reaches 102°F (95°F is the reset value that will swap back to CCW)

D. Incorrect. The ACCW valves must see the CCW valves closed before the ACCW valves will open.

Technical Reference(s): SD-CC (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-CC00 obj. 3 (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 Comments:

Revision 0 Facility: Waterford 3 Page 106 of 202

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 078 G2.1.30 Importance Rating 4.4 K/A Statement 2.1.30 Ability to locate and operate components, including local controls.

Proposed Question: RO 54 Rev: 0 Given:

The plant has experienced an instrument air leak.

OP-901-511, Instrument Air Malfunction has been entered.

The instrument air leak has been isolated and Instrument Air pressure is being restored.

IA Dryer Bypass, IA-123, is open.

IA Dryer Bypass, IA-123, ____(1)____ when the reset value is reached. Position indication for IA Dryer Bypass, IA-123, can be monitored from ________(2)________.

(1) (2)

A. requires manual control room annunciator and locally at the closure instrument air dryers.

B. automatically control room annunciator and locally at the closes instrument air dryers.

C. requires manual locally at the instrument air dryers only closure D. automatically locally at the instrument air dryers only closes Revision 0 Facility: Waterford 3 Page 107 of 202

2012 NRC Exam RO Written Exam Worksheet Proposed Answer: A Explanation: (Optional)

A. CORRECT: IA-123 automatically opens when IA pressure lowers to 95 psig but must be manually closed when Instrument air pressure is restored and the reset value for closing IA-123 is reached. Position indication is available at the IA dryer (control switch) and an annunciator in the Control Room which is illuminated when IA-123 is open and clear when IA-123 is closed.

B. Incorrect. IA-123 must be manually closed (does not auto close) when Instrument air pressure is restored and the reset value for closing IA-123 is reached. Position indication is available at the IA dryer (control switch) along with the annunciator in the Control Room.

C. Incorrect. IA-123 must be manually closed when the reset vale for IA pressure is reached. Position indication is available at the IA dryer (control switch) along with the annunciator in the Control Room D. Incorrect. IA-123 must be manually closed (does not auto close) when Instrument air pressure is restored and the reset value for closing IA-123 is reached. Position indication is available at the IA dryer (control switch) along with the annunciator in the Control Room.

Technical Reference(s): SD-AIR (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PP050 obj. 3 (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 3, 10 55.43 Comments:

Revision 0 Facility: Waterford 3 Page 108 of 202

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 103 A1.01 Importance Rating 3.7 K/A Statement A1.01 Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the containment system controls including: Containment pressure, temperature, and humidity Proposed Question: RO 55 Rev: 0 If Containment to Annulus Differential Pressure reaches ____(1)____ INWD, Containment Vacuum Relief Valves CVR-101 and CVR-201 OPEN.

CVR-101 and CVR-201 can be realigned when DP ______(2)_____ to the reset value.

(1) (2)

A. 8.5 rises B. 5.5 rises C. 8.5 lowers D. 5.5 lowers Revision 0 Facility: Waterford 3 Page 109 of 202

2012 NRC Exam RO Written Exam Worksheet Proposed Answer: C Explanation: (Optional)

A. Incorrect. CVR-101 and CVR-201 open automatically when annulus pressure is greater than containment pressure by 8.5 INWD but D/P must lower (not rise) to realign the CVR valves.

B. Incorrect. CVR-101 and CVR-201 open automatically when annulus pressure is greater than containment pressure by 8.5 INWD (not 5.5 INWD). 5.5 INWD is the annulus to ambient D/P used when cycling CVR valves to raise containment pressure.

C. CORRECT: CVR-101 and CVR-201 open automatically when annulus pressure is greater than containment pressure by 8.5 INWD. CVR-101 and CVR-201 can be manually closed when containment to annulus D/P lowers to 6.9 INWD.

D. Incorrect. CVR-101 and CVR-201 open automatically when annulus pressure is greater than containment pressure by 8.5 INWD (not 5.5 INWD). 5.5 INWD is the annulus to ambient D/P used when cycling CVR valves to raise containment pressure.

Technical Reference(s): OP-008-005, Revision 304 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-CB00 obj. 2 (As available)

Question Source: Bank # X 08262a Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 8, 9 55.43 Comments:

Revision 0 Facility: Waterford 3 Page 110 of 202

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 2 K/A # 002 G2.4.11 Importance Rating 4.0 K/A Statement 2.4.11 Knowledge of abnormal condition procedures.

Proposed Question: RO 56 Rev: 0 Given:

The plant is operating at 100% power.

OP-901-111, Reactor Coolant System Leak, has been implemented due to a 2 gpm increase in leakage.

Using a contact pyrometer, the RCA watch has located increased temperature on pipe 7BM1-246 at the Reactor Drain Tank.

Which ONE of the following valves is a possible RCS leakage source?

A. RC-510A, RCP 2A Controlled Bleedoff to Reactor Drain Tank B. RC-601, RCP Controlled Bleedoff to Reactor Drain Tank Isol C. RC-204, Reactor Coolant Hot Leg 2 Drain D. RC-213, Reactor Coolant Loop 2A Charging Line Drain Isol Revision 0 Facility: Waterford 3 Page 111 of 202

2012 NRC Exam RO Written Exam Worksheet Proposed Answer: D Explanation: (Optional)

A. Incorrect. RC-510A, RCP 2A Controlled Bleedoff to Reactor Drain Tank is an input to the RDT, but is not an input to pipe 7BM1-246.

B. Incorrect. RC-601, RCP Controlled Bleedoff to Reactor Drain Tank Isol is an input to the RDT, but is not an input to pipe 7BM1-246.

C. Incorrect. RC-204, Reactor Coolant Hot Leg 2 Drain is an input to the RDT, but is not an input to pipe 7BM1-246.

D. CORRECT: RC-213, Reactor Coolant Loop 2A Charging Line Drain Isol is an input to pipe 7BM1-246. This can be verified by using attachment 1 of OP-901-111.

Technical Reference(s): OP-901-111, Revision 301.

(Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: OP-901-111 (all of attachment 1)

Learning Objective: WLP-OPS-PP010 obj. 3 (As available)

Question Source: Bank # X 07605 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 3, 10 55.43 Comments:

Revision 0 Facility: Waterford 3 Page 112 of 202

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 2 K/A # 011 K2.01 Importance Rating 3.1 K/A Statement K2.01 Knowledge of bus power supplies to the following: Charging pumps Proposed Question: RO 57 Rev: 0 Given:

AB busses are powered from Train B Charging Pump AB is powered from ____________. With the present bus alignment, Charging Pump AB can be aligned to automatically start in place of Charging Pump

____________ for a Safety Injection Actuation Signal.

(1) (2)

A. AB3 B only B. AB31 B only C. AB3 A or B D. AB31 A or B Revision 0 Facility: Waterford 3 Page 113 of 202

2012 NRC Exam RO Written Exam Worksheet Proposed Answer: B Explanation: (Optional)

A. Incorrect. Wrong bus. Correct pump.

B. CORRECT: The listed charging pump is powered from 480 VAC bus AB31. The charging pump can be aligned to replace Charging Pump B only for an automatic start on SIAS. The pump can still be started manually in place of Charging Pump A after verifying sequencer is timed out; however to get an automatic start, the bus alignment has to the same as the pump being replaced.

C. Incorrect. Wrong bus. Wrong pump combination.

D. Incorrect. Correct bus. Wrong pump combination.

Technical Reference(s): OP-500-007, Revision 020 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-CVC00 Obj. 5 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 Comments:

Revision 0 Facility: Waterford 3 Page 114 of 202

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 2 K/A # 014 K5.02 Importance Rating 2.8 K/A Statement K5.02 Knowledge of the operational implications of the following concepts as they apply to the RPIS: RPIS independent of demand position Proposed Question: RO 58 Rev: 0 Given:

The plant is performing a plant startup and is presently at 80% power.

Group P CEAs are being withdrawn for ASI control.

The Group select switch is in the P position.

CEA #25 in Group P becomes mechanically bound.

If the CEA Manual Shim switch is taken to WITHDRAW, PMC position indication for CEA 25 on CP-2 will be ____(1)____ for CEA 25. The CEA position indications on the CEACs will be ______(2)____ for CEA 25.

(1) (2)

A. rising stationary B. stationary stationary C.

rising rising D. stationary rising Revision 0 Facility: Waterford 3 Page 115 of 202

2012 NRC Exam RO Written Exam Worksheet Proposed Answer: A Explanation: (Optional)

A. CORRECT: The PMC pulse counters will be rising as long as the shim switch is taken to withdraw. The CEA position used in the CEACs work on reed switch position indicators and will see actual CEA position. Therefore, the CEACs will show a stationary position for CEAs.

B. Incorrect. The PMC pulse counters will be rising as long as the shim switch is taken to withdraw.

C. Incorrect. The CEA position used in the CEACs work on reed switch position indicators and will see actual CEA position. Therefore, the CEACs will show a stationary position for CEAs.

D. Incorrect. The PMC pulse counters will be rising as long as the shim switch is taken to withdraw. The CEA position used in the CEACs work on reed switch position indicators and will see actual CEA position. Therefore, the CEACs will show a stationary position for CEAs.

Technical Reference(s): SD-CED (Attach if not previously provided) SD-PMC (including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-CED00 obj. 17 (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 6 55.43 Comments:

Revision 0 Facility: Waterford 3 Page 116 of 202

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 2 K/A # 016 K3.02 Importance Rating 3.4 K/A Statement K3.02 Knowledge of the effect that a loss or malfunction of the NNIS will have on the following: PZR LCS Proposed Question: RO 59 Rev: 0 Given:

The plant is operating at 100% power.

RRS Cabinet LOOP SELECTOR in the RRS Cabinets 1 and 2 is in the BOTH position.

RCS Temperature Loop 1 Cold Leg (RC-ITI-0111-Y) indicates failed low.

Based on this instrument failure, the ATC operator will expect letdown flow to

____(1)____. The preferred method of selecting the non-faulted Tcold instrument is to operate the TCOLD LOOP 1 selector switch located in the ______(2)_____ .

(1) (2)

A. lower RRS local cabinets B. lower rear of CP-2 C. rise RRS local cabinets D. rise rear of CP-2 Revision 0 Facility: Waterford 3 Page 117 of 202

2012 NRC Exam RO Written Exam Worksheet Proposed Answer: D Explanation: (Optional)

A. Incorrect. RCS Temperature Loop 1 Cold Leg failing low will cause the setpoint for the PLCS to lower. Actual level will be greater than setpoint, therefore letdown flow will rise (not lower). The loop selector switch in RRS cabinets is used to swap to an operable Thot instrument and can be used to swap to an operable Tcold instrument.

Although, Operating the TCOLD LOOP selector switch behind CP-2 is the preferred method of selecting the operable Tcold instrument per OP-901-110, section E2.

B. Incorrect. RCS Temperature Loop 1 Cold Leg failing low will cause the setpoint for the PLCS to lower. Actual level will be greater than setpoint, therefore letdown flow will rise (not lower).

C. Incorrect. The loop selector switch in RRS cabinets is used to swap to an operable Thot instrument and can be used to swap to an operable Tcold instrument. Although, Operating the TCOLD LOOP selector switch behind CP-2 is the preferred method of selecting the operable Tcold instrument per OP-901-110, section E2.

D. CORRECT: RCS Temperature Loop 1 Cold Leg failing low will cause the setpoint for the PLCS to lower. Actual level will be greater than setpoint, therefore letdown flow will rise. Operating the TCOLD LOOP selector switch behind CP-2 is the preferred method of selecting the operable Tcold instrument per OP-901-110, section E2.

Technical Reference(s): OP-901-110, Revision 6 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-RR00 obj. 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 Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7, 10 55.43 Comments:

Revision 0 Facility: Waterford 3 Page 118 of 202

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 2 K/A # 075 A2.03 Importance Rating 2.5 K/A Statement Ability to (a) predict the impacts of the following malfunctions or operations on the circulating water system; and (b) based on those predictions, use procedures to correct, control or mitigate the consequences of those malfunctions or operations: Safety features and relationship between condenser vacuum, turbine trip and steam dump.

Proposed Question: RO 60 Rev: 0 Given:

The plant was initially at 100% power.

Circ Water Pumps A, B and D are operating.

Circ Water Pump A trips.

Condenser vacuum is 9 inches Hg and stable.

The Steam Bypass control system is ____(1)____

The crew will verify ______(2)_____ .

(1) (2)

A. unavailable a Main Feedwater pump is available B. unavailable Emergency Feedwater is available C. available a Main Feedwater pump is available D. available Emergency Feedwater is available Revision 0 Facility: Waterford 3 Page 119 of 202

2012 NRC Exam RO Written Exam Worksheet Proposed Answer: D Explanation: (Optional)

A. Incorrect. Steam bypass is still available. Steam bypass is interlocked to close at 3.4 in Hg. The crew will verify EFW is available because both Main Feed Pumps tripped at 14in Hg.

B. Incorrect. Steam bypass is still available. Steam bypass is interlocked to close at 3.4 in Hg.

C. Incorrect. Steam bypass is interlocked to close at 3.4 in Hg. The crew will verify EFW is available because both Main Feed Pumps tripped at 14in Hg.

D. CORRECT: Steam bypass is still available. Steam bypass is interlocked to close at 3.4 in Hg. The crew will verify EFW is available because both Main Feed Pumps tripped at 14in Hg.

Technical Reference(s): OP-901-220, Revision 302 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PP020 objective 3 (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 Comments:

Revision 0 Facility: Waterford 3 Page 120 of 202

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 2 K/A # 029 K4.03 Importance Rating K/A Statement K3.02 Knowledge of design feature(s) and/or interlock(s) which provide for:

Automatic purge isolation Proposed Question: RO 61 Rev: 0 Given:

The plant is at 100% power.

Containment Purge is in progress per OP-002-010, RAB HVAC and Containment Purge.

RAB Vent Mode selector switch is in the CONTAINMENT PURGE position.

Which one of the following will automatically close Containment Purge Inlet inside Annulus, CAP-103.

A. RAB Normal exhaust flow decreases to 90,000 scfm.

B. Hi-Hi alarm on Plant Stack B gas channel (PRM-IRE-0100.2)

C. Containment Pressure lowers to -6.0 INWG below atmospheric pressure.

D. CAP exhaust valves CAP-203, CAP-204 and CAP-205 reach greater than the 52° open position.

Revision 0 Facility: Waterford 3 Page 121 of 202

2012 NRC Exam RO Written Exam Worksheet Proposed Answer: B Explanation: (Optional)

A. Incorrect. RAB Normal exhaust flow decreasing to 84,000 scfm (not 90,000 scfm) will isolate CAP-103.

B. CORRECT: CAP-103 automatically closes on a hi-hi radiation signal from the Plant Stack Rad Monitor (PRM-IRE-0100.2)

C. Incorrect. Containment Pressure lowering to -8.4 INWG (not -6.0 INWG) below atmospheric pressure will close CAP-103.

D. Incorrect. CAP exhaust valves CAP-203, CAP-204 and CAP-205 must be full open for CAP-103 to remain open. The 52° open position is the normal position for these valves at 100% power.

Technical Reference(s): OP-002-010, Revision 304 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-HVR00 obj. 3 (As available)

Question Source: Bank # X 4118-A Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 9,11 55.43 Comments:

Revision 0 Facility: Waterford 3 Page 122 of 202

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 2 K/A # 034 A4.02 Importance Rating 3.5 K/A Statement A4.02 Ability to manually operate and/or monitor in the control room: Neutron levels Proposed Question: RO 62 Rev: 0 Given:

The plant is in a refueling outage.

The refueling group is preparing to withdraw the first fuel assembly from the reactor vessel.

The Source Range Neutron Flux Monitors are verified operable by ensuring at least

____(1)___ channel(s) operable with continuous visible indication in the control room and ___(2)___ channel(s) operable with audible indication in the containment and the control room .

(1) (2)

A. two one B. one one C. two two D. one two Revision 0 Facility: Waterford 3 Page 123 of 202

2012 NRC Exam RO Written Exam Worksheet Proposed Answer: A Explanation: (Optional)

A. CORRECT: Source range neutron flux monitors are verified operable by ensuring at least 2 channels operable with continuous visible indication in the control room and 1 channel operable with audible indication in the containment and control room. (TS 3.9.2)

B. Incorrect. Source range neutron flux monitors are verified operable by ensuring at least 2 channels operable with continuous visible indication in the control room.

C. Incorrect. TS 3.9.2 requires only 1 channel operable with audible indication in the containment and control room.

D. Incorrect. Source range neutron flux monitors are verified operable by ensuring at least 2 channels operable with continuous visible indication in the control room and 1 channel operable with audible indication in the containment and control room. (TS 3.9.2)

Technical Reference(s): TS 3.9.2 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-REQ04 obj. 2 (As available)

Question Source: Bank # 2927-A 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 2, 6 55.43 Comments:

Revision 0 Facility: Waterford 3 Page 124 of 202

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 2 K/A # 041 A3.02 Importance Rating 3.3 K/A Statement A3.02 Ability to monitor automatic operation of the SDS, including: RCS pressure, RCS temperature, and reactor power Proposed Question: RO 63 Rev: 0 Given:

The plant is at 100% power.

RCS Tavg is 573°F.

RCS Pressure is 2250 PSIA.

Reactor cutback is out of service.

Reactor trip on Turbine trip is enabled.

A Turbine trip occurs.

The response of the steam bypass control system immediately following generation of the reactor trip signal will be:

A. All valves will quick open.

B. Only valves 1 through 5 will quick open.

C. Only valves 1 through 3 will quick open.

D. No valves will quick open.

Revision 0 Facility: Waterford 3 Page 125 of 202

2012 NRC Exam RO Written Exam Worksheet Proposed Answer: B Explanation: (Optional)

A. Incorrect. All SBCVs do not quick open on a reactor trip. SBC valve #6 quick open is blocked any time there is a reactor trip.

B. CORRECT: SBC valves 1-5 will quick open on a reactor trip with plant power at 100% power. SBC valve #6 quick open is blocked any time there is a reactor trip.

C. Incorrect. The quick open mode for the SBC system operates in two groups of 3 valves. Only one group of SBCVs quick opening will not handle a trip from 100%

power.

D. Incorrect. RCS Tave less than 561°F would block all SBCVs from quick opening.

The initial conditions in this event has Tave at 573°F.

Technical Reference(s): SD-SBC (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-SBC00 obj. 5 (As available)

Question Source: Bank # X 07938 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 7 55.43 Comments:

Revision 0 Facility: Waterford 3 Page 126 of 202

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 2 K/A # 045 K1.06 Importance Rating 2.6 K/A Statement K1.06 Knowledge of the physical connections and/or cause-effect relationships between the MT/G system and the following systems: RCS, during steam valve test Proposed Question: RO 64 Rev: 0 Given:

The plant is reducing turbine load to less than 1109 MWe to perform OP-903-007, Turbine Inlet Valve Cycling Test.

Per OP-903-007, Turbine Inlet Valve Cycling Test, RCS Tcold is maintained 543°F to 546°F by ____(1)____ the ______(2)_____ in manual control during the turbine load reduction.

(1) (2)

A. closing atmospheric dump valve B. opening steam bypass control valve C. opening atmospheric dump valve D. closing steam bypass control valve Revision 0 Facility: Waterford 3 Page 127 of 202

2012 NRC Exam RO Written Exam Worksheet Proposed Answer: B Explanation: (Optional)

A. Incorrect. Per OP-903-007, Turbine Inlet Valve Cycling Test, RCS Tcold is maintained 543°F to 546°F by operating a SBCV in manual. The SBCV will have to be opened (not closed) to maintain RCS temperature if load is being taken off of the turbine.

B. CORRECT: Per OP-903-007, Turbine Inlet Valve Cycling Test, RCS Tcold is maintained 543°F to 546°F by operating a SBCV in manual. The SBCV will have to be opened to maintain RCS temperature if load is being taken off of the turbine.

C. Incorrect. Per OP-903-007, Turbine Inlet Valve Cycling Test, RCS Tcold is maintained 543°F to 546°F by operating a SBCV in manual.

D. Incorrect. The SBCV will have to be opened (not closed) to maintain RCS temperature if load is being taken off of the turbine.

Technical Reference(s): OP-903-007, Revision 14 (Attach if not previously provided)

(including version/revision number)

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

Question Source: Bank #

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 5, 10 55.43 Comments:

Revision 0 Facility: Waterford 3 Page 128 of 202

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 2 K/A # 086 A1.05 Importance Rating 2.9 K/A Statement A1.05 Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the Fire Protection System controls including: FPS lineups Proposed Question: RO 65 Rev: 0 Given:

The crew is responding to an inadvertent CIAS.

The plant is in MODE 4 in preparation for a refueling outage.

The Airborne Radioactivity Removal System (ARRS) B charcoal filter catches fire.

The pull stations for E-13B (ARRS B) and FPM-2 (Reactor Bldg Fire Main Header B) have been operated.

The fire brigade leader reports that there is no firewater flow to ARRS B unit and the unit continues to burn.

To comply with OP-009-004, Fire Protection the BOP should:

A. Operate the pull station for FPM-1, Reactor Bldg Fire Main Header A.

B. Place the control switch for FP-601B, Reactor Bldg Fire Mn Hdr B FPM-2 CNTMT Isol, to Close and then to Open.

C. Direct the Fire Brigade Leader to locally actuate the deluge system for the B Airborne Radioactivity Removal System (ARRS) unit.

D. Direct the Fire Brigade Leader to spray down the B Airborne Radioactivity Removal System (ARRS) charcoal unit with the local fire hoses.

Revision 0 Facility: Waterford 3 Page 129 of 202

2012 NRC Exam RO Written Exam Worksheet Proposed Answer: B Explanation: (Optional)

A. Incorrect. Actuating FPM-1 will not help because FP-601A is closed and the fire is on the B ARRS unit.

B. CORRECT: The inadvertent CIAS closed FP-601B, preventing fire protection water from reaching containment. OP-009-004 has the operator verify FP-601B open after the pull stations are actuated. FP-601B control switch must be taken to close then open to reset a CIAS.

C. Incorrect. The deluge system for the B Airborne Radioactivity Removal System (ARRS) unit can only be actuated from the control room. Plus, FP-601A and FP-601B are closed due to the inadvertent CIAS.

D. Incorrect. Directing the Fire Brigade Leader to spray down the B Airborne Radioactivity Removal System (ARRS) charcoal unit with the local fire hoses will not work because FP-601A and FP-601B are closed due to the inadvertent CIAS.

Technical Reference(s): OP-009-004, Revision 313 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-FP00 obj. 1 (As available)

Question Source: Bank # X 07835 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 8, 9 55.43 Comments:

Revision 0 Facility: Waterford 3 Page 130 of 202

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 3 Group # 1 K/A # G2.1.1 Importance Rating 3.8 K/A Statement Conduct of Operations: Knowledge of conduct of operations requirements.

Proposed Question: RO 66 Rev: 0 Per EN-OP-115, Conduct of Operations, if an operating parameter exceeds any of the reactor protection set points and an automatic shutdown does not occur, the licensed operator is required to:

A. take action to restore the parameter within limit, if not successful manually trip the reactor..

B. get Control Room Supervisor concurrence and then manually trip the reactor.

C. report tripping the reactor while taking action to manually trip the reactor.

D. get a peer check on the parameter and then manually trip the reactor.

Revision 0 Facility: Waterford 3 Page 131 of 202

2012 NRC Exam RO Written Exam Worksheet Proposed Answer: C Explanation: (Optional)

A. Incorrect. Per EN-OP-115, this is not the correct action to take when operating parameters exceed any of the reactor protection set points and an automatic shutdown does not occur B. Incorrect. Per EN-OP-115, this is not the correct action to take when operating parameters exceed any of the reactor protection set points and an automatic shutdown does not occur C. CORRECT: Per EN-OP-115, step 5.2, Licensed operators SHALL immediately insert a manual scram whenever Operating parameters exceed any of the reactor protection set points and an automatic shutdown does not occur.

D. Incorrect. Per EN-OP-115, this is not the correct action to take when operating parameters exceed any of the reactor protection set points and an automatic shutdown does not occur Technical Reference(s): EN-OP-115, Revision 12 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PPA00 Obj. 2 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Comments:

Revision 0 Facility: Waterford 3 Page 132 of 202

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 3 Group # 1 K/A # G2.1.3 Importance Rating 3.7 K/A Statement Conduct of Operations: Knowledge of shift or short-term relief turnover practices.

Proposed Question: RO 67 Rev: 0 Per EN-OP-115, Conduct of Operations, if a control room operator needs to be relieved for a short term during their shift to obtain a medical physical, permission must be granted from the ____(1)____ and it requires a ______(2)_____ prior to relieving the watch.

(1) (2)

A. FSS review of the EOS index B. SM or CRS verbal turnover C. FSS verbal turnover D. SM or CRS review of the EOS index Revision 0 Facility: Waterford 3 Page 133 of 202

2012 NRC Exam RO Written Exam Worksheet Proposed Answer: B Explanation: (Optional)

A. Incorrect. Per EN-OP-115-3 step 5.7.2, the SM or CRS must give permission to be relieved on shift. A verbal turnover is required to be relieved of the watch B. CORRECT: Per EN-OP-115-3 step 5.7.2, the SM or CRS must give permission to be relieved on shift. A verbal turnover is also required to be relieved of the watch.

C. Incorrect. Per EN-OP-115-3 step 5.7.2, the SM or CRS must give permission to be relieved on shift.

D. Incorrect. Per EN-OP-115, a verbal turnover is required to be relieved of the watch.

Technical Reference(s): EN-OP-115-03, Revision 12 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PPA00 Obj. 2 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Comments:

Revision 0 Facility: Waterford 3 Page 134 of 202

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 3 Group # 1 K/A # G2.1.44 Importance Rating 3.9 K/A Statement Conduct of Operations: Knowledge of RO duties in the control room during fuel handling, such as responding to alarms from the fuel handling area, communication with the fuel storage facility, systems operated from the control room in support of fueling operations, and supporting instrumentation.

Proposed Question: RO 68 Rev: 0 Given:

The plant is in a refueling outage.

Core off load is in progress.

The Refueling SRO has communicated that a spent fuel assembly has been dropped from the Spent Fuel Handling Machine.

The crew has entered OP-901-405, Fuel Handling Incident.

The ATC reports that FHB Isolation B (300.1) radiation monitor is in alarm.

Based on these conditions, the operator will verify that:

A. A Fuel Handling Accident Signal has occurred and that Fuel Handling Building Emergency Filtration Unit B has started.

B. Upon a second FHB radiation area monitor going into alarm that a Fuel Handling Accident Signal has occurred.

C. Upon any of the two FHB Exhaust PIG (5107A or B) radiation monitors going into alarm that a Fuel Handling Accident Signal has occurred.

D. A Fuel Handling Accident Signal has occurred and that both Fuel Handling Building Emergency Filtration Units have started.

Revision 0 Facility: Waterford 3 Page 135 of 202

2012 NRC Exam RO Written Exam Worksheet Proposed Answer: A Explanation: (Optional)

A. CORRECT: Any of the two Train B FHB area monitors (300.1, 300.3) will generate a Fuel Handling Accident signal and start FHB B Emergency Filtration unit.

B. Incorrect. Any of the four FHB area monitors (300.1, 300.2, 300.3 or 300.4) will generate a train related Fuel Handling Accident signal.

C. Incorrect. FHB Exhaust PIG (5107A or B) radiation monitors going into alarm does not cause a Fuel Handling Accident signal.

D. . Incorrect. A Fuel Handling Accident B signal has occurred but both FHB emergency filtration units do not start (only B).

Technical Reference(s): OP-901-404, Revision 002 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-HVF00 obj. 3 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7,11 55.43 Comments:

Revision 0 Facility: Waterford 3 Page 136 of 202

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 3 Group # 2 K/A # G2.2.6 Importance Rating 3.0 K/A Statement Equipment Control: Knowledge of the process for making changes to procedures.

Proposed Question: RO 69 Rev: 0 Prior to forwarding a Procedure Improvement Request (PIR) to the Operations Procedure Administrative Group (OPAG), the PIR must be verified to be complete, accurate and proper justification is provided by obtaining a review from ____(1)____ and

______(2)_____.

(1) (2)

A. the requestors immediate supervisor an independent SRO B. another licensed operator an independent SRO C. another licensed operator a 50.59 qualified reviewer D. the requestors immediate supervisor a 50.59 qualified reviewer Revision 0 Facility: Waterford 3 Page 137 of 202

2012 NRC Exam RO Written Exam Worksheet Proposed Answer: A Explanation: (Optional)

A. CORRECT: Per OI-019-000, Operations Procedure Administration (steps 5.1.2 and 5.1.3) directs the PIR to have a review from the Immediate supervisor and an Independent SRO before forwarding the request to OPAG.

B. Incorrect. Per OI-019-000, Operations Procedure Administration (steps 5.1.2 and 5.1.3) directs the PIR to have a review from the Immediate supervisor and an Independent SRO before forwarding the request to OPAG.

C. Incorrect. Per OI-019-000, Operations Procedure Administration (steps 5.1.2 and 5.1.3) directs the PIR to have a review from the Immediate supervisor and an Independent SRO before forwarding the request to OPAG.

D. Incorrect. Per OI-019-000, Operations Procedure Administration (steps 5.1.2 and 5.1.3) directs the PIR to have a review from the Immediate supervisor and an Independent SRO before forwarding the request to OPAG.

Technical Reference(s): OI-019-000, Revision 303 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PPA00 obj. 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 Comments:

Revision 0 Facility: Waterford 3 Page 138 of 202

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 3 Group # 2 K/A # G2.2.13 Importance Rating 4.1 K/A Statement Equipment Control: Knowledge of tagging and clearance procedures.

Proposed Question: RO 70 Rev: 0 Which of the following components qualifies as a NO TAG component on a Danger Tagout?

A. A handwheel on a Motor Operated Valve being used as an isolation boundary.

B. A drain valve that is going to be removed by maintenance.

C. A breaker that maintenance personnel will need to operate while signed on to the tagout.

D. A pump discharge valve that is used as the inside double valve isolation on a high energy system.

Revision 0 Facility: Waterford 3 Page 139 of 202

2012 NRC Exam RO Written Exam Worksheet Proposed Answer: B Explanation: (Optional)

A. Incorrect. Not allowed as a no tag per EN-OP-102.

B. CORRECT: EN-OP-102 identifies a purpose of a no tag is to identify components that cannot be tagged but require positioning when hanging a tagout. Removing a drain valve meets this purpose.

C. Incorrect. Not allowed as a no tag per EN-OP-102.

D. Incorrect. Not allowed as a no tag per EN-OP-102.

Technical Reference(s): EN-OP-102, Revision 15 (Attach if not previously provided)

(including version/revision number)

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

Question Source: Bank # X 08462a 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 Comments:

Revision 0 Facility: Waterford 3 Page 140 of 202

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 3 Group # 3 K/A # G2.2.44 Importance Rating 4.2 K/A Statement Equipment Control: Ability to interpret control room indications to verify the status and operation of a system, and understand how operator actions and directives affect plant and system conditions.

Proposed Question: RO 71 Rev: 0 Given:

The plant is at 100% power.

FW IFR1121, Steam Generator 2 Feedwater flow instrument has failed low.

The crew has entered OP-901-201, Steam Generator Level Control Malfunction.

The BOP placed Main Feed Regulating Valve (MFRV) 2 in MANUAL in response to the failure If Steam Generator 2 level is allowed to rise to the High Level Override setpoint of

____(1)____ the FWCS 2 response would be to close ______(2)_____ .

(1) (2)

A. 87.7% NR Main Feed Regulating Valve 2 and Startup Feed Regulating Valve 2 B. 87.7% NR Startup Feed Regulating Valve 2 C. 81% NR Main Feed Regulating Valve 2 and Startup Feed Regulating Valve 2 D. 81% NR Startup Feed Regulating Valve 2 Revision 0 Facility: Waterford 3 Page 141 of 202

2012 NRC Exam RO Written Exam Worksheet Proposed Answer: D Explanation: (Optional)

A. Incorrect. The HLO occurs at 81% NR. (87.7 NR is the SG level that will cause a Reactor trip to occur.) The High Level Override closes both the MFRV and SUFRV for the SG with the High level. In this instance, MFRV 2 is in manual, therefore only the SUFRV will close on a HLO.

B. Incorrect. The HLO occurs at 81% NR. (87.7 NR is the SG level that will cause a Reactor trip to occur.)

C. Incorrect. The High Level Override closes both the MFRV and SUFRV for the SG with the High level. In this instance, MFRV 2 is in manual, therefore only the SUFRV will close on a HLO.

D. CORRECT: The HLO occurs at 81% NR. The High Level Override closes both the MFRV and SUFRV for the SG with the High level. In this instance, MFRV 2 is in manual, therefore only the SUFRV will close on a HLO.

Technical Reference(s): OP-901-201, Revision 5 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-FWC00 obj. 4 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 4,7 55.43 Comments:

Revision 0 Facility: Waterford 3 Page 142 of 202

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 3 Group # 3 K/A # G2.3.4 Importance Rating 3.2 K/A Statement Radiation Control: Knowledge of radiation exposure limits under normal or emergency conditions.

Proposed Question: RO 72 Rev: 0 The routine administrative limit for TEDE with no undocumented quarters is ____(1)____

mrem/year. The Emergency Exposure TEDE guideline for lifesaving activities is

______(2)______ Rem.

(1) (2)

A. 2000 75 B. 4500 75 C. 2000 25 D. 4500 25 Revision 0 Facility: Waterford 3 Page 143 of 202

2012 NRC Exam RO Written Exam Worksheet Proposed Answer: C Explanation: (Optional)

A. Incorrect. Correct routine admin limit. 75 Rem is the lens of the eye limit for lifesaving activities.

B. Incorrect. 4500 mrem/yr is the maximum allow administrative limit. Wrong emergency exposure limit.

C. CORRECT: Per EP-002-030 step 5.2.2, Emergency team members chosen to perform life saving activities shall not exceed 25 Rem. EN-RP-201 states the routine limit with no undocumented quarters is 2000 mrem/yr.

D. Incorrect. Wrong routine admin limit. Correct emergency exposure limit.

Technical Reference(s): EP-002-030, Rev. 10 (Attach if not previously provided) EN-RP-201, Rev. 3 (including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-EP02 obj. 8 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 9,12 55.43 Comments:

Revision 0 Facility: Waterford 3 Page 144 of 202

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 3 Group # 4 K/A # G2.3.13 Importance Rating 3.4 K/A Statement 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 73 Rev: 0 Two areas classified in Mode 1 as Very High Radiation Areas, where access is forbidden, are:

A. at the SDC Isolations (SI-401A and 405A) and the Reactor Cavity.

B. at the SDC Isolations (SI-401A and 405A) and outside the Regen HX Cubicle.

C. the Reactor Vessel Annulus and the Reactor Cavity.

D. the Reactor Vessel Annulus and the outside the Regen HX Cubicle.

Revision 0 Facility: Waterford 3 Page 145 of 202

2012 NRC Exam RO Written Exam Worksheet Proposed Answer: C Explanation: (Optional)

A. Incorrect. The SDC valves are not a forbidden area.

B. Incorrect. The SDC valves are not a forbidden area and outside the Regen HX cubicle is not a forbidden area.

C. CORRECT: HP-001-213 step 5.2.2 states that the following areas have been designated as Very High Radiation Areas. Entries into these areas are forbidden when the reactor is in mode 1: (Hot and Cold Leg D ring wall penetrations, Reactor Vessel Annulus, and the Reactor Cavity.)

D. Incorrect. Outside the Regen HX is not a forbidden area.

Technical Reference(s): HP-001-213, Revision 303 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PPA00 obj. 3 (As available)

Question Source: Bank # X 3445-B 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 12 55.43 Comments:

Revision 0 Facility: Waterford 3 Page 146 of 202

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 3 Group # 4 K/A # G2.4.5 Importance Rating 3.7 K/A Statement Emergency Procedures / Plan: Knowledge of the organization of the operating procedures network for normal, abnormal, and emergency evolutions.

Proposed Question: RO 74 Rev: 0 The network of procedures used to perform operations refueling tasks is primarily located in the _____________ series procedures.

A. OI B. OP-010 C. OP-100 D. UNT Revision 0 Facility: Waterford 3 Page 147 of 202

2012 NRC Exam RO Written Exam Worksheet Proposed Answer: B Explanation: (Optional)

A. Incorrect. The OI series procedures is the general operating instructions for operations.

B. CORRECT: OP-010 series procedures contains refuel tasks for operations. Most notably OP-010-006, Outage operations.

C. Incorrect. OP-100 contains operations programs such as EOS, watch station turnover sheets, and locked valve and breaker deviations.

D. Incorrect. The UNT are site program procedures.

Technical Reference(s): OP-010-006, Revision 320 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PPA00 obj. 1 (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 Comments:

Revision 0 Facility: Waterford 3 Page 148 of 202

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 3 Group # 4 K/A # G2.4.14 Importance Rating 3.8 K/A Statement Emergency Procedures / Plan: Knowledge of general guidelines for EOP usage.

Proposed Question: RO 75 Rev: 0 During EOP usage, Parameter values that are ____(1)____ should be used if a harsh environment in containment exists. A harsh environment is defined as Containment Temperature greater than or equal to ______(2)_____ .

(1) (2)

A. underlined 120°F B. underlined 200°F C. bracketed 120°F D. bracketed 200°F Revision 0 Facility: Waterford 3 Page 149 of 202

2012 NRC Exam RO Written Exam Worksheet Proposed Answer: D Explanation: (Optional)

A. Incorrect. Per OI-038-000, EOP Expectations and Guidelines, Parameter values that are bracketed should be used if a harsh environment in containment exists. Containment temperature of 120°F is the acceptable containment temperature found in the Containment Temperature and Pressure control safety function status checklist.

B. Incorrect. Per OI-038-000, EOP Expectations and Guidelines, Parameter values that are bracketed should be used if a harsh environment in containment exists.

C. Incorrect. Containment temperature of 120°F is the acceptable containment temperature found in the Containment Temperature and Pressure control safety function status checklist.

D. CORRECT: Per OI-038-000, EOP Expectations and Guidelines, Parameter values that are bracketed should be used if a harsh environment in containment exists. A harsh environment is defined as Containment Temperature greater than or equal to 200°F.

Technical Reference(s): OI-038-000, Revision 005 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PPE08 obj. 8 (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 Comments:

Revision 0 Facility: Waterford 3 Page 150 of 202

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 000008 AA2.23 Importance Rating 4.3 K/A Statement Ability to determine and interpret the following as they apply to the Pressurizer Vapor Space Accident: Criteria for throttling high-pressure injection after a small LOCA.

Proposed Question: SRO 1 Rev: 0 The following plant conditions exist:

A LOCA is progress due to a Pressurizer Safety Valve failing open

Pressurizer Level is 100%

Thot is 532°F

Representative Core Exit Thermocouple temperature is 538°F

Pressurizer Pressure is 1170 PSIA

QSPDS Reactor Vessel Levels 1-3 are voided, 4-8 are filled

High Pressure Safety Injection Pump (HPSI) B was secured and Train A flow was throttled 10 minutes ago

RCPs are secured Based on these conditions:

(1) What direction should the CRS provide?

(2) What is the basis for this direction?

(1) (2)

A. Restore Pressurizer heaters and RCS subcooling requirements are not raise RCS Pressure. met.

B. Raise HPSI flow and start HPSI RCS subcooling requirements are not Pump B and commence a cooldown. met.

C. Restore Pressurizer heaters and Reactor Vessel Level requirements raise RCS Pressure. are not met.

D. Raise HPSI flow and start HPSI Reactor Vessel Level requirements Pump B and commence a cooldown. are not met.

Revision 0 Facility: Waterford 3 Page 151 of 202

2012 NRC Exam RO Written Exam Worksheet Proposed Answer: B Explanation: (Optional)

E. Incorrect. Restoring Pressurizer heaters and raising RCS pressure is a valid method for restoring subcooling when the RCS is intact. However, for the conditions given it would only serve to increase RCS leakage from the safety valve. Correct parameter evaluation.

F. CORRECT. CET temperature should be used with natural circulation in progress.

Although Thot indicates greater than 28°F subcooling, CETs are < 28°F. This does not meet HPSI Throttle Criteria, requiring the operator to restore HPSI flow as needed to restore RCS subcooling. Commencing a cooldown will also assist in restoring subcooling margin G. Incorrect. Wrong action. Reactor Vessel Level throttle criteria is already met based on Level 5 NOT voided.

H. Incorrect. Correct action. Wrong parameter evaluation.

Technical Reference(s): OP-902-002, Rev. 14 (Attach if not previously provided) TG-OP-902-002, Rev. 013 (including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PPE02, Objective 8 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 5 Comments:

Revision 0 Facility: Waterford 3 Page 152 of 202

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 000029 G2.4.2 Importance Rating 4.6 K/A Statement Knowledge of system set points, interlocks and automatic actions associated with EOP entry conditions. (ATWS)

Proposed Question: SRO 2 Rev: 0 The following plant conditions exist:

The plant is performing a shutdown due to a 25 gpm primary to secondary leak on Steam Generator 1

The ATC operator reports the following indications:

Pressurizer Level is 45% and lowering

Pressurizer Pressure is 1750 PSIA and lowering

Containment Pressure is 16.7 PSIA and rising

RCS Temperature is lowering rapidly

Steam Generator 1 and 2 Pressure is 750 PSIA and lowering

All Charging Pumps are operating

Reactor Power is 78% and steady

Per CRS direction the ATC performs a manual reactor trip, SIAS and CIAS An Anticipated Transient Without Scram (ATWS) ______(1)______ occurred. After verification of Standard Post Trip Actions the CRS will implement

_________(2)____________.

(1) (2)

A. has OP-902-004, Excess Steam Demand Recovery B. has OP-902-008, Functional Recovery C. has NOT OP-902-004, Excess Steam Demand Recovery D. has NOT OP-902-008, Functional Recovery Revision 0 Facility: Waterford 3 Page 153 of 202

2012 NRC Exam RO Written Exam Worksheet Proposed Answer: B Explanation: (Optional)

A. Incorrect. Correct ATWS evaluation. Wrong procedure implementation.

B. CORRECT. Prior to the ATC tripping the reactor RCS pressure went below the Aux Trip setpoint in the CPCs for RCS Pressure, therefore an ATWS has occurred. Two events are in progress. The diagnostic flowchart would send the crew to OP-902-008. If for some reason the CRS misuses the Diagnostic Flowchart, OP-902-004 Safety Function Status Checklist will not support implementation of the procedure due to secondary activity, based on given conditions.

C. Incorrect. Wrong ATWS evaluation. Wrong procedure implementation.

D. Incorrect. Wrong ATWS evaluation. Correct procedure implementation.

Technical Reference(s): OP-902-009, Rev. 307 (Attach if not previously provided) TS Bases 2.2.1 Amendment 64 (including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PPE08 Obj. 7 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 5 Comments:

Revision 0 Facility: Waterford 3 Page 154 of 202

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 000055 EA2.06 Importance Rating 4.1 K/A Statement Ability to determine or interpret the following as they apply to a Station Blackout:

Faults and lockouts that must be cleared prior to re-energizing buses Proposed Question: SRO 3 Rev: 0 The following plant conditions exist:

Emergency Diesel Generator B is tagged out to repair the Speed Governor.

A grid fault resulting in a fault isolation and loss of offsite power.

A Sudden Pressure relay actuation occurs on Startup Transformer B.

Emergency Diesel Generator A started; then tripped on Generator Differential.

PME reports that Startup Transformer B Sudden Pressure actuation was a relay failure, and the relay has been replaced.

Gretna TOC reports that the fault has been removed and offsite power can be restored.

The SRO should implement ____(1)____ and the crew will need to reset Hand Reset (HR) lockout relays 86STB, ____(2)____ to restore offsite power to BOTH SAFETY busses.

(1) (2)

A. OP-902-003, Loss of Offsite 86A2 and 86B2 only Power/ Loss of Forced Circulation Recovery B. OP-902-003, Loss of Offsite 86A1, 86B1, 86A2, and 86B2 Power/ Loss of Forced Circulation Recovery C. OP-902-005, Station 86A2 and 86B2 only Blackout Recovery D. OP-902-005, Station 86A1, 86B1, 86A2, and 86B2 Blackout Recovery Revision 0 Facility: Waterford 3 Page 155 of 202

2012 NRC Exam RO Written Exam Worksheet Proposed Answer: C Explanation: (Optional)

A. Incorrect: A grid fault caused a loss of offsite power, but EDG B is out of service, and EDG A trips leaving the plant in a station blackout condition, therefore OP-902-005 is the correct EOP entry. 86STB/HR must be reset due to the relay tripping on the Sudden Pressure fault. 86A2/HR and 86B2/HR must be reset after power is restored to the 2 buses in order to close the 2-To-3 Tie Breakers, and restore offsite power to the 3 buses.

B. Incorrect: A grid fault caused a loss of offsite power, but EDG B is out of service, and EDG A trips leaving the plant in a station blackout condition, therefore OP-902-005 is the correct EOP entry. 86A1/HR and 86B1/HR are lockout resets for the A1 and B1 7KV buses and are not required for restoration of offsite power to the safety buses since they do not supply any other electrical buses.

C. CORRECT: A grid fault caused a loss of offsite power, EDG B is out of service, and EDG A trips leaving the plant in a station blackout condition, therefore OP-902-005 is the correct EOP entry. 86STB/HR must be reset due to it tripping on the Sudden Pressure fault. 86A2/HR and 86B2/HR must be reset after power is restored to the 2 buses in order to close the 2-To-3 Tie Breakers, and restore offsite power to the 3 buses.

D. Incorrect: A grid fault caused a loss of offsite power, EDG B is out of service, and EDG A trips leaving the plant in a station blackout condition, therefore OP-902-005 is the correct EOP entry. 86A1/HR and 86B1/HR are lockout resets for the A1 and B1 7KV buses and are not required for restoration of offsite power to the safety buses since they do not supply any other electrical buses.

Technical Reference(s): OP-902-009 Rev. 307 (Attach if not previously provided) SD-TSS, Rev. 9 (including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-TSS00, Obj.1,6 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 5 Comments:

Revision 0 Facility: Waterford 3 Page 156 of 202

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 000058 AA2.03 Importance Rating 3.9 K/A Statement AA2.03 Ability to determine and interpret the following as they apply to the Loss of DC Power: DC loads lost; impact on ability to operate and monitor plant systems Proposed Question: SRO 4 Rev: 0 The following plant conditions exist:

A loss of A-DC-S Bus has occurred

Reactor Trip has occurred

On the Generator Trip, 2A Bus failed to transfer to Startup Transformer A

The crew has entered OP-902-000, Standard Post Trip Actions and OP-901-313, Loss of 125V DC Bus concurrently Emergency Diesel Generator A is ___(1)___. The Control Room Supervisor should implement __(2)__.

(1) (2)

A. NOT running and its output OP-902-003, Loss of Offsite Power/ Loss breaker Open of Forced Circulation Recovery B. running with its output breaker OP-902-003, Loss of Offsite Power/ Loss Closed of Forced Circulation Recovery C. NOT running with its output OP-902-001, Reactor Trip Recovery and breaker Open OP-901-310, Loss of Train A Safety Bus D. running with its output breaker OP-902-001, Reactor Trip Recovery and Closed OP-901-310, Loss of Train A Safety Bus Revision 0 Facility: Waterford 3 Page 157 of 202

2012 NRC Exam RO Written Exam Worksheet Proposed Answer: C Explanation: (Optional)

A. Incorrect: EDG A did not start due to the loss of DC power to the starting air solenoids.

EDG A Output Breaker cannot auto close due to the loss of DC control power. Loads previously running have not stripped from 3A bus due to the loss of DC control power, therefore CCW Pump A breaker is closed, but has no power. The Sequencer cannot strip and sequence on loads due to the loss of DC.

B. Incorrect: EDG A did not start due to the loss of DC power to the starting air solenoids.

EDG A Output Breaker cannot auto close due to the loss of DC control power. Loads previously running have not stripped from 3A bus due to the loss of DC control power, therefore CCW Pump A breaker is closed, but has no power. For the conditions given only 2a, 3a and loads downstream have been lost. The loss of DC would cause a reactor trip. Diagnostics would result in implementing OP-902-001 Reactor Trip.

Power is lost to the 3A bus so OP-901-310 is also applicable.

C. CORRECT: EDG A cannot start or load due to the loss of DC power. The loads that were closed in prior to the loss of DC cannot strip since control power is required to automatically trip load breakers.

D. Incorrect: EDG A did not start due to the loss of DC power to the starting air solenoids.

EDG A Output Breaker cannot auto close due to the loss of DC control power. Loads previously running have not stripped from 3A bus due to the loss of DC control power.

The Sequencer cannot strip and sequence on loads due to the loss of DC.

Technical Reference(s): OP-901-310, Rev. 308 (Attach if not previously provided) OP-902-009 Rev. 307 (including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PPO30 Obj. 1 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 5 Comments:

Revision 0 Facility: Waterford 3 Page 158 of 202

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 000062 G2.1.32 Importance Rating 4.0 K/A Statement Ability to explain and apply system limits and precautions (Loss of Nuclear Service Water)

Proposed Question: SRO 5 Rev: 0 Given:

Reactor Power is 100%

Wet Cooling Tower (WCT) A and B Basin levels are verified to be 95%

The required action is to enter Technical Specification ___(1)___. The minimum water level ensures a ____(2)____ day cooling water supply to essential equipment.

(1) (2)

A. 3.0.3 7 B. 3.0.3 30 C. 3.7.4 7 D. 3.7.4 30 Revision 0 Facility: Waterford 3 Page 159 of 202

2012 NRC Exam RO Written Exam Worksheet Proposed Answer: D Explanation: (Optional)

A. Incorrect: Both trains of UHS are inoperable due to water levels below the minimum 97% required by TS LCO 3.7.4a. TS 3.0.3 is not entered since TS 3.7.4 provides an action for both UHS Trains inoperable. The limitation on minimum water level is based on providing a 30-day cooling water supply vice 7 days, which is a EDG fuel oil volume limitation.

B. Incorrect: TS 3.0.3 is not entered since TS 3.7.4 provides an action for both UHS Trains inoperable. The limitation on minimum water level is based on providing a 30-day cooling water supply.

C. Incorrect: Both trains of UHS are inoperable due to water levels below the minimum 97% required by TS LCO 3.7.4a. TS 3.7.4, Action b is entered for both UHS Trains inoperable. The limitation on minimum water level is based on providing a 30-day cooling water supply vice 7 days, which is a EDG fuel oil volume limitation.

D. CORRECT: Both trains of UHS are inoperable due to water levels below the minimum 97% required by TS LCO 3.7.4a. TS 3.7.4, Action b is entered for both UHS Trains inoperable. The limitation on minimum water level is based on providing a 30-day cooling water supply.

Technical Reference(s): TS 3.7.4 and Bases (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-CC00, Obj. 9 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 2 Comments:

Revision 0 Facility: Waterford 3 Page 160 of 202

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 000077 G2.2.36 Importance Rating 4.2 K/A Statement Ability to analyze the effect of maintenance activities, such as degraded power sources, on the status of limiting conditions for operations. (Generator Voltage and Electric Grid Disturbances)

Proposed Question: SRO 6 Rev: 0 The following plant conditions exist:

The plant is in MODE 1

The shift manager received a call from the grid operator informing him that maintenance activities are in progress that would result in Waterfords post-trip grid voltage to be 222 kV

Current grid and bus voltages meet the requirements of OP-903-066, Electrical Breaker Alignment Check The CRS should (1) and is required to (2)

(1) (2)

A. declare both AC off-site circuits initiate an operability evaluation of the Offsite inoperable and enter both Tech Circuits.

Spec 3.8.1.1 and TRM 3.8.1.1 B. declare both AC off-site circuits restore the predicted post trip Offsite AC inoperable and enter both Tech circuit voltage to > 223 KV within one hour Spec 3.8.1.1 and TRM 3.8.1.1 C. enter TRM 3.8.1.1 only restore the predicted post trip Offsite AC circuit voltage to > 223 KV within one hour D. enter TRM 3.8.1.1 only initiate an operability evaluation of the Offsite AC Circuits.

Revision 0 Facility: Waterford 3 Page 161 of 202

2012 NRC Exam RO Written Exam Worksheet Proposed Answer: D Explanation: (Optional)

A. Incorrect: Not required to enter TS unless AC circuits declared inoperable.

B. Incorrect: Not required to enter TS unless AC circuits declared inoperable.

C. Incorrect: Operations is required to initiate a CR and perform an operability assessment for off-site AC circuits. Sufficient margin is assumed available unless determined to be otherwise.

D. CORRECT: Operations is required to initiate a CR and perform an operability assessment for off-site AC circuits. Sufficient margin is assumed available unless determined to be otherwise.

Technical Reference(s): OP-901-314 E1 Step 7 TRM 3.8.1.1 & Bases (Attach if not previously provided) Tech Spec 3.8.1.1 & Bases (including version/revision number)

Proposed references to be provided TS 3.8.1.1, TRM 3.8.1.1, excerpt from TRM to applicants during examination: Bases Learning Objective: WLP-OPS-PPO30 Obj. 05 (As available)

Question Source: Bank # X WF3-OPS-08876 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam 2009 NRC Exam SRO #6 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 2 Comments:

Revision 0 Facility: Waterford 3 Page 162 of 202

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 2 K/A # 000036 G2.4.41 Importance Rating 4.6 K/A Statement Knowledge of the emergency action level thresholds and classifications. (Fuel Handling Accident)

Proposed Question: SRO 7 Rev: 0 The following plant conditions exist:

The plant is in MODE 6

A full core offload is in progress

Containment Purge is in operation with the Maintenance Hatch open

The Control Room received the following annunciators 20 minutes ago:

o CLASS 1E RAD MONITORING SYS ACTIVITY HI-HI (CP-18, Panel SA) o CLASS 1E RAD MONITORING SYS ACTIVITY HI-HI (CP-18, Panel SB) o RAD MONITORING SYS ACTIVITY HI-HI (CP-36, Panel L) o EFFLUENT RAD MONITORING SYS ACTIVITY HI-HI (CP-36, Panel L)

The Refueling SRO reported that a fuel assembly slipped off the grapple during withdrawal and fell back into the core.

The ATC operator reports the following radiation monitor indications:

o All four Containment Purge Isolation Radiation Monitors ARM-IRE-5024(5025)(5026)(5027) are in High Alarm o Plant Stack PIG Gas Channel (PRM-IRE-0100.1S(2S) are reading 1.00 E-1 Ci/cc o Plant Stack WRGM (PRM-IRE-0110, RE0110-4) indicates release rate 2.62E+08 Ci/sec and slowly rising

The BOP operator reports that Containment Purge will not secure automatically or manually.

Which of the following Emergency Plan classifications is correct for this event?

A. Unusual Event, AU2 B. Alert, AA2 C. Site Area Emergency, AS1 D. General Emergency, AG1 Revision 0 Facility: Waterford 3 Page 163 of 202

2012 NRC Exam RO Written Exam Worksheet Proposed Answer: C Explanation: (Optional)

A. Incorrect: Containment Purge Isolation Monitors are in High alarm, which meets part of AA2 criteria, but the Plant Stack WRGM reading supersedes the Containment Purge Monitor data, requiring entry into a higher classification.

B. Incorrect: Containment Purge Isolation Monitors are in High alarm, which exceeds part of AU2 criteria, but the Plant Stack PIGs and Plant Stack WRGM readings supersede the Containment Purge Monitor data, requiring entry into a higher classification.

C. CORRECT: The Plant Stack WRGM release rate reading is >2.55E+08 Ci/sec, requiring the candidate to classify the event as a Site Area Emergency, AS1 based on EAL 1.

D. Incorrect: No information provided reaches the General Emergency threshold. The Plant Stack WRGM release rate would have to be >2.55E+09 Ci/sec to classify a General Emergency.

EP-001-001, Recognition and Classification of Technical Reference(s): Emergency Conditions, Rev 030 - TAB A (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided EP-001-001, Recognition and Classification of to applicants during examination: Emergency Conditions Learning Objective: WLP-OPS-EP02 Obj. 17 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 4 Comments:

Revision 0 Facility: Waterford 3 Page 164 of 202

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 2 K/A # 000037 G2.2.22 Importance Rating 4.7 K/A Statement Knowledge of limiting conditions for operations and Safety Limits. (Steam Generator Tube Leak)

Proposed Question: SRO 8 Rev: 0 Given:

The Plant is at 100% power and steady state

The ATC operator reports that the N-16 Radiation Monitor for Steam Generator 2 reads 65 GPD.

PSLR reads 85 GPD

Chemistry reports that the primary to secondary leakage in Steam Generator 2 is consistent with PSLR.

The CRS should enter Technical Specification _____(1)______.

The Nuclear Energy Institute recommended a limit of 150 GPD.

The limit in Technical Specification 3.4.5.2 is more restrictive ____(2)_____.

(1) (2)

A. 3.4.5.2, action a to ensure the magnitude of leakage does not interfere with the detection of UNIDENTIFIED LEAKAGE by the leakage detection systems B. 3.4.5.2, action b to ensure the magnitude of leakage does not interfere with the detection of UNIDENTIFIED LEAKAGE by the leakage detection systems C. 3.4.5.2, action a because the proximity of the east ADV to the east control room air intake could result in unacceptable radiological consequences D. 3.4.5.2, action b because the proximity of the east ADV to the east control room air intake could result in unacceptable radiological consequences Revision 0 Facility: Waterford 3 Page 165 of 202

2012 NRC Exam RO Written Exam Worksheet Proposed Answer: C Explanation: (Optional)

A. Incorrect: Part 1 is correct - T.S. 3.4.5.2 Action a. With any PRESSURE BOUNDARY LEAKAGE, or primary to secondary leakage not within limit, be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . T.S. LCO 3.4.5.2 LCO c requires entry for 75 gallons per day primary to secondary leakage through any one steam generator (SG), which 85 GPD exceeds.

The description in Part 2 is the bases for IDENTIFIED LEAKAGE limits, not primary to secondary leakage.

B. Incorrect: Primary to secondary leakage is specifically excluded from TS 3.4.5.2, Action b. The description in Part 2 is the bases for IDENTIFIED LEAKAGE limits, not primary to secondary leakage.

C. CORRECT: T.S. 3.4.5.2 Action a. is correct based on given leakage greater than 75 gallons per day primary to secondary leakage through SG #2. The bases is correct based on not exceeding 5 REM TEDE per person for the duration of the accident.

D. Incorrect: Primary to secondary leakage is specifically excluded from TS 3.4.5.2, Action b. The bases is correct based on not exceeding 5 REM TEDE per person for the duration of the accident.

Technical Reference(s): Technical Specification 3.4.5.2, Amendment 204 (Attach if not previously provided) and Bases (including version/revision number)

Proposed references to be provided to applicants during examination: Technical Specification 3.4.5.2 Learning Objective: WLP-OPS-RCS00 Obj. 9 (As available)

WLP-OPS-PPO20 Obj. 5 Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 2 Comments:

Revision 0 Facility: Waterford 3 Page 166 of 202

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 2 K/A # 000061 AA2.06 Importance Rating 4.1 K/A Statement Ability to determine or interpret the following as they apply to the Area Radiation Monitoring (ARM) System Alarms: Required actions if alarm channel is out of service Proposed Question: SRO 9 Rev: 0 The following plant conditions exist:

The plant is in MODE 3

Fuel movement is occurring in the Spent Fuel Pool Containment Purge Monitor Reading Monitor Alarm/Trip Setpoint ARM-IRE-5024S reads 150 mR/hr 320 mR/hr ARM-IRE-5025S reads 67 mR/hr 120 mR/hr ARM-IRE-5026S reads 15 mR/hr 50 mR/hr ARM-IRE-5027S reads 19 mR/hr 40 mR/hr ARM-IRE-____(1)___ are inoperable. Enter Technical Specification ___(2)____.

(1) (2)

A. 5025S and 5027S 3.3.3.1 & 3.9.4 B. 5024S and 5026S 3.3.3.1 & 3.9.4 C. 5025S and 5027S 3.3.3.1 only D. 5024S and 5026S 3.3.3.1 only Revision 0 Facility: Waterford 3 Page 167 of 202

2012 NRC Exam RO Written Exam Worksheet Proposed Answer: D Explanation: (Optional)

A. Incorrect: ARM-IRE-5025S and ARM-IRE-5027S are operable. 5025S based on setpoint less than 2 times background and 5027S based on setpoint appropriately at 40 mR/hr. No Tech Spec entry is required for these monitors.

B. Incorrect: ARM-IRE-5024S is inoperable because its setpoint is greater than 2 times background. ARM-IRE-5026S is inoperable because its setpoint should be the higher of 2 times background (30 mR/hr) or 40 mR/hr. Based on both monitors inoperable on the same train, Tech Spec 3.3.3.1 entry is required. T.S. 3.9.4 entry not required since is only applicable during CORE ALTERATIONS or load movements with or over irradiated fuel within the containment, which cannot occur in MODE 3.

C. Incorrect: ARM-IRE-5025S and ARM-IRE-5027S are operable. 5025S based on setpoint less than 2 times background and 5027S based on setpoint appropriately at 40 mR/hr. No Tech Spec entry is required for these monitors.

D. CORRECT: ARM-IRE-5024S is inoperable because its setpoint is greater than 2 times background. ARM-IRE-5026S is inoperable because its setpoint should be the higher of 2 times background (30 mR/hr) or 40 mR/hr. Based on both monitors inoperable on the same train, Tech Spec 3.3.3.1 entry is required. T.S. 3.9.4 is only applicable during CORE ALTERATIONS or load movements with or over irradiated fuel within the containment, which cannot occur in MODE 3.

Technical Reference(s): Tech Specs 3.3.3.1 & 3.9.4 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: Tech Specs 3.3.3.1 & 3.9.4 Learning Objective: WLP-OPS-RMS00 Obj. 2,7 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 2 Comments:

Revision 0 Facility: Waterford 3 Page 168 of 202

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 2 K/A # CE/E09 EA2.1 Importance Rating 4.4 K/A Statement Ability to determine and interpret the following as they apply to the (Functional Recovery): Facility conditions and selection of appropriate procedures during abnormal and emergency operations.

Proposed Question: SRO 10 Rev: 0 The following plant conditions exist:

The plant has tripped

The crew has entered OP-902-008, Functional Recovery procedure

A Loss of Offsite Power has occurred

Emergency Diesel Generator A failed to start

RCS pressure is 1150 PSIA

Containment Pressure is 17.8 PSIA

HPSI flow is 100 GPM per leg

CET temperature is 451 degrees F

Activity is indicated on Main Steam Line Rad Monitor 1 (ARM-IRE-5500A)

Steam Generator 1 pressure is 300 PSIA and dropping

Subcooled margin is 111 degrees F

Containment Temperature is 216 degrees F

RVLMS level indicates 100 percent for both the head and the plenum

Steam Generator 2 level is 78 percent WR and lowering slowly Assuming the following:

All actuations and other equipment functioned as required

Operators have only performed actions required by OP-902-000, Standard Post Trip Actions Which section of OP-902-008, Functional Recovery procedure should be implemented first?

A. MVA-AC-2 B. IC-2 C. HR-2 D. CI-1 Revision 0 Facility: Waterford 3 Page 169 of 202

2012 NRC Exam RO Written Exam Worksheet Proposed Answer: D Explanation: (Optional)

A. Incorrect: AC-2 is the highest priority Safety Function that does not meet Success Path one criterion based on only one EDG energizing a safety bus; and is therefore second in prioritization.

B. Incorrect: IC-2 is the second highest priority Safety Function that does not meet Success Path one criteria due to the SIAS actuation based on Containment Pressure

>17.1 PSIA and/or Pressurizer Pressure <1684 PSIA; and is therefore third in prioritization.

C. Incorrect: HR-2 is Safety Function is met, but is a lower priority Safety Function than IC-2 and therefore is the fourth in priority.

D. CORRECT: Candidates should prioritize Safety Function Status Checklist from top to bottom as RC-1=6, MVA-DC-1=7, MVA-AC-2=2, IC-2=3, PC-1=8, HR-2=4, CI-1=1, CTPC-2=5. CI-1 is first priority based upon based upon being the only Safety Function that is not met. This is due to the fact that all release paths from the most affected SG to the environment are NOT closed.

OP-902-008, Functional Recovery Procedure, Technical Reference(s): Revision 018 (Attach if not previously provided) Pages 14-23 (including version/revision number)

OP-902-008, Functional Recovery Procedure, Proposed references to be provided Pages 1-23 to applicants during examination: OP-902-009, Standard Appendices, Appendix 2 Learning Objective: WLP-OPS-PPE08, Objective 6 (As available)

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

New Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 5 Comments:

Revision 0 Facility: Waterford 3 Page 170 of 202

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 006 A2.02 Importance Rating 4.3 K/A Statement A2.02 Ability to (a) predict the impacts of the following malfunctions or operations on the ECCS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Loss of flow path Proposed Question: SRO 11 Rev: 0 The following plant conditions exist:

A Loss of Coolant Accident occurred 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months ago

The crew diagnosed into OP-902-002, Loss of Coolant Accident Recovery procedure

Recirculation Actuation Signal has occurred, and the actions have been addressed

The RCS is 15°F subcooled, but CET temperatures are starting to rise very slowly

RVLMS shows level 1-6 voided

The BOP notes that High Pressure Safety Injection Pump (HPSI) A and B cold leg flow has dropped off slowly with discharge pressure rising and no oscillations apparent

HPSI cold leg flow criteria is met

Containment Spray Pump A and B are running with flow ~ 2000 GPM and stable The CRS should ________(1)________. The conditions given indicate

_______(2)______.

(1) (2)

A. remain in OP-902-002 and align core flow blockage is starting to occur due hot and cold leg injection to boron precipitation B. secure one HPSI pump and go core flow blockage is starting to occur due to OP-902-008, Functional to boron precipitation Recovery C. remain in OP-902-002 and align the SI Sump Strainers are starting to clog hot and cold leg injection from debris collection D. secure one HPSI pump and go the SI Sump Strainers are starting to clog to OP-902-008, Functional from debris collection Recovery Revision 0 Facility: Waterford 3 Page 171 of 202

2012 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 172 of 202

2012 NRC Exam RO Written Exam Worksheet Proposed Answer: A Explanation: (Optional)

A. CORRECT. The indications given indicate a problem developing on the discharge side of the High Pressure Safety Injection Pumps. At 2-3 hours post LOCA the crew should align hot and cold leg injection to avoid concentrating boric acid in the reactor vessel and possible boron precipitation in the core which could restrict coolant flow through the core.

B. Incorrect. Core flow blockage indications are given making Part 2 correct. Either condition listed in Part 2 is addressed in OP-902-002, and therefore, OP-902-008 entry is not required.

C. Incorrect. The indications given indicate a problem developing on the discharge side of the High Pressure Safety Injection Pumps, not the suction side (SI Sump Strainers). Step 47 of OP-902-002 directs performance of Hot and Cold Leg Injection 2 to3 hours post-LOCA. Either condition listed in Part 2 is addressed in OP-902-002, and therefore, OP-902-008 entry is not required.

D. Incorrect. Both parts are incorrect. The indications given indicate a problem developing on the discharge side of the High Pressure Safety Injection Pumps, not the suction side (SI Sump Strainers). Step 45 of OP-902-002 directs going down to one HPSI pump operating to prevent heat buildup and subsequent pump seal damage in low flow conditions. This condition is addressed in OP-902-002, and therefore, OP-902-008 entry is not required.

Technical Reference(s): OP-902-002 Rev. 14 (Attach if not previously provided) TGOP-902-002 Rev. 013 (including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PPE02, Objective 19 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 5 Comments:

Revision 0 Facility: Waterford 3 Page 173 of 202

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 013 G2.2.22 Importance Rating 4.7 K/A Statement Knowledge of limiting conditions for operations and safety limits. (Engineering Safeguards Features Actuation)

Proposed Question: SRO 12 Rev: 0 Given:

The plant is in MODE 3 preparing for startup.

While performing OP-903-107, Plant Protection System Function Test, the operator notes that the Low Steam Generator Water Level No. 1 trip bistable is tripping when the test voltage is lowered to 2.03 VDC.

The following Technical Specifications have been reviewed:

o 2.2.1, Reactor Trip Setpoints o 3.3.1, Reactor Protective System Instrumentation o 3.3.2, Engineering Safety Features Actuation System Instrumentation This requires that Technical Specifications _______(1)______ be entered. The action is to bypass _________(2)__________.

(1) (2)

A. 2.2.1, 3.3.1, and Steam Generator 1 Level Low, Steam Generator P 1 and 2 3.3.2 B. 3.3.2 only Steam Generator 1 Level Low, Steam Generator P 1 and 2 C. 2.2.1, 3.3.1, and Steam Generator 1 Level Low, Steam Generator P 1 only 3.3.2 D. 3.3.2 only Steam Generator 1 Level Low, Steam Generator P 1 only Revision 0 Facility: Waterford 3 Page 174 of 202

2012 NRC Exam RO Written Exam Worksheet Proposed Answer: D Explanation: (Optional)

A. Incorrect: 2.2.1 is only applicable for the instrumentation that is inoperable for the modes that the associated trip function instrumentation is applicable for. Steam Generator Level Low is only applicable in MODE 1 and 2 for TS 3.3.1. Steam Generator P 2 would only have to be bypassed if a Steam Generator Pressure channel were out of service.

B. Incorrect. The listed action is correct for the current plant mode. Steam Generator P 2 would only have to be bypassed if a Steam Generator Pressure channel were out of service.

C. Incorrect: 2.2.1 is only applicable for the instrumentation that is inoperable for the modes that the associated trip function instrumentation is applicable for. Steam Generator Level Low is only applicable in MODE 1 and 2 for TS 3.3.1.

D. CORRECT: Steam Generator Level Low is applicable in Modes 1, 2, and 3, for TS 3.3.1. The action is to bypass Steam Generator 1 Level Low, Steam Generator P 1 only. Steam Generator P 2 would only have to be bypassed if a Steam Generator Pressure channel were out of service. TS 3.3.1 and 2.2.1 are not applicable for Steam Generator Level low in MODE 3.

Technical Reference(s): TS 2.2.1 (Attach if not previously provided) TS 3.3.1 (including version/revision number) TS 3.3.2 OP-903-107, Plant Protection System, Rev. 304 Proposed references to be provided to applicants during examination: TS 2.2.1, 3.3.1, 3.3.2 Learning Objective: WLP-OPS-PPS00 Obj. 8 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 2 Comments:

Revision 0 Facility: Waterford 3 Page 175 of 202

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 026 G2.1.20 Importance Rating 4.6 K/A Statement Ability to interpret and execute procedure steps.

Proposed Question: SRO 13 Rev: 0 The following plant conditions exist:

The plant has experienced a Loss of Coolant Accident concurrently with a Loss of Offsite Power

EDG A started and loaded, but EDG B tripped on overspeed

The crew has entered OP-902-002, Loss of Coolant Accident Recovery

Containment Pressure is 17.8 PSIA and rising

RCS Pressure is 850 PSIA and steady

Containment Spray Pump A tripped on overcurrent In response to this event, the SRO should ____(1)___ and ______(2)______.

(1) (2)

A. continue in OP-902-002, Loss of realign LPSI Pump A to replace Coolant Accident Recovery Containment Spray Pump A B. continue in OP-902-002, Loss of override and close CS-125A Coolant Accident Recovery C. go to OP-902-008, Functional realign LPSI Pump A to replace Recovery Containment Spray Pump A D. go to OP-902-008, Functional override and close CS-125A Recovery Revision 0 Facility: Waterford 3 Page 176 of 202

2012 NRC Exam RO Written Exam Worksheet Proposed Answer: C Explanation: (Optional)

A. Incorrect: The Safety Function Status Checklist for OP-902-002 requires at least one Containment Spray (CS) pump running with greater than or equal to 1750 GPM flow.

Therefore, the crew would either enter directly into, or rediagnose into OP-902-008.

OP-902-008 Continuing Actions will lead the crew to Appendix 28 to align LPSI Pump A to replace CS Pump A.

B. Incorrect: The Safety Function Status Checklist for OP-902-002 requires at least one Containment Spray (CS) pump running with greater than or equal to 1750 GPM flow.

Therefore, the crew would either enter directly into, or rediagnose into OP-902-008.

Although the crew could override closed CS-125A to meet the CI-1 safety function, they would have to re-open CS-125A to meet the CTPC-2 safety function while aligning LPSI Pump A to replace CS Pump A.

C. CORRECT: Safety Function Status Checklist for OP-902-002 requires at least one Containment Spray (CS) pump running with greater than or equal to 1750 GPM flow. The crew would either enter directly into, or rediagnose into OP-902-008.

OP-902-008 Continuing Actions will lead the crew to Appendix 28 to align LPSI Pump A to replace CS Pump A with TSC concurrence.

D. Incorrect: Entry into OP-902-008 is correct. The crew could override closed CS-125A to meet the CI-1 safety function, but they would have to re-open CS-125A to meet the CTPC-2 safety function while aligning LPSI Pump A to replace CS Pump A.

OP-902-002, Loss of Coolant Accident Recovery, Technical Reference(s): Revision 014 (Attach if not previously provided) OP-902-008, Functional Recovery, Revision 018 OP-902-009, Standard Appendices, Revision 307, (including version/revision number) Standard Appendix 28, Attachment 28-A TG-OP-902-009, Revision 306 Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PPE02, Objective 19 (As available)

Question Source: Bank # X WF3-OPS-07137 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 5 Comments:

Revision 0 Facility: Waterford 3 Page 177 of 202

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Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 064 A2.06 Importance Rating 3.3 K/A Statement A2.06 Ability to (a) predict the impacts of the following malfunctions or operations on the ED/G system; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Operating unloaded, lightly loaded, and highly loaded time limit Proposed Question: SRO 14 Rev: 0 The following plant conditions exist:

The crew is performing a post-maintenance 24-hour EDG B run in accordance with OP-903-116, Train B Integrated Emergency Diesel Generator/Engineering Safety Features Test.

15 minutes ago, the crew loaded EDG B to 4.84 MW.

EDG B operation at this load is allowed for an additional ____(1)____ hours. After the time limit is reached then reduce load to less than or equal ____(2)_____ KW.

(1) (2)

A. 1.75 2200 B. 1.75 4400 C. 5.75 2200 D. 5.75 4400 Revision 0 Facility: Waterford 3 Page 178 of 202

2012 NRC Exam RO Written Exam Worksheet Proposed Answer: B Explanation: (Optional)

OP-009-002, Emergency Diesel Generator, Limitation 3.2.5 states, During normal operation, Emergency Diesel Generator ratings of 4.4 MW for continuous loading and 4.84 MW for 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months out of any 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months should not be exceeded.

A. Incorrect: Continuous rating is 4400 KW. Operation past two hours at this rate will cause generator winding damage.

B. CORRECT: 4840 KW is the EDG 2-hour maximum rating.

C. Incorrect: 2-hour rating is 4840 KW. Plausible if applicant confuses the ratings and time allowed. Both rates listed are actual operating rates.

D. Incorrect: 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months rating is 4840 KW. EDG B is above its continuous operating rate of 4400 KW.

OP-009-002, Emergency Diesel Generator, Technical Reference(s): Revision 318 SD-EDG, Emergency Diesel Generator System (Attach if not previously provided) Description (including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-EDG00, Objective 2 (As available)

Question Source: Bank # X WLP-OPS-08208a 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 55.43 5 Comments:

Revision 0 Facility: Waterford 3 Page 179 of 202

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 103 A2.04 Importance Rating 3.6*

K/A Statement A2.04 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: Containment evacuation (including recognition of the alarm)

Proposed Question: SRO 15 Rev: 0 The following plant conditions exist:

The plant is in MODE 6

A full core offload is in progress in preparation to replace Steam Generators

The following alarms and indications are received:

o CLASS 1E RAD MONITORING SYSTEM ACTIVITY HI-HI (Cabinet SA & SB) o RAD MONITORING SYSTEM ACTIVITY HI-HI (Cabinet L) o Plant Stack PIG Iodine Channels read 3.54E-02 ci/cc o Plant Stack WRGM Effluent Rate reads 1.62E+06 o Containment Purge Monitors ARM-IRE-5024(5026)(5027) are in Alert o Containment Purge Monitor ARM-IRE-5025 is in High Alarm o The Refueling SRO reports that a fuel assembly has dropped from the grapple and fallen back into the core o Bubbles are rising out of the water in the area of the dropped assembly The crew will enter ___(1)___, and action should be directed to evacuate all ___(2)___?

(1) (2)

A. OP-901-403, High Airborne non-essential personnel from the site Activity in Containment only B. OP-901-403, High Airborne non-essential personnel from the site Activity in Containment and OP-901-405, Fuel Handling Incident concurrently C. OP-901-405, Fuel Handling personnel from Containment Incident only D. OP-901-403, High Airborne personnel from Containment Activity in Containment and OP-901-405, Fuel Handling Incident concurrently Revision 0 Facility: Waterford 3 Page 180 of 202

2012 NRC Exam RO Written Exam Worksheet Proposed Answer: D Explanation: (Optional)

A. Incorrect: Step 12 directs concurrent entry into OP-901-405 if the cause of high activity is a leaking Fuel bundle. The Emergency Director determines the need for site evacuation based on their E-Plan classification, not the procedures used here.

The readings given would only result in an Alert classification, which would not require a site evacuation.

B. Incorrect: Alarms and indications match those listed in the Symptoms section of OP-901-403. Step 4 of OP-901-403 directs evacuation of Containment. Step 12 directs concurrent entry into OP-901-405 if the cause of high activity is a leaking Fuel bundle. The Emergency Director determines the need for site evacuation based on their E-Plan classification, not the procedures used here. The readings given would only result in an Alert classification, which would not require a site evacuation.

C. Incorrect: Alarms and indications match those listed in the Symptoms section of OP-901-403. Step 4 of OP-901-403 directs evacuation of Containment. Step 12 directs concurrent entry into OP-901-405 if the cause of high activity is a leaking Fuel bundle.

D. CORRECT: Alarms and indications match those listed in the Symptoms section of OP-901-403. Step 4 of OP-901-403 directs evacuation of Containment. Step 12 directs concurrent entry into OP-901-405 if the cause of high activity is a leaking Fuel bundle. OP-901-405 indications include Bubbles emerging from submerged, dropped, OR damaged Fuel Assembly.

Technical Reference(s): OP-901-403, High Airborne Activity in (Attach if not previously provided) Containment, Revision 003 (including version/revision number) OP-901-405, Fuel Handling Incident, Revision 005 Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PPO40, Obj. 1 & 4 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 5 Comments:

Revision 0 Facility: Waterford 3 Page 181 of 202

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 2 K/A # 015 G2.1.7 Importance Rating 4.7 K/A Statement Conduct of Operations: 2.1.7 Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation.

Proposed Question: SRO 16 Rev: 0 The following plant conditions exist:

The plant was operating at 100% when CEA 32 dropped into the core

The crew has entered OP-901-102, CEA or CEDMCS Malfunction

The crew is commencing a Rapid Plant Power Reduction to comply with TS 3.1.3.1 The CRS must commence a power reduction within ___(1)___ minutes. The appropriate power indication to use for this power reduction is ______(2)______.

(1) (2)

A. 15 COLSS Descending Power Track B. 15 Calibrated Excore Power C. 60 COLSS Descending Power Track D. 60 Calibrated Excore Power Revision 0 Facility: Waterford 3 Page 182 of 202

2012 NRC Exam RO Written Exam Worksheet Proposed Answer: A Explanation: (Optional)

A. CORRECT: COLR Figure 3 provides a minimum required power reduction increment with the power reduction commencing within 15 minutes of a deviated CEA. OP-901-212 directs the use of COLSS Descending Power Track during power reduction because it automatically selects and displays the correct power indication.

B. Incorrect. COLR Figure 3 indicates that the power reduction must be completed within 60 minutes of a deviated CEA, but started within 15 minutes. OP-901-212 directs the use of COLSS powers due to the CEA shadowing and power peaking effects on Excore Nuclear Instruments.

C. Incorrect. COLR Figure 3 indicates that the power reduction must be completed within 60 minutes of a deviated CEA, but started within 15 minutes. OP-901-212 directs the use of COLSS Descending Power Track during power reduction because it automatically selects and displays the correct power indication.

D. Incorrect. COLR Figure 3 indicates that the power reduction must be completed within 60 minutes of a deviated CEA, but started within 15 minutes. OP-901-212 directs the use of COLSS powers due to the CEA shadowing and power peaking effects on Excore Nuclear Instruments.

Technical Reference(s): TS 3.1.3.1 and COLR Figure 3 (Attach if not previously provided) OP-901-212, Rapid Plant Power Reduction, (including version/revision number) Rev 004 OP-901-102, CEA or CEDMCS Malfunction, Rev. 301 Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PPO10, Objective 3 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 5 Comments:

Revision 0 Facility: Waterford 3 Page 183 of 202

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 2 K/A # 033 A2.03 Importance Rating 3.5 K/A Statement A2.03 Ability to (a) predict the impacts of the following malfunctions or operations on the Spent Fuel Pool Cooling System ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Abnormal spent fuel pool water level or loss of water level Proposed Question: SRO 17 Rev: 0 The following plant conditions exist:

The plant is in MODE 6

A full core offload to the Spent Fuel Pool is in progress

The following alarms have been received at CP-2:

o Fuel Pool Temperature Hi o Fuel Pool Pumps Discharge Press Lo o Fuel Pool Level Lo Based on these indications, the CRS should enter ___(1)___ and suspend ___(2)___.

(1) (2)

A. OP-901-510, Component Cooling Water spent fuel movement and fill the System Malfunction Spent Fuel Pool B. OP-901-510, Component Cooling Water core alterations and open the System Malfunction CCW isolations to the AB Header C. OP-901-513, Spent Fuel Pool Cooling spent fuel movement and fill the Malfunction Spent Fuel Pool D. OP-901-513, Spent Fuel Pool Cooling core alterations and open the Malfunction CCW isolations to the AB Header Revision 0 Facility: Waterford 3 Page 184 of 202

2012 NRC Exam RO Written Exam Worksheet Proposed Answer: C Explanation: (Optional)

A. Incorrect: The combination of the Fuel Pool Level Lo and Fuel Pool Pumps Discharge Press Lo indicates that Fuel Pool level has dropped to the Fuel Pool Cooling Pumps lo level trip setpoint. Makeup is required to restore level and restart Fuel Pool Cooling. Fuel Pool Hi Temperature is due to loss of the cooling mechanism. Part 1, OP-901-510 is incorrect because it only addresses a loss of CCW Pump, not level restoration. Annunciator response procedure also directs the operator to OP-901-513. Part 2 is correct based on OP-901-513, steps 2 (for level restoration) and 4 (requiring suspension of fuel movement in the pool.

B. Incorrect: Part 1, OP-901-510 is incorrect because it only addresses a loss of CCW Pump. Annunciator response procedure directs the operator to OP-901-513. Part 2 is incorrect because there are no corresponding indications of refueling cavity level below requirements or loss of CCW cooling to the Fuel Pool.

C. CORRECT: Annunciator response procedure directs the operator to OP-901-513.

Part 2 is correct based on OP-901-513, steps 2 (for level restoration) and 4 (requiring suspension of fuel movement in the pool.

D. Incorrect: Annunciator response procedure directs the operator to OP-901-513. Part 2 is incorrect because there are no corresponding indications of refueling cavity level below requirements or loss of CCW cooling to the Fuel Pool.

Technical Reference(s): OP-901-513, Spent Fuel Pool Cooling Malfunction (Attach if not previously provided) Rev. 006 (including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PPO50, Objective 3 (As available)

Question Source: Bank # X WF3-OPS-09491 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam 2010 NRC Exam SRO #7 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 5 Comments:

Revision 0 Facility: Waterford 3 Page 185 of 202

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 2 K/A # 034 K4.02 Importance Rating 3.3 K/A Statement K4.02 Knowledge of design feature(s) and/or interlock(s) which provide for the following: Fuel movement Proposed Question: SRO 18 Rev: 0 Which of the following describes a decision making responsibility of the Refueling SRO?

A. Directing use of the key override to move the Refueling Machine hoist in the upward direction with a fuel assembly in the core region.

B. Directing use of the bypass position key-operated switch on the New Fuel Elevator panel to lower the elevator with a fuel assembly in the elevator.

C. Directing override of Spent Fuel Handling Machine interlocks to place a spent fuel assembly in the New Fuel Elevator.

D. Directing override of Spent Fuel Handling Machine interlocks to lift the CEA tool with a weight in excess of 200 lbs above the weight of the tool.

Revision 0 Facility: Waterford 3 Page 186 of 202

2012 NRC Exam RO Written Exam Worksheet Proposed Answer: A Explanation: (Optional)

Distractors derived from RF-005-002, Refueling Equipment Operation, Section 3.2 Limitations A. CORRECT: Limitation 3.35 - The manual handwheel or key override shall not be used to move RFM hoist in upward direction with a fuel assembly in the core region without Refueling SRO permission. [T.S. 3.9.6]

B. Incorrect. Limitation 3.32 requires Refueling SRO permission to place the key-operated switch in bypass position to RAISE the elevator with a fuel assembly in the elevator.

C. Incorrect. Limitation 3.23 - Spent Fuel shall not be placed in the New Fuel Elevator (NFE).

D. Incorrect. Limitation 3.29 - During lifting of the CEA handling tool, if a weight in excess of 200 lbs above the weight of the tool is observed, then stop and contact the Refueling Supervisor and Refueling SRO. No allowance is made to continue lifting the CEA handling tool with excess weight since an unplanned lift of a fuel assembly may occur.

RF-005-002, Refueling Equipment Operation, Technical Reference(s): Revision 320 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None WLP-OPS-FHS00, Fuel Learning Objective: Handling Systems (As available)

Objectives 1, 2, & 4 Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 7 Comments:

Revision 0 Facility: Waterford 3 Page 187 of 202

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 3 Group # 1 K/A # G2.1.5 Importance Rating 3.9 K/A Statement Conduct of Operations: Ability to use procedures related to shift staffing, such as minimum crew complement, overtime limitations, etc.

Proposed Question: SRO 19 Rev: 0 The following plant conditions exist:

Reactor is shutdown with all CEAs fully inserted

RCS Boron Concentration is 2100 ppm

Keff is 0.945

RCS Temperature is 195°F Which ONE of the following correctly describes the minimum required shift staffing in accordance with Tech Spec 6.2.2?

A. 1 Shift Manager, 1 SRO, 1 RO, 1 STA and 2 NAOs B. 1 Shift Manager, 1 RO, and 1 NAO C. 1 Shift Manager, 1 RO, 1 STA and 1 NAO D. 1 Shift Manager, 1 SRO, 1 RO and 2 NAOs Revision 0 Facility: Waterford 3 Page 188 of 202

2012 NRC Exam RO Written Exam Worksheet Proposed Answer: B Explanation: (Optional)

A. Incorrect: This crew composition exceeds the requirement of T.S. 6.2.2, Table 6.2-1 for MODE 5 by 1 SRO, 1 RO and 1 STA.

B. CORRECT: This crew composition meets the requirement of T.S. 6.2.2, Table 6.2-1 for MODE 5.

C. Incorrect: This crew composition exceeds the requirement of T.S. 6.2.2, Table 6.2-1 for MODE 5 by 1 STA.

D. Incorrect: This crew composition exceeds the requirement of T.S. 6.2.2, Table 6.2-1 for MODE 5 by 1 SRO and 1 NAO.

Technical Reference(s): Technical Specification 6.2.2 and associated (Attach if not previously provided) Table 6.2-1 Minimum Shift Crew Composition (including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-TS03, Objective 6 (As available)

Question Source: Bank # X WF3-OPS-7256-A Modified Bank # See Comments below.

New Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 5 Comments:

Revision 0 Facility: Waterford 3 Page 189 of 202

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 3 Group # 1 K/A # G2.1.37 Importance Rating 4.6 K/A Statement Conduct of Operations: Knowledge of procedures, guidelines, or limitations associated with reactivity management.

Proposed Question: SRO 20 Rev: 0 The following plant conditions exist:

The plant is in a startup Chemistry hold at 3% power

The ATC operator reports the following indications:

o Reactor power is now 5.4% and rising o Tcold is 539°F and lowering rapidly o No control rod movement is in progress o No boration or dilution evolutions are in progress Which of the following orders should the CRS give to the crew for this event?

A. Insert CEAs to restore Reactor Power to less than 5%.

B. Withdraw CEAs to restore RCS Tcold to 541°F to 543°F.

C. Insert all Reg Group and Group P CEAs fully into the core.

D. Trip the reactor and commence Emergency Boration.

Revision 0 Facility: Waterford 3 Page 190 of 202

2012 NRC Exam RO Written Exam Worksheet Proposed Answer: D Explanation: (Optional)

A. Incorrect: Indications given are for an uncontrolled cooldown of the RCS, which requires Emergency Boration. The reactor should be tripped as a part of establishing reactivity control.

B. Incorrect: Withdrawing CEAs to restore primary system temperature caused by a plant transient shall not be attempted.

C. Incorrect: This is the required action if the reactor goes critical outside of the plus or minus 0.5%k/k calculated on the ECC. The given indications are that the reactor is at the Point-of-Adding-Heat.

D. CORRECT: Emergency Boration is required due to the uncontrolled cooldown.

Tripping of the reactor is a conservative measure to ensure reactivity control safety function is met promptly.

Technical Reference(s): OP-901-103, Emergency Boration, Revision 002 (Attach if not previously provided) OP-010-003, Plant Startup, Revision 324 (including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PPO10, Objective 1 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 6 Comments:

Revision 0 Facility: Waterford 3 Page 191 of 202

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 3 Group # 2 K/A # G2.2.7 Importance Rating 3.6 K/A Statement Equipment Control: Knowledge of the process for conducting special or infrequent tests.

Proposed Question: SRO 21 Rev: 0 An Infrequently Performed Test or Evolution (IPTE) briefing is required to perform

___(1)___. The ___(2)___is responsible for ensuring proper performance of the IPTE Pre-job Brief.

(1) (2)

A. the initial Reactor Coolant System Drain Shift Manager Down to 18 feet MSL B. the initial Reactor Coolant System Drain Senior Line Manager Down to 18 feet MSL C. OP-903-007, Turbine Inlet Valve Shift Manager Cycling Test D. OP-903-007, Turbine Inlet Valve Senior Line Manager Cycling Test Revision 0 Facility: Waterford 3 Page 192 of 202

2012 NRC Exam RO Written Exam Worksheet Proposed Answer: B Explanation: (Optional)

A. Incorrect: EN-OP-116 states that Reactor Coolant System Drain Down to Lowered Inventory is an IPTE. Lowered inventory is less than 20 feet, therefore, drain down to 18 feet is an IPTE. EN-OP-116 Responsibilities section states that the Senior Line Manager is responsible for ensuring proper performance of the IPTE Pre-job Brief, not the Shift Manager.

B. CORRECT: EN-OP-116 states that Reactor Coolant System Drain Down to Lowered Inventory is an IPTE. Lowered inventory is less than 20 feet, therefore, drain down to 18 feet is an IPTE. EN-OP-116 Responsibilities section states that the Senior Line Manager is responsible for ensuring proper performance of the IPTE Pre-job Brief.

C. Incorrect: EN-OP-116 states that any test that actually over-speeds a turbine or Emergency Diesel Generator is an IPTE. OP-903-007 tests the cycling of Turbine valves, not the overspeed protection system, which is tested by OP-904-002, Main Turbine Overspeed Test. EN-OP-116 Responsibilities section states that the Senior Line Manager is responsible for ensuring proper performance of the IPTE Pre-job Brief, not the Shift Manager.

D. Incorrect: OP-903-007 does not overspeed the Main Turbine, and therefore is not an IPTE. EN-OP-116 Responsibilities section states that the Senior Line Manager is responsible for ensuring proper performance of the IPTE Pre-job Brief.

Technical Reference(s): EN-OP-116, Infrequently Performed Tests or (Attach if not previously provided) Evolutions, Revision 009 (including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-IPTE, Revision 2 (As available)

Objectives 1 and 4 Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 3 Comments:

Revision 0 Facility: Waterford 3 Page 193 of 202

2012 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 194 of 202

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 3 Group # 3 K/A # G2.2.17 Importance Rating 3.8 K/A Statement Equipment Control: Knowledge of the process for managing maintenance activities during power operations, such as risk assessments, work prioritization, and coordination with the transmission system operator.

Proposed Question: SRO 22 Rev: 0 The following plant conditions exist:

The plant is in MODE 5

Maintenance requests to remove Battery Charger A1 from service for scheduled work

Shutdown EOOS assigned an ORANGE risk level to the configuration resulting from removal of Battery Charger A1.

The ___(1)___ can authorize the request to remove Battery Charger A1 from service.

To remove Battery Charger A1 from service, a ___(2)___.

(1) (2)

A. Duty Plant Manager qualitative assessment is required B. Duty Plant Manager qualitative assessment is NOT required C. Shift Manager qualitative assessment is required D. Shift Manager qualitative assessment is NOT required Revision 0 Facility: Waterford 3 Page 195 of 202

2012 NRC Exam RO Written Exam Worksheet Proposed Answer: A Explanation: (Optional)

A. CORRECT: OI-037-000, Page 20, step 5.7.2.3 and bullet 1 state: Orange is considered high risk, with the following requirements: Duty Plant Manager approval for voluntary entry, or notification upon entry into emergent activities. Page 10, Step 5.2.1 states, Therefore, when one or more of these SSCs becomes unavailable due to emergent activities, an assessment which considers both the quantitative (EOOS tool) and qualitative (i.e., Level 2 and External Events) aspects of risk is performed.

B. Incorrect. DPM is the correct authority level, but a qualitative risk assessment is required per Step 5.2.1 above.

C. Incorrect. The Shift Manager is not the correct level of authority for Orange level risk assessment determination.

D. Incorrect. The Shift Manager is not the correct level of authority for Orange level risk assessment determination, and a qualitative risk assessment is required per Step 5.2.1 above.

Technical Reference(s): OI-037-000, Operations Risk Assessment (Attach if not previously provided) Guideline, Revision 304 (including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-ORA, Revision 4 (As available)

Objectives 2 & 3 Question Source: Bank # X WF3-OPS-08384 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam 2008 NRC Exam Question #96 SRO Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 3 Comments:

Revision 0 Facility: Waterford 3 Page 196 of 202

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 3 Group # 3 K/A # G2.3.14 Importance Rating 3.8 K/A Statement Radiation Control: Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or emergency conditions or activities.

Proposed Question: SRO 23 Rev: 0 The following plant conditions exist:

The plant is at 100% power

A Reactor Coolant Pump (RCP) 1A shaft seizure occurs

The reactor failed to trip automatically

The reactor was manually tripped using DRTS 30 seconds after the shaft seizure

Containment Purge Isolation monitors ARM-IRE-5024(5025)(5026)(5027) indicate rising activity levels

Containment PIG monitor PRM-IRE-0100Y readings remain unchanged In addition to OP-902-001, Reactor Trip Recovery Procedure, the CRS should enter

_______(1)_______, and direct _______(2)_______.

(1) (2)

A. OP-901-130, Reactor Coolant Pump evacuation of the Reactor Auxiliary Malfunction Building B. OP-901-410, High Activity in evacuation of the Reactor Auxiliary Reactor Coolant System Building C. OP-901-130, Reactor Coolant Pump securing of all Reactor Coolant Pumps Malfunction D. OP-901-410, High Activity in securing of all Reactor Coolant Pumps Reactor Coolant System Revision 0 Facility: Waterford 3 Page 197 of 202

2012 NRC Exam RO Written Exam Worksheet Proposed Answer: B Explanation: (Optional)

A. Incorrect. OP-901-130 does not address an RCP shaft seizure event. The RCA should be evacuated due to changing radiological conditions to prevent excess radiation exposure/dose.

B. CORRECT: Per FSAR Section 15.3.3.1.3.2, The single reactor coolant pump seized/sheared shaft may result in a minimum DNBR below the DNBR limit due to the rapid decrease in flowrate. This could result in some fuel damage and corresponding radiological releases. OP-901-410 addresses the activity released into the coolant released by the overheating of fuel prior to reactor shutdown. No indications are given of a breach of the RCS. The RCA is evacuated due to changing radiological conditions per step 3 of OP-901-410.

C. Incorrect: OP-901-130 does not address an RCP shaft seizure event. Forced circulation is desirable for RCS cleanup. All RCPs would be secured if an RCP sheared shaft event occurred.

D. Incorrect. OP-901-410 is the correct procedure. Forced circulation is desirable for RCS cleanup. All RCPs would be secured if an RCP sheared shaft event occurred.

Technical Reference(s): OP-901-410, High Activity in Reactor Coolant System, Revision 004 - Page 3 indications &

(Attach if not previously provided) Page 6, Step 3 (including version/revision number) FSAR Chapter 15, Section 15.3.3.1 Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PPE02, Objective 3 (As available)

Slide 31 Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 4 Comments:

Revision 0 Facility: Waterford 3 Page 198 of 202

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 3 Group # 4 K/A # G2.4.9 Importance Rating 4.2 K/A Statement Emergency Procedures / Plan: Knowledge of low power/shutdown implications in accident (e.g., loss of coolant accident or loss of residual heat removal) mitigation strategies.

Proposed Question: SRO 24 Rev: 0 The following plant conditions exist:

RCS temperature is 185°F and rising slowly

RCS level is 14 feet and steady.

Shutdown Cooling Train A is in service with flow 1800 to 2200 gpm and oscillating.

LPSI Pump A amps indicate 10 to 15 amps and oscillating.

The RCA watch reports that LPSI Pump A is running with a rattling sound.

The crew has entered OP-901-131, Shutdown Cooling Malfunction.

Based on these indications, which subsection in OP-901-131 needs to be entered?

A. E1: System Leakage B. E2: Loss of Shutdown Cooling Flow C. E3: Loss of Shutdown Cooling Heat Removal Capability D. E4: System Malfunction in MODE 4.

Revision 0 Facility: Waterford 3 Page 199 of 202

2012 NRC Exam RO Written Exam Worksheet Proposed Answer: B Explanation: (Optional)

A. Incorrect: A cavitating LPSI pump could be a sign of system leakage, but in this instance, RCS level is indicating steady.

B. CORRECT: The rattling sound and oscillating amps for LPSI Pump A are signs of cavitation. Section E2 of OP-901-131 is where the crew will find the actions to perform in the event of a cavitating LPSI pump.

C. Incorrect: The rattling sound and oscillating amps for LPSI Pump A are signs of cavitation. RCS temperature would rise rapidly on a loss of shutdown cooling heat removal capability. RCS temperature is rising slowly in this instance, which is expected for a cavitating LPSI pump.

D. Incorrect: The plant is not in mode 4 because RCS temperature is indicating 185°F.

Technical Reference(s): OP-901-131, Revision 302 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-REQ21, Revision 14 (As available)

Objectives 5 & 6 Question Source: Bank # X WF3-OPS-08873 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 55.43 5 Comments:

Revision 0 Facility: Waterford 3 Page 200 of 202

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 3 Group # 4 K/A # G2.4.28 Importance Rating 4.1 K/A Statement Emergency Procedures / Plan: Knowledge of procedures relating to a security event (non-safeguards information).

Proposed Question: SRO 25 Rev: 0 The NRC shall be notified within 15 minutes from the ______(1)_______ of a security-based event so that the NRC can provide prompt ________(2)_______.

(1) (2)

A. discovery notifications to other licensees and federal agencies B. discovery assistance in obtaining necessary resources C. E-Plan classification notifications to other licensees and federal agencies D. E-Plan classification assistance in obtaining necessary resources Revision 0 Facility: Waterford 3 Page 201 of 202

2012 NRC Exam RO Written Exam Worksheet Proposed Answer: A Explanation: (Optional)

A. CORRECT: Per Note on Page 8 of OP-901-523, Security Events, The NRC Headquarters Operations Center shall be notified within 15 minutes of discovery of a security-based event to allow the NRC to more quickly notify other licensees and Federal agencies.

B. Incorrect: Within 15 minute of discovery is the required notification time, but the NRC is not responsible for obtaining necessary resources for security-based events.

C. Incorrect: The NRC Headquarters Operations Center (HOC) shall be notified within 15 minutes of discovery of a security-based event to allow the NRC to more quickly notify other licensees and Federal agencies. Delaying until E-Plan classification and notifications are complete could delay NRC HOC notifications by as much as 60 minutes.

D. Incorrect: The NRC Headquarters Operations Center (HOC) shall be notified within 15 minutes of discovery of a security-based event to allow the NRC to more quickly notify other licensees and Federal agencies. Delaying until E-Plan classification and notifications are complete could delay NRC HOC notifications by as much as 60 minutes. The NRC is not responsible for obtaining necessary resources for security-based events.

Technical Reference(s): OP-901-523, Security Events, Revision 010 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PPO52, (As available)

Objectives 2 & 3 Question Source: Bank # X WF3-OPS-06733 Modified Bank #

New Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 5 Comments:

Revision 0 Facility: Waterford 3 Page 202 of 202

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