ML15225A522

2015-09 Draft Written Exam and Revised Outlines

ML15225A522
Person / Time
Site:	Waterford
Issue date:	09/14/2015
From:	Vincent Gaddy Operations Branch IV
To:
References
Download: ML15225A522 (227)

Text

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 1 of 150 Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 000007 (CE/E02)EK2.2 Importance Rating 3.5 K/A Statement EK2.2: Knowledge of the interrelations between the (Reactor Trip Recovery) and the following: Facility's heat removal systems, including primary coolant, emergency coolant, the decay heat removal systems, and relations between the proper operation of these systems to the operation of the facility.

Proposed Question:

RO 1 Rev: 0 Given: Plant is at 100% power SG ILR 1111, Steam Generator 1 Downcomer Level, fails high Crew has entered OP-901-201, Steam Generator Level Control Malfunction Upon a reactor trip, Main Feed Pump A speed will reduce automatically. This design feature was added to reduce the differential pressure across the ____(1)____ valves. Upon subsequent entry into OP-902-001, Reactor Trip Recovery procedure, the Main Feed Pump A speed controller (FW IHIC1107) will be controlled in the _____(2)_____ mode. (1) (2) A. Main Feed Regulating automatic B. Main Feed Isolation automatic C. Main Feed Regulating manual D. Main Feed Isolation manual 2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 2 of 150 Proposed Answer: D Explanation: (Optional)

A. Incorrect: The main feed regulating valve is plausible because it also receives a close signal on MSIS and a reactor trip override (RTO), but is not the reason the design feature was added. Main Feed Pump A speed controller will swap to auto upon the reactor trip for 5 seconds. Because there is a level deviation in this instance the controller will swap back to manual. B. Incorrect. Part 1 is correct. Main Feed Pump A speed controller will swap to auto upon the reactor trip for 5 seconds. Because there is a level deviation in this instance the controller will swap back to manual. C. Incorrect. The main feed regulating valve is plausible because it also receives a close signal on MSIS and a reactor trip override, but is not the reason the design feature was added. Part 2 is correct.

D. CORRECT: On a reactor trip override (RTO), FWCS will swap the Main Feed Pump speed controllers to auto and drive them to minimum speed if the controller was in manual. If the controller was in manual due to a level deviation, which in this case it is, the controller will still swap to auto but will return to manual after 5 seconds. As stated in the basis for TS 3.7.6.1, this design feature is to ensure rapid closure of a MFIV on a MSIS by ensuring DP across the valve is reduced.

Technical Reference(s): TS 3.7.1.6 basis (Attach if not previously provided) SD-FWC pp.24-25 (Rev 13) (including version/revision number)

Proposed references to be provided to applicants during examination:

None Learning Objective: WLP-OPS-FWC00 obj. 6 (As available)

WLP-OPS-FW obj. 8 Question Source: Bank #

Modified Bank #

(Note changes or attach parent)

New X Question History: Last NRC Exam None Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Comments:

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 3 of 150 Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 000008 AA2.15 Importance Rating 3.9 K/A Statement AA2.15: Ability to determine and interpret the following as they apply to the Pressurizer Vapor Space Accident:

ESF control board, valve controls, and indicators.

Proposed Question:

RO 2 Rev: 0 The plant was at 100% power when Pressurizer Safety Valve, RC-317A, failed open. The crew tripped the reactor and initiated SIAS and CIAS.

The following conditions exist:

RCS pressure is 1340 psia and stable Reactor Coolant Pumps have been secured CET is 556°F, Thot is 550°F, Tcold is 543°F (all stable) Pressurizer level is 70% and rising Quench Tank pressure is 20 psig QSPDS indicates reactor head level 1 voided Steam Generator (SG) 1 and 2 pressures are 980 psia and stable SG 1 and 2 levels are 60% WR and slowly lowering The expected temperature indication of the Pressurizer Relief Line is ____(1)____ °F. The crew ______(2)______ throttle close Train B High Pressure Safety Injection flow control valves.

(1) (2) A. 258 will not B. 228 will not C. 258 will D. 228 will 2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 4 of 150 Proposed Answer: A Explanation: (Optional)

A. CORRECT: Quench tank pressure is 20 psig. This pressure must be changed to absolute pressure to determine relief line temperature. The saturation temperature for 35 psia is 258°F. The crew should not close HPSI flow control valves because subcooled margin based on CET is less than 28F The applicant will determine that a leak exists in the pressurizer steam space. Pressurizer level will rise in this event due to RCS voiding and HPSI flow should not be throttled based solely on pressurizer level. B. Incorrect. Quench tank pressure is 20 psig. This pressure must be changed to absolute pressure to determine relief line temperature. The saturation temperature for 35 psia is 258°F. The distractor is the saturation temperature for 20 psia. Second part is correct. C. Incorrect. First part is correct. Second part is plausible because pressurizer level is rising rapidly and water solid operations should be avoided unless minimum subcooling (28F) cannot be maintained. D. Incorrect: Quench tank pressure is 20 psig. This pressure must be changed to absolute pressure to determine relief line temperature. The saturation temperature for 35 psia is 258°F. The distractor is the saturation temperature for 20 psia. Second part is plausible because pressurizer level is rising rapidly and water solid operations should be avoided unless minimum subcooling (28F) cannot be maintained.

Technical Reference(s): SD-RCS page 42 rev. 20 (Attach if not previously provided) OI-038-000 step 5.4.31 revision 10 (including version/revision number) Steam Tables, OP-902-002 step 23 rev. 19 Proposed references to be provided to applicants during examination: Steam Tables 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 None Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 3,8 55.43 Comments:

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 5 of 150 Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 000009 EA2.11 Importance Rating 3.8 K/A Statement EA2.11: Ability to determine or interpret the following as they apply to a small break LOCA: Containment temperature, pressure, and humidity.

Proposed Question:

RO 3 Rev: 0 Given: Small Break LOCA has occurred and the crew has diagnosed to OP-902-002, LOCA Recovery Procedure Pressurizer pressure is 1690 PSIA and slowly lowering Containment pressure is 17.4 psia and slowly rising Containment temperature is 210°F Which of the following ESFAS signal(s) will have been generated?

A. MSIS only B. SIAS and CIAS only C. SIAS, CIAS, and MSIS only D. SIAS, CIAS, MSIS, and CSAS

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 6 of 150 Proposed Answer: C Explanation: (Optional)

A. Incorrect. MSIS is present due to containment pressure above 17.1 PSIA. SIAS and CIAS occur at 1684 PSIA or 17.1 PSIA in containment. Applicant must be aware that PZR pressure is above SIAS and CIAS setpoint but SIAS and CIAS are still in due to containment pressure. B. Incorrect. SIAS and CIAS occur at 1684 PSIA or 17.1 PSIA in containment. MSIS is also present due to containment pressure above 17.1 PSIA.

C. CORRECT: SIAS, CIAS, and MSIS all actuate at a containment pressure of 17.1 PSIA. A harsh environment exists in containment, the applicant must determine that the containment pressure instruments are not affected by containment temperature above 200°F. Many instruments in containment are affected by the harsh environment, but the pressurizer pressure nor containment pressure instruments that initiate ESFAS signals are affected. The applicant may determine that the instruments are affected and determine the wrong ESFAS initiations. Waterford does not use containment humidity in any of its procedures. D. Incorrect. CSAS does not occur until 17.7 PSIA.

Technical Reference(s): TS 2.2.1 & 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-PPS00 obj. 1, 5 (As available)

Question Source: Bank #

Modified Bank #

X (Note changes or attach parent)

New Question History: Last NRC Exam None Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 4,7 55.43 Comments: Modified from 2012 NRC Exam - RO 39

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 7 of 150 Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 000011 2.4.50 Importance Rating 4.2 K/A Statement 2.4.50: Ability to verify system alarm setpoints and operate controls identified in the alarm response manual.

Proposed Question:

RO 4 Rev: 0 Given: A LOCA has occurred RCS pressure is 100 PSIA The crew is performing OP-902-002, Loss of Coolant Accident Recovery The following alarms are received on Control Room Cabinet K:

o K-19, RAS TRAIN A LOGIC INITIATED o K-20, RAS TRAIN B LOGIC INITIATED The crew will verify that Refueling Water Storage Pool level is _____(1)______ percent. If required, a manual initiation of an RAS can be performed _____(2)______. (1) (2) A. 10 from the local ESFAS cabinet only B. 10 from CP-7 or local ESFAS cabinet C. 15 from the local ESFAS cabinet only D. 15 from CP-7 or local ESFAS cabinet

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 8 of 150 Proposed Answer: A Explanation: (Optional)

A. CORRECT: The annunciator response procedure OP-500-009 (Att. 4.89 and 4.90) indicates RAS Train A(B) Logic Initiated alarms annunciate at 10% RWSP level. An RAS can only be initiated at the local ESFAS cabinet. It is the only ESFAS that cannot be initiated from CP-7. B. Incorrect. The annunciator response procedure OP-500-009 (Att. 4.89 and 4.90) indicates RAS Train A(B) Logic Initiated alarms annunciate at 10% RWSP level. An RAS can only be initiated at the local ESFAS cabinet. It is the only ESFAS that cannot be initiated from CP-7. C. Incorrect. An RWSP level of 15% is credible because the RWSP level Lo pretrip will annunciate in the control room at this level. An RAS can only be initiated at the local ESFAS cabinet. D. Incorrect. An RWSP level of 15% is credible because the RWSP level Lo pretrip will annunciate in the control room at this level. An RAS can only be initiated at the local ESFAS cabinet. It is the only ESFAS that cannot be initiated from CP-7.

Technical Reference(s): OP-500-009 Att. 4.89 and 4.90 rev 5 (Attach if not previously provided) SD-PPS pg 52 rev. 16 (including version/revision number)

Proposed references to be provided to applicants during examination:

None Learning Objective: WLP-OPS-PPS00 obj 1 and 16 (As available)

Question Source: Bank #

Modified Bank #

(Note changes or attach parent)

New X Question History: Last NRC Exam None Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Comments:

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 9 of 150 Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 000015/17 AA1.03 Importance Rating 3.7 K/A Statement AA1.03: Ability to operate and / or monitor the following as they apply to the Reactor Coolant Pump Malfunctions (Loss of RC Flow): Reactor trip alarms, switches, and indicators.

Proposed Question:

RO 5 Rev: 0 Given: The reactor was manually tripped due to 1A RCP ARRD High Temperature 1A RCP has been secured Five minutes later, the ATC reports that the RCP PMC Mimic indicates RCP 1A speed is 600 rpm Per OP-901-130, Reactor Coolant Pump Malfunction, the crew is required to trip the reactor if ARRD temperature exceeds ____(1)____ °F. Based on the above conditions, the crew will ______(2)______. (1) (2) A. 225 continue to monitor RCP 1A for reverse rotation B. 210 continue to monitor RCP 1A for reverse rotation C. 225 remove all reactor coolant pumps from service D. 210 remove all reactor coolant pumps from service

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 10 of 150 Proposed Answer: C Explanation: (Optional)

A. Incorrect. Part 1 is correct. With the given conditions, RCP 1A is reverse rotating even though speed on the RCP mimic is indicating a positive number. This guidance on RCP speed indication is given in a note in the Reverse rotation section of OP-901-130. If a RCP is reverse rotating, all RCPs must be secured. B. Incorrect. Per OP-901-130, the crew is required to trip the reactor and secure the affected RCP when ARRD temperature exceeds 225 °F. At 210°F, a plant shutdown must be commenced. With the given conditions, RCP 1A is reverse rotating.

C. CORRECT: Per OP-901-130, the crew is required to trip the reactor and secure the affected RCP when ARRD temperature exceeds 225 °F. With the given conditions, RCP 1A is reverse rotating even though speed on the RCP mimic is indicating a positive number. This guidance on RCP speed indication is given in a note in the Reverse rotation section of OP-901-130. If a RCP is reverse rotating, all RCPs must be secured. D. Incorrect. Per OP-901-130, the crew is required to trip the reactor and secure the affected RCP when ARRD temperature exceeds 225 °F. At 210°F, a plant shutdown must be commenced. Part 2 is correct.

Technical Reference(s): OP-901-130 section E5 and E6 revision 8 (Attach if not previously provided) (including version/revision number)

Proposed references to be provided to applicants during examination:

None Learning Objective: WLP-OPS-PPO10 obj. 3 (As available)

Question Source: Bank #

Modified Bank #

X (Note changes or attach parent)

New Question History: Last NRC Exam 2008 RO NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 Comments: Modified from 2010 NRC Exam - RO1

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 11 of 150 Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 000022 AA1.01 Importance Rating 3.4 K/A Statement AA1.01: Ability to operate and / or monitor the following as they apply to the Loss of Reactor Coolant Makeup: CVCS Letdown and Charging Proposed Question:

RO 6 Rev: 0 Given: Plant is at 50% power A loss of all Charging Pumps has occurred The crew will verify closed ____(1)____ in accordance with ______(2)______. (1) (2) A. Letdown Inside Containment Isolation, CVC-103 OP-901-110, Pressurizer Level Control Malfunction B. Letdown Stop Valve, CVC-101 OP-901-110, Pressurizer Level Control Malfunction C. Letdown Stop Valve, CVC-101 OP-901-112, Charging or Letdown Malfunction D. Letdown Inside Containment Isolation, CVC-103 OP-901-112, Charging or Letdown Malfunction

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 12 of 150 Proposed Answer: C Explanation: (Optional)

A. Incorrect At 470°F Regen HX outlet temperature, CVC-101 gets a closed signal. CVC-103 is located downstream of CVC-101 and will close on a SIAS and CIAS, not RHX outlet temp. OP-901-110 is plausible because pzr level will be affected but the crew will find no guidance there for a loss of all charging pumps. B. Incorrect. Part 1 is correct. This guidance is located in OP-901-112, Charging or Letdown Malfunction. OP-901-110 is plausible because pzr level will be affected but the crew will find no guidance there for a loss of all charging pumps.

C. CORRECT: A loss of all charging pumps will mean no cooling to the Regen HX. At 470°F Regen HX outlet temperature, CVC-101 gets a closed signal. This guidance is located in OP-901-112, Charging or Letdown Malfunction.

D. Incorrect. At 470°F Regen HX outlet temperature, CVC-101 gets a closed signal. CVC-103 is located downstream of CVC-101 and will close on a SIAS and CIAS, not RHX outlet temp. Part 2 is correct.

Technical Reference(s): OP-901-112 section E1 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-PPO10 obj. 3 (As available)

Question Source: Bank #

Modified Bank #

(Note changes or attach parent)

New X Question History: Last NRC Exam None Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Comments:

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 13 of 150 Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 000025 2.1.30 Importance Rating 4.4 K/A Statement 2.1.30: Ability to locate and operate components, including local controls.

Proposed Question:

RO 7 Rev: 0 Given: The plant is in Mode 5 with Shutdown Cooling Train A in service A loss of Instrument Air has occurred SI-129A, LPSI Pump A Discharge Flow Control, is not accessible due to plant conditions The crew will direct the Operator to force closed SI-129A using a (an) ____(1)____ accumulator. Manipulation of the required 3-way valve will be performed ______(2)______. (1) (2) A. nitrogen gas from the -35 hallway outside of Safeguards Room A B. instrument air from the -35 hallway outside of Safeguards Room A C. instrument air from the RAB -15 Valve Gallery D. nitrogen gas from the RAB -15 Valve Gallery

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 14 of 150 Proposed Answer: C Explanation: (Optional)

A. Incorrect. Nitrogen gas is plausible because many W3 valves backup air supply comes from Nitrogen accumulators. The location of outside of Safeguards Room A is plausible because SI-129A is in the safeguards room, but the operator would not have to enter the room to operate the valve. B. Incorrect. Part 1 is correct. The location of outside of Safeguards Room A is plausible because SI-129A is in the safeguards room, but the operator would not have to enter the room to operate the valve.

C. CORRECT: SI-129A will fail open on a loss of instrument air. If local access to SI-129A is not available, OP-901-511 directs the crew to force closed SI-129A in accordance with OP-009-008, Safety Injection System. The motive force is from an Instrument Air accumulator and the 3-way valve to perform the manipulation is in the -15 Valve Gallery. D. Incorrect. Nitrogen gas is plausible because many W3 valves backup air supply comes from Nitrogen accumulators. Part 2 is correct.

Technical Reference(s): OP-009-008 section 8.17 revision 37 (Attach if not previously provided) OP-003-016 attachment 11.1 page 94 revision 20 (including version/revision number)

Proposed references to be provided to applicants during examination:

None Learning Objective: WLP-OPS-SI00 obj. 6 (As available)

Question Source: Bank #

Modified Bank #

(Note changes or attach parent)

New X Question History: Last NRC Exam None Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7, 10 55.43 Comments:

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 15 of 150 Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 000026 AK3.04 Importance Rating 3.5 K/A Statement AK3.04: Knowledge of the reasons for the following responses as they apply to the Loss of Component Cooling Water: Effect on the CCW flow header of a loss of CCW. Proposed Question:

RO 8 Rev: 0 Given: CCW Surge Tank Level Switch CC-ILS-7013A failed low CC-200A/727, CCW Hdr A TO AB Supply and Return Isolations, failed closed CC-134 A, CCW A Dry Cooling Tower Bypass failed open CC-135 A, CCW A Dry Cooling Tower Isolation, failed closed Both Train A and Train B CC-620, Fuel Pool Heat Exchanger Temperature Control Valve, control switches are in the CONTROL position at CP-8 Based on this level switch failure, A. Spent Fuel Pool temperature will rise since CC-620 failed closed. B. Spent Fuel Pool temperature will lower since CC-134 A, CCW A Dry Cooling Tower Bypass failed open. C. Spent Fuel Pool temperature will rise since CC-135 A, CCW A Dry Cooling Tower Isolation, failed closed. D. Spent Fuel Pool temperature is controlled in automatic since the Train B CC-620 control switch is in CONTROL.

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 16 of 150 Proposed Answer: A Explanation: (Optional)

A. CORRECT: CC-620 Fuel Pool HX Temperature control valve goes closed on low CCW surge tank level on either side. CC-620 has closed since CC-200A has closed due to the level switch failure. CC-620 is interlocked with C-200A. This will cause Spent fuel pool temperature to rise due to no CCW flow to the Spent Fuel Pool HX. B. Incorrect: CC-134 A does open if this level switch fails, but ACCW Pump A will auto start and control CCW Train A temperature. SFP temperature rises because of CC-620 closing. The applicant may assume that there is greater flow to the SFP HX since the bypass around the heat exchanger has opened. C. Incorrect. CC-135 A does close if this level switch fails, but ACCW Pump A will auto start and control CCW Train A temperature. SFP temperature rises because valve CC-620 closes. D. Incorrect. Even with the Train B control Switch for CC-620 in Control, CC-620 will still close due to the failure of the Train A level switch. CC-620 has an A and B solenoid valve in series.

Technical Reference(s): OP-901-510 Attachment 1 revision 303 (Attach if not previously provided) SD-CC pg. 37 revision 21 (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 #

X Question 61 Modified Bank #

(Note changes or attach parent)

New Question History: Last NRC Exam 2011 NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Comments:

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 17 of 150 Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 000027 AA2.11 Importance Rating 4.0 K/A Statement AA2.11: Ability to determine and interpret the following as they apply to the Pressurizer Pressure Control Malfunctions: RCS pressure.

Proposed Question:

RO 9 Rev: 0 Given: The plant is at 100% power Control Systems are in normal 100% power alignments The Pressurizer Pressure Controller, RC-IPIC-0100, output fails HIGH Assuming no action by the crew, a reactor trip will occur due to ____(1)____ Pressurizer Pressure generated by ______(2)______. (1) (2) A. high PPS B. high CPCs C. low PPS D. low CPCs 2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 18 of 150 Proposed Answer: D Explanation: (Optional)

A. Incorrect. If no action taken with a failed high pressure controller, PZR pressure will lower because controller will open pressurizer spray valves. There is a high pressurizer pressure trip associated with PPS B. Incorrect. If no action taken with a failed high pressure controller, PZR pressure will lower because controller will open pressurizer spray valves. There is a high pressurizer pressure trip associated with CPCs C. Incorrect. If no action taken with a failed high pressure controller, PZR pressure will lower because controller will open pressurizer spray valves. CPCs will generate a trip at 1860 psia prior to PPS trip on low pressure at 1684 psia D. CORRECT: If no action taken with a failed high pressure controller, PZR pressure will lower because controller will open pressurizer spray valves. CPCs will generate a trip at 1860 psia prior to PPS trip on low pressure at 1684 psia.

Technical Reference(s): OP-901-120 section E2 revision 302 (Attach if not previously provided) TS Tbl. 2.2-1 & Bases; OP-004-006 pg 29 rev 305 (including version/revision number) SD-PLC pg 29 rev 9 Proposed references to be provided to applicants during examination:

None Learning Objective: WLP-OPS-PPO10 obj. 4 (As available)

Question Source: Bank #

X Question 46 Modified Bank #

(Note changes or attach parent)

New Question History: Last NRC Exam 2007 RO Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Comments:

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 19 of 150 Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 000029 EK1.03 Importance Rating 3.6 K/A Statement EK1.03: Knowledge of the operational implications of the following concepts as they apply to the ATWS: Effects of boron on reactivity.

Proposed Question: RO 10 Rev: 0 Given: Reactor power is 1% A large steam leak upstream of MSIV B prompts the crew to trip the reactor A manual reactor trip using CP-2 pushbuttons was unsuccessful The crew tripped the reactor using the Diverse Reactor Trip System (DRTS) All CEAs have inserted into the core The DRTS de-energized CEDM cabinets by opening the MG set ____(1)____. Boron injection ______(2)______. (1) (2) A. supply breakers is required to maintain shutdown margin B. supply breakers is not required because all CEAs have inserted C. load contactors is required to maintain shutdown margin D. load contactors is not required because all CEAs have inserted 2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 20 of 150 Proposed Answer: C Explanation: (Optional)

A. Incorrect: DRTS opens the Motor-Generator Output contactors to de-energize CEDMCS. Second part is correct. B. Incorrect: DRTS opens the Motor-Generator Output contactors to de-energize CEDMCS. OP-902-000, Standard Post Trip Action does not direct the operator to emergency borate when all CEAs insert into the core however, OP-901-103, Emergency Boration directs the operators to initiate emergency boration in response to an uncontrolled cooldown. C. CORRECT: DRTS opens the Motor-Generator Output contactors to de-energize CEDMCS. The applicant must recognize that an uncontrolled cooldown is in progress and therefore emergency boration is required. OP-901-103, Emergency Boration directs the operators to initiate emergency boration in response to an uncontrolled cooldown.

D. Incorrect: First part is correct. OP-902-000, Standard Post Trip Action does not direct the operator to emergency borate when all CEAs insert into the core however, OP-901-103, Emergency Boration directs the operators to initiate emergency boration in response to an uncontrolled cooldown.

Technical Reference(s): OP-901-103 pg. 2-3, revision 3 (Attach if not previously provided) OP-902-000 pg. 5, revision 15 (including version/revision number) SD-ATS pg. 9, revision 5 Proposed references to be provided to applicants during examination:

None Learning Objective: WLP-OPS-PPO10 obj. 1,2 (As available)

WLP-OPS-PPE01 obj. 9,10 Question Source: Bank #

Modified Bank #

(Note changes or attach parent)

New X Question History: Last NRC Exam None Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 Comments:

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 21 of 150 Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 000038 EA1.39 Importance Rating 3.6 K/A Statement EA1.39: Ability to operate and monitor the following as they apply to a SGTR:

Drawing S/G into the RCS, using the "feed and bleed" method.

Proposed Question: RO 11 Rev: 0 Given: A Steam Generator Tube Rupture (SGTR) has occurred in Steam Generator (SG) #2 The crew has entered OP-902-007, Steam Generator Tube Rupture Recovery Procedure, and has isolated SG #2 Steam Generator #2 level is 80% NR The CRS directs lowering SG #2 level to 65% NR using back flow to the RCS. RCS boron concentration is presently 1210 ppm The CRS directs the ATC to calculate final RCS boron concentration for the SG #2 level reduction Prior to lowering SG#2 level using backflow, the ATC will verify ____(1)____. Final RCS boron concentration will be approximately ______(2)______ ppm.

(1) (2) A. natural circulation conditions are being met 1101 B. at least one RCP is running 907 C. at least one RCP is running 1101 D. natural circulation conditions are being met 907 2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 22 of 150 Proposed Answer: C Explanation: (Optional)

A. Incorrect: At least one RCP operating is required before cooling and depressurizing the isolated SG using the feed and bleed method. Natural circulation conditions are verified in this ORP, but is not a prerequisite for feeding and bleeding a SG. The note before step 50 indicates that RCS boron concentration will lower by 0.6% for each 1% drop in steam generator NR level. B. Incorrect. Step 50 of OP-902-007 directs the crew to verify at least one RCP operating before cooling and depressurizing the isolated SG using the feed and bleed method. If the crew were to use a 1% drop in RCS boron concentration for each 0.6% drop in steam generator level they would come up with this wrong answer. C. CORRECT: Step 50 of OP-902-007 directs the crew to verify at least one RCP operating before cooling and depressurizing the isolated SG using the feed and bleed method. The note before step 50 indicates that RCS boron concentration will lower by 0.6% for each 1% drop in steam generator NR level. D. Incorrect. Step 50 of OP-902-007 directs the crew to verify at least one RCP operating before cooling and depressurizing the isolated SG using the feed and bleed method. If the crew were to use a 1% drop in RCS boron concentration for each 0.6% drop in steam generator level they would come up with this wrong answer. Technical Reference(s): OP-902-007 step 50, revision 16 (Attach if not previously provided) (including version/revision number)

Proposed references to be provided to applicants during examination:

None Learning Objective: WLP-OPS-PPE07 obj. 8 (As available)

Question Source: Bank #

Modified Bank #

(Note changes or attach parent)

New X Question History: Last NRC Exam None Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 Comments:

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 23 of 150 Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 000040 CE/E05 EK2.1 Importance Rating 3.3 K/A Statement CE/E05, EK2.1: Knowledge of the interrelations between the (Excess Steam Demand) 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 An Excess Steam Demand event is in progress and OP-902-004, Excess Steam Demand Recovery is being implemented.

RCS Subcool Margin 80ºF and stable SG 1 pressure is 540 psia and lowering SG 2 pressure is 670 psia and stable SG 1 level 45% WR and lowering SG 2 level 60% WR and lowering slowly Containment Pressure 25 psia and rising Containment Spray Pump A is tagged out with CS-125A gagged shut Containment Spray Header B flow is 1000 gpm CVC-101, Letdown Stop Valve, is stuck open EFW-228A, Emergency Feedwater Isolation SG 1 Primary is tagged closed All other systems and components are operating as designed Based on this, which Safety Function is NOT being met?

A. RCS Pressure Control B. RCS Heat Removal C. Containment Isolation D. Containment Temperature and Pressure Control

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 24 of 150 Proposed Answer: D Explanation: (Optional)

A. Incorrect. RCS pressure Control can be met by LPSI and HPSI pumps providing flow > OP-902-009 Appendix 2 curves. This is implied by the last bullet in the conditions given. B. Incorrect. RCS Heat Removal. Although S/G 2 level is lowering the final bullet implies that that EFAS-2 actuated per design and EFW is operating per design. The level given would not be low enough yet for EFW to feed S/G 2. Operators are trained to evaluate this safety function as SAT if EFW is automatic and working as designed. Tc can also be implied to be lowering during the blowdown phase of an excess steam demand. The conditions given imply that information. C. Incorrect. CVC-101, Letdown Stop Valve, is not a containment isolation valve. This valve gets an automatic closure signal on a high Regen HX outlet temperature and is located just upstream of the CVC inside containment isolation valve, CVC-103.

D. CORRECT: At least one Containment Spray pump must be providing > 1750 gpm of flow to meet the CT&PC safety function.

Technical Reference(s): OP-902-004 pg. 42 revision 15 (Attach if not previously provided) (including version/revision number)

Proposed references to be provided to applicants during examination:

None Learning Objective: WLP-OPS-PPE04 obj. 6 (As available)

Question Source: Bank #

X Question #49 Modified Bank #

(Note changes or attach parent)

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

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 25 of 150 Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 000054 (CE/E06) EK3.2 Importance Rating 3.2 K/A Statement CE/E06, EK3.2: Knowledge of the reasons for the following responses as they apply to the (Loss of Feedwater): Normal, abnormal and emergency operating procedures associated with (Loss of Feedwater).

Proposed Question: RO 13 Rev: 0 Given: A Loss of all Main Feedwater has occurred The crew has entered OP-902-006, Loss of Main Feedwater Recovery Procedure EFW Pump AB is the only EFW pump available The crew will secure ____(1)____. The reason for this response is to ______(2)______. (1) (2) A. one Reactor Coolant Pump in each loop reduce heat input into the RCS B. one Reactor Coolant Pump in each loop prevent operating RCPs without adequate NPSH C. all Reactor Coolant Pumps reduce heat input into the RCS D. all Reactor Coolant Pumps prevent operating RCPs without adequate NPSH

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 26 of 150 Proposed Answer: A Explanation: (Optional)

A. CORRECT: One RCP in each loop is secured upon entry into OP-902-006 (step 6). The tech guide for securing a reactor coolant pump in each loop states the objective is to reduce heat input into the RCS. B. Incorrect. One RCP in each loop is secured upon entry into OP-902-006 (step 6). NPSH will be reduced in this instance, but not enough to require tripping RCPs. RCS pressure will not be low enough. Credible because there are regions in the RCP operating curves where 4-pump operation is not allowed and one RCP in each loop is secured. C. Incorrect. One RCP in each loop is secured upon entry into OP-902-006 (step 6). Step 7 of OP-902-006 would have all RCPs secured if the only available EFW pump is one Motor Driven EFW pump. In this case the turbine driven EFW pump is available which has a higher capacity than the motor driven EFW pump. Part 2 is correct. D. Incorrect. Step 7 of OP-902-006 would have all RCPs secured if the only available EFW pump is one Motor Driven EFW pump. In this case the turbine driven EFW pump is available which has a higher capacity than the motor driven EFW pump. NPSH will be reduced in this instance, but not enough to require tripping RCPs. RCS pressure will not be low enough. Credible because there are regions in the RCP operating curves where 4-pump operation is not allowed and one RCP in each loop is secured.

Technical Reference(s): OP-902-006 pg. 6, revision 16 (Attach if not previously provided) TG-OP-902-006 pg. 16, revision 16 (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 #

X (Note changes or attach parent)

New Question History: Last NRC Exam 2008 RO Exam Question 75 Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 8 55.43 Comments:

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 27 of 150 Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 000055 2.2.37 Importance Rating 3.6 K/A Statement 2.2.37: Ability to determine operability and/or availability of safety related equipment.

Proposed Question: RO 14 Rev: 0 Given: At 1200, a reactor trip occurred due to a station blackout At 1210, heavy rain started and "Dry Cooling Tower Sump 1 and 2 Level HI" annunciators have illuminated marking the beginning of a Probable Maximum Precipitation (PMP) event At 1215, the crew started Emergency Diesel Generator (EDG) A and EDG A is powering the 3A bus The CRS directs the auxiliary operator to restore operation of the DCT motor driven Sump Pumps.

Which ONE of the following describes the clock time limit to accomplish these actions and the reason for those actions?

A. 1240; prevent flooding of MCC 315A and its Transformer only.

B. 1240; prevent flooding of MCCs 315A and 315B and their Transformers.

C. 1230; prevent flooding of MCC 315A and its Transformer only.

D. 1230; prevent flooding of MCCs 315A and 315B and their Transformers.

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 28 of 150 Proposed Answer: B Explanation: (Optional)

A. Incorrect. Part 1 is correct: Aligning the DCT sump pumps at MCC-314A will energize one DCT sump pump for both the East and West DCT (#1 and #2) sumps. The applicant could assume that both sump pumps aligned at the 314A bus are for the 315A bus. B. CORRECT: The 30 minute clock for aligning DCT sump pumps for operation begins at the time of the PMP event. Aligning the DCT sump pumps at MCC-314A only will energize one DCT sump pump for both the East and West DCT (#1 and #2) sumps. Each train would have one operable sump pump. C. Incorrect. The 30 minute clock for aligning DCT sump pumps for operation begins at the time of the PMP event (not at the time of the SBO). Aligning the DCT sump pumps at MCC-314A will energize one DCT sump pump for both the East and West DCT (#1 and #2) sumps. The Tech Guide for step 5 indicates that "Dry Cooling Tower Sump 1 and 2 Level HI" annunciators are a good indication of a PMP event in progress. D. Incorrect. The 30 minute clock for aligning DCT sump pumps for operation begins at the time of the PMP event (not at the time of the SBO). Part 2 is correct.

Technical Reference(s): OP-902-009 Appendix 20 revision 310 (Attach if not previously provided) TG OP-902-005 step 5 revision 307 (including version/revision number) OP-902-005 step 5 revision 18 Proposed references to be provided to applicants during examination:

None Learning Objective: WLP-OPS-PPE01 obj. 3 (As available)

Question Source: Bank #

Modified Bank #

(Note changes or attach parent)

New X Question History: Last NRC Exam None Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 Comments:

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 29 of 150 Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 000056 AK1.03 Importance Rating 3.1 K/A Statement AK1.03: Knowledge of the operational implications of the following concepts as they apply to Loss of Offsite Power: Definition of subcooling: use of steam tables to determine it.

Proposed Question: RO 15 Rev: 0 Given: A Loss of Offsite Power has occurred 30 minutes ago The crew has entered OP-902-003, Loss of Offsite Power Recovery Procedure RCS pressure is 1725 psia and slowly rising Thot is 595°F and constant CET temperature is 600°F and constant Tcold is 590°F and constant The present subcooling margin is ____(1)____ °F. The crew will verify the operating loop steam generator pressure is approximately equal to saturation pressure for the existing ______(2)______. (1) (2) A. 15 Thot B. 15 Tcold C. 25 Tcold D. 25 Thot 2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 30 of 150 Proposed Answer: B Explanation: (Optional)

A. Incorrect. Part 1 is correct. The contingency for not meeting natural circ conditions is to verfify operating loop steam generator pressure is approximately equal to saturation pressure for the existing T cold, not Thot. B. CORRECT: Per OI-038-000, with no RCPs running, the crew is required to use CET temperature to determine subcooling. Using a steam table and RCS pressure, the applicant will determine that subcooling is presently 15°F. The contingency for not meeting natural circ conditions is to verify operating loop steam generator pressure is approximately equal to saturation pressure for the existing T cold. C. Incorrect. Per OI-038-000, with no RCPs running, the crew is required to use CET temperature to determine subcooling. Using a steam table and RCS pressure, the applicant would incorrectly determine that subcooling is 25°F if Tcold was used. Part 2 is correct. D. Incorrect. Per OI-038-000, with no RCPs running, the crew is required to use CET temperature to determine subcooling. Using a steam table and RCS pressure, the applicant would incorrectly determine that subcooling is 25°F if Tcold was used. The contingency for not meeting natural circ conditions is to verify operating loop steam generator pressure is approximately equal to saturation pressure for the existing T cold, not Thot.

Technical Reference(s): OP-902-003 step 15, revision 9 (Attach if not previously provided) OI-038-000 step 5.2.4 revision 10 (including version/revision number) Steam Tables Proposed references to be provided to applicants during examination: Steam Tables 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 None Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 Comments:

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 31 of 150 Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 000058 AK1.01 Importance Rating 2.8 K/A Statement AK1.01: Knowledge of the operational implications of the following concepts as they apply to Loss of DC Power: Battery charger equipment and instrumentation.

Proposed Question: RO 16 Rev: 0 Given: A Loss of all Offsite Power has occurred The crew has entered OP-902-003, Loss of Offsite Power/Loss of Forced Circulation Recovery Procedure The CRS has directed the auxiliary operator to secure TGB loads in accordance with Attachment 33 of OP-902-009, Standard Appendices.

With no turbine building battery chargers energized and no operator action, the maximum discharge time for the turbine building battery is ____(1)____ minutes. Securing TGB loads will preserve the remote operation of the ____(2)____ bus feeder breakers.

(1) (2) A. 30 2 B. 30 1 C. 90 1 D. 90 2 2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 32 of 150 Proposed Answer: D Explanation: (Optional)

A. Incorrect. Attachment 33-B of OP-902-009 states "With the turbine building battery chargers not energized and no operator action, the maximum discharge time for the Turbine Building battery is 90 minutes (not 30 minutes). 30 minutes is the time requirement to finish stripping safety related DC loads during a SBO. B. Incorrect. Attachment 33-B of OP-902-009 states "With the turbine building battery chargers not energized and no operator action, the maximum discharge time for the Turbine Building battery is 90 minutes (not 30 minutes). 30 minutes is the time requirement to finish stripping safety related DC loads during a SBO. Part 2 is correct. C. Incorrect. Part 1 is correct. DC control power for the 2 Bus feeder breakers is TGB-DC. The control power for the 1 Bus feeder breakers is from safety related A and B DC. D. CORRECT: Attachment 33-B of OP-902-009 states "With the turbine building battery chargers not energized and no operator action, the maximum discharge time for the Turbine Building battery is 90 minutes. DC control power for the 2 Bus feeder breakers is TGB-DC.

Technical Reference(s): TGOP-902-003 step 17 revision 305 (Attach if not previously provided) OP-902-009 attachment 33 (pg.169) revision 310 (including version/revision number) OP-902-003 step 17, revision 9 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 None Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Comments:

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 33 of 150 Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 000062 AK3.04 Importance Rating 3.5 K/A Statement AK3.04: Knowledge of the reasons for the following responses as they apply to the Loss of Nuclear Service Water: Effect on the nuclear service water discharge flow header of a loss of CCW.

Proposed Question: RO 17 Rev: 0 Given: Power is 100% A level switch malfunction has caused the B Dry Cooling Tower to isolate and bypass. The Control Room has entered OP-901-510, Component Cooling Water (CCW) System Malfunction A Loss of Coolant Accident occurs. Plant conditions are:

Pressurizer pressure is 1600 psia and slowly lowering Containment pressure is 17.3 psia and stable CCW heat exchanger B outlet temperature is 105 ºF If ACCW basin temperature is 80 ºF, ACC-126B, ACC Hdr B CCW HX Outlet Temp Control valve, will be ____(1)____ . This occurs to ______(2)______. (1) (2) A. open conserve ultimate heat sink water supply B. closed conserve ultimate heat sink water supply C. open ensure proper component cooling in the Wet mode D. closed ensure proper component cooling in the Wet mode 2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 34 of 150 Proposed Answer: B Explanation: (Optional)

A. Incorrect: ACC-126B will go closed (ACCW flow will stop) because its setpoint is automatically increased to 115 ºF on SIAS when ACCW basin temperature is >74 ºF. Normally, a setpoint of 95 ºF would cause the valve to open. Second part is correct. B. CORRECT: The loss of the Dry Cooling Tower (DCT) will cause CCW temp to rise steadily. SIAS and an ACCW basin temperature of >74 ºF will automatically increase ACC-126B's setpoint from a nominal 95ºF to 115ºF. Another function of ACC-126B is for the valve to remain closed unless the ACCW basin temperature is above 74 ºF this is to ensure proper cooling of Essential Chillers in Wet mode. Even though this interlock is met, ACC-126B will not open until the new setpoint of 115ºF is reached. TS 3.7.4 bases state that ACCW basin temperature and level limitations are based on a 30-day cooling water supply. C. Incorrect: First part: Plausible because initially, the loss of the DCT will cause ACC-126B to throttle open and maintain 95 ºF CCW temperature. On SIAS ACCW flow will stop due to a setpoint increase on ACC-126B. Second part: Plausible because a CCW temperature of 102 ºF will cause the Essential Chillers to swap from CCW to ACCW cooling (Wet Mode) to ensure proper cooling, but this is not the reason why ACC-126B closes. D. Incorrect: First part is correct. Second part is plausible because ACC-126B does not need to be open for ACCW to supply cooling water to the Essential Chiller.

Technical Reference(s): SD-CC pg 51 rev21 (Attach if not previously provided) TS 3.7.4 Bases (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 None Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Comments:

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 35 of 150 Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 000077 AK2.01 Importance Rating 3.1 K/A Statement AK2.01: Knowledge of the interrelations between Generator Voltage and Electric Grid Disturbances and the following: Motors. Proposed Question: RO 18 Rev: 0 Given: Power is 100% Main Generator load is 1200 MW; 200 MVAR Pine Bluff System Operations Center informed the control room that grid parameters are outside of the prescribed operating range The crew has entered OP-901-314, Degraded Grid Conditions and TRM 3.8.1.1 due to predicted post-trip offsite AC circuit voltage < 223KV The concern with a degraded grid (low voltage) is ______(1)______. To help stabilize the grid, the CRS orders raising main generator voltage. The BOP will ____(2)____ reactive load (MVARs).

(1) (2) A. degraded voltage relay reliability on the safety bus raise B. degraded voltage relay reliability on the safety bus lower C. emergency core cooling system motor overheating raise D. emergency core cooling system motor overheating lower 2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 36 of 150 Proposed Answer: C Explanation: (Optional)

A. Incorrect: The bases for TRM 3.8.1.1 discusses the safety bus degraded voltage relays and how their designed function is taken into account in the analysis but their functionality is not a concern for the degraded grid voltage scenario. Second part is correct. B. Incorrect: The bases for TRM 3.8.1.1 discusses the safety bus degraded voltage relays and how their designed function is taken into account in the analysis but their functionality is not a concern for the degraded grid voltage scenario. Lowering reactive load will lower main generator voltage and worsen the degraded grid condition.

C. CORRECT: The bases for TRM 3.8.1.1 mentions that the concern with starting and running motors at degraded voltage conditions is overheating. Raising Main Generator reactive load (MVARs) will help stabilize/increase grid voltage and improve the degraded condition. D. Incorrect: First part is correct. Lowering reactive load will lower main generator voltage and worsen the degraded grid condition.

Technical Reference(s): TRM 3.8.1.1 bases (Attach if not previously provided) OP-901-314 Sect. E0, rev 3 (including version/revision number) OP-010-004 limitation 3.2.27.2 rev 324 Proposed references to be provided to applicants during examination:

None Learning Objective: WLP-OPS-ED00 obj. 2 (As available)

WLP-OPS-TYC05 obj. 20 Question Source: Bank #

Modified Bank #

(Note changes or attach parent)

New X Question History: Last NRC Exam None Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 4 55.43 Comments:

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 37 of 150 Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 2 K/A # 000001 AA2.03 Importance Rating 4.5 K/A Statement AA2.03: Ability to determine and interpret the following as they apply to the Continuous Rod Withdrawal: Proper actions to be taken if automatic safety functions have not taken place.

Proposed Question: RO 19 Rev: 0 Given: Reactor Power is 10% with power ascension in progress Reg Group 6 is being withdrawn manually in Manual Group mode to raise power to 15% The ATC releases the IN-HOLD-OUT switch at 135" and Reg Group 6 CEA# 20 continues outward movement. The remaining Reg Group 6 CEAs indicate 135" The operator will FIRST ____(1)____ . If CEA 20 has stopped outward motion at 143" the crew will be required to ______(2)______ in accordance with the guidance in OP-901-102, CEA or CEDMCS Malfunction.

(1) (2) A. place CEDMCS mode select switch to OFF go to section E1, CEA misalignment greater than 7 inches B. remove the Group Select switch from the Reg Group 6 position go to section E1, CEA misalignment greater than 7 inches C. remove the Group Select switch from the Reg Group 6 position align CEAs to comply with TS 3.1.3.6, Regulating and Group P CEA Insertion limits D. place CEDMCS mode select switch to OFF align CEAs to comply with TS 3.1.3.6, Regulating and Group P CEA Insertion limits

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 38 of 150 Proposed Answer: A Explanation: (Optional)

A. CORRECT: OP-901-102 Section E3 requires the operator to place CEDMCS Mode Select switch to OFF if CEA continuous motion occurs. This same section directs the crew to go to section E1, CEA misalignment > 7 inches if any CEA is 7 inches misaligned. In this case, one CEA is 8 inches misaligned. B. Incorrect. The group select switch is taken into and out of the Reg Group 6 position when moving Reg Group 6 CEAs in Manual Group, but this is not the manipulation that OP-901-102 directs. Part 2 is correct. C. Incorrect. The group select switch is taken into and out of the Reg Group 6 position when moving Reg Group 6 CEAs in Manual Group, but this is not the manipulation that OP-901-102 directs. The step to align CEAs IAW TS 3.1.3.6 comes before the step for going to section E1. Section E3, Continuous Movement of CEA group was written for a continuous insertion or withdrawal of CEAs. Since this question is referencing a withdrawal, insertion limits of Reg Group 6 are not challenged. Also, TS 3.1.3.6 is for CEA groups, not individual CEAs. D. Incorrect. Part 1 is correct. Since this question is referencing a withdrawal, insertion limits of Reg Group 6 are not challenged.

Technical Reference(s): OP-901-102 Sect. E3, revision 302 (Attach if not previously provided) TS 3.1.3.6 (including version/revision number) OP-004-004 section 6.7 revision 19 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 None Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 2,6 55.43 Comments:

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 39 of 150 Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 2 K/A # 000005 AK3.05 Importance Rating 3.4 K/A Statement AK3.05: Knowledge of the reasons for the following responses as they apply to the Inoperable / Stuck Control Rod: Power limits on rod misalignment.

Proposed Question: RO 20 Rev: 0 Given: The plant is at 55% power performing a power ascension Group P is at 140 INCHES for ASI control At 0335 Group P Control Element Assembly (CEA) 35 slips to 125" Power is stabilized at 55% The crew implements OP-901-102, CEA or CEDMCS Malfunction Attempts to move the CEA are unsuccessful, CEA 35 is declared INOPERABLE At 0420 o Malfunctioning Automatic Control Timing Module (ACTM) card is replaced and o CEA 35 was realigned to 140" and CEA 35 is declared OPERABLE The earliest that the crew can re-commence the power ascension is _____(1)_____ to _______(2)______. (1) (2) A. 0535 minimize the effects of Xenon redistribution B. 0535 allow for clad relaxation C. 0620 minimize the effects of Xenon redistribution D. 0620 allow for clad relaxation

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 40 of 150 Proposed Answer: D Explanation: (Optional)

A. Incorrect. Per step16 of OP-901-102 Section E1, power must be maintained constant for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after realignment to allow for clad relaxation. The time in this selection is two hours from the initial event. Minimizing the effects of Xenon redistribution is part of the TS basis for allowing a one hour time limit for small misalignments. B. Incorrect. Per step16 of OP-901-102 Section E1, power must be maintained constant for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after realignment to allow for clad relaxation. The time in this selection is two hours from the initial event. The basis given in this selection is correct. C. Incorrect. Per step16 of OP-901-102 Section E1, power must be maintained constant for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after realignment to allow for clad relaxation. Minimizing the effects of Xenon redistribution is part of the TS basis for allowing a one hour time limit for small misalignments.

D. CORRECT: Per step16 of OP-901-102 Section E1, power must be maintained constant for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after realignment to allow for clad relaxation. This time would be 0620. Technical Reference(s): OP-901-102 Sect. E1 step 16 revision 302 (Attach if not previously provided) TS 3.1.3 Bases (including version/revision number)

Proposed references to be provided to applicants during examination:

None Learning Objective: WLP-OPS-PPO10 obj. 3 (As available)

Question Source: Bank #

X Question #20 Modified Bank #

(Note changes or attach parent)

New Question History: Last NRC Exam 2011 NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 2,6 55.43 Comments:

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 41 of 150 Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 2 K/A # 000032 AA1.01 Importance Rating 3.1 K/A Statement AA1.01: Ability to operate and / or monitor the following as they apply to the Loss of Source Range Nuclear Instrumentation: Manual restoration of power.

Proposed Question: RO 21 Rev: 0 Given: Reactor Power is 5.5 X 10

-8 ENI Log Channel C has failed hi Startup Channel high voltage is automatically removed when power exceeds approximately _____(1)_____%. To restore High Voltage power to Startup Channel _____(2)_____, the HV Control Switch in the Startup Channel drawer will be selected to Alternate.

(1) (2) A. 1.0 X 10-4 2 B. 1.0 X 10-4 1 C. 5.3 X 10-6 1 D. 5.3 X 10-6 2 2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 42 of 150 Proposed Answer: C Explanation: (Optional)

A. Incorrect. At 1X10

-4% power other operating bypasses operate such as placing DNBR and LPD trips in service; however, the startup channels are de-energized by the 5X10-6% bistables of the selected log channel. The power source to startup channel 1(not 2) will be log channel D when the HV control switch is taken to alternate and High Voltage power will be restored. B. Incorrect. At 1X10

-4% power other operating bypasses operate such as placing DNBR and LPD trips in service; however, the startup channels are de-energized by the 5X10-6% bistables of the selected log channel. Part 2 is correct C. CORRECT: The startup channels are de-energized by the 5X10

-6% bistables of the selected log channel. Log Channel C controls high voltage shutoff to Startup Channel 1 if the HV control Switch in Startup Channel 1 is selected to Primary. The power source to startup channel 1 will be log channel D when the HV control switch is taken to alternate and High Voltage power will be restored. D. Incorrect. Part 1 is correct. The power source to startup channel 1(not 2) will be log channel D when the HV control switch is taken to alternate and High Voltage power will be restored Technical Reference(s): OP-010-003 step 9.4.43 revision 335 (Attach if not previously provided) OP-500-008 L-3 revision 26 (for window) (including version/revision number)

Proposed references to be provided to applicants during examination:

None Learning Objective: WLP-OPS-ENI00 obj. 2 and 3 (As available)

Question Source: Bank #

Question #22 Modified Bank #

X (Note changes or attach parent)

New Question History: Last NRC Exam 2011 NRC RO Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 2,7 55.43 Comments:

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 43 of 150 Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 2 K/A # 000036 (BW/A08) 2.4.11 Importance Rating 4.0 K/A Statement 2.4.11 Knowledge of abnormal condition procedures Proposed Question: RO 22 Rev: 0 Given: The plant is in MODE 6 with fuel shuffle in progress in both the Fuel Handling Building and Containment Containment Purge is in the Refueling Mode Containment equipment hatch and airlock doors are closed and no containment penetrations are impaired A fuel bundle is dropped from the Refueling Machine Fuel Hoist The crew has entered OP-901-405, Fuel Handling Incident Which of the following radiation monitors will automatically terminate the radioactive gas release to the environment?

A. Containment Atmosphere Hi Range Area Radiation Monitor, ARM-IRE-5400AS B. Containment Purge Area Radiation Monitor, ARM-IRE-5024 C. Refueling Machine Area Radiation Monitor, ARM-IRE-5013 D. Containment PIG Process Radiation Monitor, PRM-IRE-0100S

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 44 of 150 Proposed Answer: B Explanation: (Optional)

A. Incorrect. No automatic functions are associated with this Rad Monitor. However, the RM detectors are in containment and could see the radiation from the fuel handling incident. B. CORRECT: ARM 5024 provides isolation of Containment Purge on Hi Rad and is in the general area of fuel handling on the +46' elevation of Containment. The rad monitors and their isolation functions are located in section C of OP-901-405, Fuel Handling Incident. C. Incorrect. This rad monitor is in the general area of the Refueling Cavity; however, it has indication and alarm functions only. D. Incorrect. The containment PIG monitors containment atmosphere but does not have Containment Purge isolation functions.

Technical Reference(s): OP-901-405 pg.6, revision 7 (Attach if not previously provided) (including version/revision number)

Proposed references to be provided to applicants during examination:

None Learning Objective: WLP-OPS-RMS00 obj. 2 (As available)

Question Source: Bank #

X Question #58 Modified Bank #

(Note changes or attach parent)

New Question History: Last NRC Exam 2010 NRC RO Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 11 55.43 Comments:

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 45 of 150 Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 2 K/A # 000037 AK1.01 Importance Rating 2.9 K/A Statement AK1.01 Knowledge of the operational implications of the following concepts as they apply to Steam Generator Tube Leak: Use of Steam Tables Proposed Question: RO 23 Rev: 0 Given: A cooldown is being performed to MODE 5 per OP-901-202, Steam Generator Tube Leakage or High Activity due to a leak on SG 2 All RCPs are secured CET Temp is 382 ºF T HOT Loop1 is 375 ºF T HOT Loop2 is 378 ºF TCOLD Loop 1 is 350 ºF TCOLD Loop 2 is 382 ºF The CRS has placed a lower limit of 30ºF on Subcool Margin Determine the minimum value of RCS pressure that supports the requested RCS Subcool Margin.

A. 196 psia B. 262 psia C. 270 psia D. 283 psia

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 46 of 150 Proposed Answer: D Explanation: (Optional)

A. Incorrect. This is the value associated with Cold leg 1 temperature. CET temperature should be used to determine subcool margin when on natural circulation. Using Cold leg 1 would also lower RCS pressure below the isolated S/G pressure. B. Incorrect. This is the value associated with Hot Leg Loop 1 temperature. CET temperature should be used to determine subcool margin when on natural circulation. C. Incorrect. This is the value associated with Hot Leg Loop 2 temperature. CET temperature should be used to determine subcool margin when on natural circulation.

D. CORRECT: This is the value associated with CET and Cold Leg 2 temperatures. The hottest RCS temperature should be used to determine subcool margin. OI-038-000 has guidance to use CET temperatures on natural circulation.

Technical Reference(s): OI-038-000 rev 10 (Attach if not previously provided) Steam Tables (including version/revision number)

Proposed references to be provided to applicants during examination: Steam Tables Learning Objective: WLP-OPS-PPO20 Obj. 3 (As available)

Question Source: Bank #

X Question # 59 Modified Bank #

(Note changes or attach parent)

New Question History: Last NRC Exam 2010 RO NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10,14 55.43 Comments:

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 47 of 150 Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 2 K/A # 000061 AK1.01 Importance Rating 2.5 K/A Statement AK1.01: Knowledge of the operational implications of the following concepts as they apply to Area Radiation Monitoring (ARM) System Alarms: Detector limitations.

Proposed Question: RO 24 Rev: 0 During a LOCA or Steam Line Break in Containment, the _____(1)______ Radiation Monitor(s) will read erroneously _____(2)______ while Containment temperature is rising. (1) (2) A. Containment PIG low B. Containment PIG high C. Containment High Range low D. Containment High Range high 2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 48 of 150 Proposed Answer: D Explanation: (Optional)

A. Incorrect: The Containment High Range Radiation Monitors have been determined to be susceptible to TIC post accident which will cause erroneously high readings for at least 15 minutes from the time containment temperature stabilizes. B. Incorrect: The Containment High Range Radiation Monitors have been determined to be susceptible to TIC post accident which will cause erroneously high readings for at least 15 minutes from the time containment temperature stabilizes. C. Incorrect: The correct Rad monitor but the wrong effect.

D. CORRECT: The Containment High Range Radiation Monitors have been determined to be susceptible to TIC post accident which will cause erroneously high readings for at least 15 minutes from the time containment temperature stabilizes.

Technical Reference(s): OI-038-000 step 5.5.1 Revision 10 (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. 4 (As available)

WLP-OPS-MCD06 obj. 3 Question Source: Bank #

X Question #23 Modified Bank #

(Note changes or attach parent)

New Question History: Last NRC Exam 2014 RO Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10,11 55.43 Comments:

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 49 of 150 Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 2 K/A # 000068 (BW/A06) AA2.08 Importance Rating 3.9 K/A Statement AA2.08: Ability to determine and interpret the following as they apply to the Control Room Evacuation: S/G pressure.

Proposed Question: RO 25 Rev: 0 Given: Power is 100% A special test of the Control Room Envelope has inadvertently rendered continued operation from the Control Room impossible The CRS enters OP-901-502, Evacuation of Control Room & Subsequent Plant Shutdown Plant cooldown will be controlled using the ____(1)____. Per OP-901-502 Attachment 14, Reliable Instrumentation, reliable indication of Steam Generator pressure can be read on ____(2)____. (1) (2) A. atmospheric dump valves both trains B. steam bypass control system both trains C. atmospheric dump valves train B only D. steam bypass control system train B only

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 50 of 150 Proposed Answer: A Explanation: (Optional)

A. CORRECT: Initially MSIVs remain open through the performance of the immediate operator actions (due to no fire) however, OP-901-502, subsection E2, step 25 will direct operators to close MSIVs by initiating MSIS locally at the ESFAS cabinets and step 34 directs plant cooldown using ADVs. SG pressure has both trains A and B available as reliable instrumentation per attachment 14. B. Incorrect. Plausible because MSIVs are not closed during the immediate operator actions. Also, off-site power (i.e. condenser cooling) remains available throughout the entire procedure. Second part is correct. C. Incorrect. First part correct. Plausible because the list of reliable instrumentation (per att. 14) is not the same for both trains. For example, reliable indication of CCW temperature, saturation margin and shutdown flow are found only on train B. D. Incorrect. Plausible because MSIVs are not closed during the immediate operator actions. Also, off-site power (i.e. condenser cooling) remains available throughout the entire procedure. Plausible because the list of reliable instrumentation (per att. 14) is not the same for both trains. For example, reliable indication of CCW temperature, saturation margin and shutdown flow are found only on train B.

Technical Reference(s): OP-901-502 pp. 47,54,123 rev29 (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. 17 (As available)

Question Source: Bank #

Modified Bank #

(Note changes or attach parent)

New X Question History: Last NRC Exam None Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 4,7 55.43 Comments:

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 51 of 150 Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 2 K/A # 000069 (W/E14)AAK2.03 Importance Rating 2.8 K/A Statement AAK2.03: Knowledge of the interrelations between the Loss of Containment Integrity and the following: Personnel access hatch and emergency access hatch.

Proposed Question: RO 26 Rev: 0 Which ONE of the following conditions will result in a Loss of Containment Integrity per Technical Specification while operating in MODE 1?

A. One of two normally open redundant containment isolation valves has failed in the CLOSE position with power removed.

B. A mechanic props OPEN the Personnel Airlock outer door to perform corrective maintenance on an inoperable inner door.

C. A blank flange was installed in place of a containment isolation valve which had failed its surveillance test.

D. A manual isolation valve is CLOSED to isolate a penetration where an electrician had disconnected the auto-close feature for a containment isolation valve.

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 52 of 150 Proposed Answer: B Explanation: (Optional)

A. Incorrect: This statement meets the Tech Spec requirement per TS 3.6.1.1. B. CORRECT: Per TS 3.6.1.3, One of Two Airlock Doors must be operable and closed at all times. C. Incorrect: This statement meets the Tech Spec requirement per TS 3.6.1.1. D. Incorrect: This statement meets the Tech Spec requirement per TS 3.6.1.1 Technical Reference(s): T.S. 3.6.1.3 (Attach if not previously provided) TS 3.6.1.1 (including version/revision number) TS definition of Containment Integrity Proposed references to be provided to applicants during examination:

None Learning Objective: WLP-OPS-TS04 Obj: 2 (As available)

Question Source: Bank #

X Question 26 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 9 55.43 Comments:

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 53 of 150 Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 2 K/A # CE/A16 AK3.3 Importance Rating 3.3 K/A Statement AK3.3: Knowledge of the reasons for the following responses as they apply to the (Excess RCS Leakage): Manipulation of controls required to obtain desired operating results during abnormal, and emergency situations.

Proposed Question: RO 27 Rev: 0 Given: Plant is at 100% power The crew has entered OP-901-111, Reactor Coolant System Leak RCS Tavg is 575 degrees and steady Charging Pump A is running Pressurizer level is 55% and slowly lowering The Standby Charging Pumps selector switch is selected to the B-AB position The CRS will direct the ATC to place Charging Pump B C/S to the ON position ____(1)____. The ATC will determine an RCS leak rate by subtracting the total of ____(2)____. (1) (2) A. to allow for a more accurate leakrate determination Charging flow and RCP CBO flow from Letdown flow B. because Charging Pump B should have auto started Charging flow and RCP CBO flow from Letdown flow C. because Charging Pump B should have auto started Letdown flow and RCP CBO flow from Charging flow D. to allow for a more accurate leakrate determination Letdown flow and RCP CBO flow from Charging flow

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 54 of 150 Proposed Answer: D Explanation: (Optional)

A. Incorrect. Part 1 is correct. A leakrate will be determined by subtracting the total of letdown flow and RCP CBO flow from charging flow. B. Incorrect. The setpoint for an auto start of the backup charging pump has not been reached. The auto start is -2.5% from setpoint. Setpoint at 100% power is 56%. Therefore the auto start would occur at 53.5%. A leakrate will be determined by subtracting the total of letdown flow and RCP CBO flow from charging flow. C. Incorrect. The setpoint for an auto start of the backup charging pump has not been reached. The auto start is -2.5% from setpoint. Setpoint at 100% power is 56%. Therefore the auto start would occur at 53.5%. Part 2 is correct.

D. CORRECT: Note in OP-901-111 states if RCS leakage will result in a backup charging Pump cycling to maintain pressurizer level, starting and continuously running (place to on) an additional Charging Pump will allow for a more accurate leakrate determination. A leakrate will be determined by subtracting the total of letdown flow and RCP CBO flow from charging flow.

Technical Reference(s): OP-901-111 pp. 5,8 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-PPO10 obj. 3 (As available)

Question Source: Bank #

Modified Bank #

(Note changes or attach parent)

New X Question History: Last NRC Exam None Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Comments:

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 55 of 150 Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 003 A1.09 Importance Rating 2.8 K/A Statement A1.09: Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the RCPS controls including:

Seal flow and D/P.

Proposed Question: RO 28 Rev: 0 Given: The reactor is operating at 100% power At 0100 RCP 1A Lower Seal fails and Controlled Bleedoff flow rises to 2.5 GPM OP-901-130, Reactor Coolant Pump Malfunction is implemented and the plant remains at 100% power At 0530 RCP 1A Middle Seal fails RCP 1A Controlled Bleedoff flow went out of range high and then went to 0 GPM RCP 1A Controlled Bleedoff flow went low due to closure of ________(1)_________. The crew should ________(2)_________ and secure RCP 1A.

(1) (2) A. a RCP Controlled Bleedoff Containment Isolation Valve perform a shutdown to MODE 3 per OP-010-005, Plant Shutdown B. a RCP Controlled Bleedoff Containment Isolation Valve manually trip the reactor and perform Standard Post Trip Actions C. the check valve on RCP 1A Controlled Bleedoff line perform a shutdown to MODE 3 per OP-010-005, Plant Shutdown D. the check valve on RCP 1A Controlled Bleedoff line manually trip the reactor and perform Standard Post Trip Actions

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 56 of 150 Proposed Answer: D Explanation: (Optional)

A. Incorrect. Seal flow would not go to zero on the reactor coolant pump with normal RCS pressure. Instead the flow would be diverted to the Quench Tank via a relief valve on the common controlled bleedoff line (RC-603). A reactor trip is required vice a normal shutdown due the rapid failure of two RCP seals. B. Incorrect. Seal flow would not go to zero on the reactor coolant pump with normal RCS pressure. Instead the flow would be diverted to the Quench Tank via a relief valve on the common controlled bleedoff line (RC-603). C. Incorrect. A reactor trip is required vice a normal shutdown due the rapid failure of two RCP seals.

D. CORRECT: The RCP 1A Controlled Bleedoff excess flow check valve will seat at 10 GPM flow and isolate controlled bleedoff flow from the RCP. Two seals failing within a 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> period requires a reactor trip.

Technical Reference(s): OP-901-130 rev. 9 (Attach if not previously provided) SD-CVC pg. 39 rev. 16 (including version/revision number)

Proposed references to be provided to applicants during examination:

None Learning Objective: WLP-OPS-PPO10 obj. 3 (As available)

Question Source: Bank #

X Question #29 Modified Bank #

(Note changes or attach parent)

New Question History: Last NRC Exam 2011 NRC RO Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 3,10 55.43 Comments:

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 57 of 150 Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 003 A3.05 Importance Rating 2.7 K/A Statement A3.05: Ability to monitor automatic operation of the RCPS, including: RCP lube oil and bearing lift pumps.

Proposed Question: RO 29 Rev: 0 Given: The reactor was at 100% power RCP 1A trips on overcurrent

__(1)__ RCP 1A Lift Oil Pump(s) started when RCP 1A __(2)__. (1) (2) A. One load breaker opened B. Both load breaker opened C. One speed lowered to 600 RPM D. Both speed lowered to 600 RPM

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 58 of 150 Proposed Answer: B Explanation: (Optional)

A. Incorrect. Both lift oil pumps will start when the RCP breaker opens. Candidate must know the normal alignment of the lift oil pump control switches to eliminate this selection. B. CORRECT: At 100% power the lift oil pump control switches will be in Auto. The RCP breaker opening causes both lift oil pumps to start immediately. C. Incorrect. Wrong number of pumps. The initiating signal is incorrect. However, this is the original Waterford design which makes it plausible. D Incorrect. Correct number of pumps start. The initiating signal is incorrect.

Technical Reference(s): OP-001-002 pg.14, Revision 22 (Attach if not previously provided) WLP-OPS-RCP00, slide 96 Revision 24 (including version/revision number)

Proposed references to be provided to applicants during examination:

None Learning Objective: WLP-OPS-RCP00 obj. 5 (As available)

Question Source: Bank #

X Question #28 Modified Bank #

(Note changes or attach parent)

New Question History: Last NRC Exam 2011 NRC RO Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 5,7 55.43 Comments:

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 59 of 150 Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 004 K1.02 Importance Rating 3.5 K/A Statement K1.02: Knowledge of the physical connections and/or cause-effect relationships between the CVCS and the following systems: PZR and RCS temperature and pressure relationships.

Proposed Question: RO 30 Rev: 0 Given: Power is 100% A grid transient causes turbine load to lower by 10%

As a result of this transient, RCS temperature will ____(1)____ and the Letdown Backpressure control valve will ______(2)______. (1) (2) A. lower open B. lower close C. rise open D. rise close 2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 60 of 150 Proposed Answer: C Explanation: (Optional)

A. Incorrect: Plausible because pressurizer water temperature will lower due to an in-surge into the pressurizer due to the RCS temperature rise. Second part is correct. B. Incorrect: Plausible because pressurizer water temperature will lower due to an in-surge into the pressurizer due to the RCS temperature rise. The backpressure control valve will open in response to the higher pressure in the Letdown heat exchanger due to the Letdown flow control valve opening.

C. CORRECT: The loss of 10% turbine load will cause RCS temperature to rise. The higher pressurizer level will cause the Letdown flow control valve to open and in response, the backpressure control valve will also open. D. Incorrect: First part is correct. The backpressure control valve will open in response to the higher pressure in the Letdown heat exchanger due to the Letdown flow control valve opening.

Technical Reference(s): SD-PLC pg 9, rev 9 (Attach if not previously provided) SD-CVC pg 14, rev 16 (including version/revision number)

Proposed references to be provided to applicants during examination:

None Learning Objective: WLP-OPS-PLC00 obj. 2,3 (As available)

Question Source: Bank #

Modified Bank #

(Note changes or attach parent)

New X Question History: Last NRC Exam None Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 5 55.43 Comments:

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 61 of 150 Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 005 A1.03 Importance Rating 2.5 K/A Statement A1.03: Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the RHRS controls including:

Closed cooling water flow rate and temperature.

Proposed Question: RO 31 Rev: 0 Given: The plant is in Mode 4 Shutdown Cooling Train A is in service RCS temperature is 204°F and steady Component Cooling Water temperature is dropping through the night as ambient temperature drops To prevent entry into Mode 5, the SDC Train A Temperature Control Valve (SI-415A) must be throttled ____(1)____. This will result in the SDC Train A Flow Control Valve (SI-129A) automatically throttling ______(2)______. (1) (2) A. open open B. open closed C. closed closed D. closed open 2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 62 of 150 Proposed Answer: D Explanation: (Optional)

A. Incorrect. To prevent entry into Mode 5 (RCS temp <200°F), less SDC flow is required through the SDC HX to compensate for CCW temperature dropping. This is done by closing the TCV (SI-415A). Part 2 is correct B. Incorrect. To prevent entry into Mode 5 (RCS temp <200°F), less SDC flow is required through the SDC HX to compensate for CCW temperature dropping. This is done by closing the TCV (SI-415A). To maintain flow constant, the FCV (SI-129A) will automatically open since there is now less flow going through the HX. C. Incorrect. Part 1 is correct. SI-129A will open to maintain flow constant since there is now less flow going through the SDC HX.

D. CORRECT: To prevent entry into Mode 5 (RCS temp <200°F), less SDC flow is required through the SDC HX to compensate for CCW temperature dropping. This is done by closing the TCV (SI-415A). To maintain flow constant, the FCV (SI-129A) will automatically open since there is now less flow going through the HX.

Technical Reference(s): SD-SDC pp. 5,6,13 revision 7 (Attach if not previously provided) (including version/revision number)

Proposed references to be provided to applicants during examination:

None Learning Objective: WLP-OPS-SDC00 obj 4 (As available)

Question Source: Bank #

Modified Bank #

(Note changes or attach parent)

New X Question History: Last NRC Exam None Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Comments:

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 63 of 150 Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 006 K2.04 Importance Rating 3.6 K/A Statement K2.04: Knowledge of bus power supplies to the following: ESFAS-operated valves.

Proposed Question: RO 32 Rev: 0 What is the power supply to SI-138 A, LPSI Header to RC Loop 2A Control Isolation?

A. Bus 213 A B. Bus 311 A C. Bus 314 A D. Bus 315 A 2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 64 of 150 Proposed Answer: B Explanation: (Optional)

A. Incorrect. SI-138A is not powered from the 213A bus. The 213A bus is non-safety related but is located in the same switchgear room as the 311A bus. B. CORRECT: SI-138A is powered from the 311A bus. The 311A bus is a safety related 480V MCC. C. Incorrect. SI-138A is not powered from the 314A bus. The 315A bus is a safety related 480V MCC. D. Incorrect. SI-138A is not powered from the 315A bus. The 315A bus is a safety related 480V MCC.

Technical Reference(s): OP-009-008 Attachment 11.3 revision 37 (Attach if not previously provided) (including version/revision number)

Proposed references to be provided to applicants during examination:

None Learning Objective: WLP-OPS-SI00 obj. 3 (As available)

Question Source: Bank #

X Question #5 Modified Bank #

(Note changes or attach parent)

New Question History: Last NRC Exam 2010 RO NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 8 55.43 Comments:

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 65 of 150 Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 006 K6.03 Importance Rating 3.6 K/A Statement K6.03: Knowledge of the effect of a loss or malfunction on the following will have on the ECCS: Safety Injection Pumps.

Proposed Question: RO 33 Rev: 0 Given: A Loss Of Cooling Accident is in progress Reactor Coolant System (RCS) pressure is 200 PSIA High Pressure Safety Injection Pump B trips shortly after starting All other equipment functions as designed The minimum required Safety Injection flow for High Pressure Safety Injection is ___(1)____ and ___(2)____ is required for Low Pressure Safety Injection.

(1) (2) A. >185 GPM total flow to the cold legs no flow B. >185 GPM total flow to the cold legs

>185 GPM flow to loop 2 cold legs C. >185 GPM to each cold leg no flow D. >185 GPM to each cold leg >185 GPM flow to loop 2 cold legs

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 66 of 150 Proposed Answer: C Explanation: (Optional)

A. Incorrect HPSI >185 GPM to each cold leg not total flow to all legs is required .Part 2 is correct. B. Incorrect. HPSI flow should be to each cold leg not total flow to all legs and LPSI flow would be 0 GPM at 200 PSIA based on the flow curves.

C. CORRECT: Flow should be >185 GPM on each cold leg and 0 flow on the LPSI based on the flow curves. D. Incorrect. Part 1 is correct. LPSI flow required is 0 GPM based on the flow curves.

Technical Reference(s): OP-902-009 Attachment 2-E and 2-F, rev 310 (Attach if not previously provided) (including version/revision number)

Proposed references to be provided to applicants during examination: OP-902-009 Attachment 2-E and 2-F Learning Objective: WLP-OPS-PPE02 obj. 17 (As available)

Question Source: Bank #

X Question #33 Modified Bank #

(Note changes or attach parent)

New Question History: Last NRC Exam 2011 NRC RO Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7,8 55.43 Comments:

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 67 of 150 Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 007 K3.01 Importance Rating 3.3 K/A Statement K3.01: Knowledge of the effect that a loss or malfunction of the PRTS will have on the following: Containment.

Proposed Question: RO 34 Rev: 0 A Steam Generator Tube Rupture has occurred that resulted in an automatic SIAS/CIAS.

Which ONE of the following could result in a Quench Tank Rupture Disc failure and rising containment pressure, due to automatic alignment to the Quench Tank?

A. RCP Control Bleedoff B. RCP Vapor Seal Leak Off C. Reactor Head Vent Header D. Pressurizer Vent Header

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 68 of 150 Proposed Answer: A Explanation: (Optional)

A. CORRECT: RC-606, RCP Control Bleedoff Inside Containment Isolation Valve closes on a CIAS, redirecting RCP control bleedoff to the quench tank through a relief valve (RC-603). If no action is taken, this flow could eventually lead to a Quench Tank rupture disc failure

. B. Incorrect. RCP Vapor seal leakage is directed to the containment sump at all times. C. Incorrect. Reactor Vent header must be aligned to the quench tank manually. D. Incorrect. Pressurizer vent header must be aligned to the quench tank manually.

Technical Reference(s): OP-001-002, Att. 11.3 rev 22 (Attach if not previously provided) SD-CVC pp. 39-40 rev 16 (including version/revision number)

Proposed references to be provided to applicants during examination:

None Learning Objective: WLP-OPS-RCP00 obj. 3, 5 (As available)

Question Source: Bank #

X Question #33 Modified Bank #

(Note changes or attach parent)

New Question History: Last NRC Exam 2014 NRC RO Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Comments:

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 69 of 150 Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 008 A1.01 Importance Rating 2.8 K/A Statement A1.01: Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the CCWS controls including:

CCW flow rate.

Proposed Question: RO 35 Rev: 0 CCW Pump A has tripped with CCW Pump AB OOS.

Which of the following actions is performed to protect CCW Pump B from runout conditions?

A. Split out the A and B CCW headers B. Close the NNS loop isolations C. Secure Train B Containment Fan Coolers D. Align Chiller B Cooling to the Wet Tower

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 70 of 150 Proposed Answer: A Explanation: (Optional)

A. CORRECT: OP-901-510 section E2 step 8 directs the crew to split CCW headers if only one Component Cooling Water Pump is operating. B. Incorrect. Closing NNS loop isolations would not change flow enough. C. Incorrect. A header is more flow, stopping fan coolers would lower flow, but not minimize runout concern. D. Incorrect. Chiller B would be aligned to dry tower for this event.

Technical Reference(s): OP-901-510 section E2 step 8 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-PPO50 obj. 3 (As available)

Question Source: Bank #

X Question #8 Modified Bank #

(Note changes or attach parent)

New Question History: Last NRC Exam 2007 NRC RO Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Comments:

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 71 of 150 Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 010 K6.03 Importance Rating 3.2 K/A Statement K6.03: Knowledge of the effect of a loss or malfunction of the following will have on the PZR PCS: PZR sprays and heaters.

Proposed Question: RO 36 Rev: 0 Given: Plant is at 100% power Pressure Control Channel Select switch is in 'X' Low Level Heater Cutout switch is in 'BOTH' Spray Valve Selector switch is in 'Loop 1A' Pressurizer pressure control channel 'X' fails HIGH.

___(1)____ spray valve(s) will open. Pressurizer back-up heaters ___(2)____ energize on lowering pressure.

(1) (2) A. One will B. One will not C. Both will D. Both will not 2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 72 of 150 Proposed Answer: B Explanation: (Optional)

A. Incorrect: Part 1 is correct. Since the selected pressurizer pressure control channel is failed high, all pressurizer heaters will be cutout and not energize regardless of actual pressure. B. CORRECT: With the selected pressure control channel (X) failed high, both spray valves would normally receive an open signal. However, with the spray valve selector switch in '1A', only the 1A spray valve will receive an open signal. Since the selected pressurizer pressure control channel is failed high (>2270 psia), all pressurizer heaters will be cutout and not energize regardless of actual pressure. C. Incorrect. With the spray valve selector switch in '1A', only the 1A spray valve will receive an open signal. With the selected pressurizer pressure control channel failed high, all pressurizer heaters are cutout. D. Incorrect. With the spray valve selector switch in '1A', only the 1A spray valve will receive an open signal. Part 2 is correct.

Technical Reference(s): SD-PLC pp. 14,29,Fig.6 (rev 9) (Attach if not previously provided) (including version/revision number)

Proposed references to be provided to applicants during examination:

None Learning Objective: WLP-OPS-PLC00 obj. 6 (As available)

Question Source: Bank #

Modified Bank #

(Note changes or attach parent)

New X Question History: Last NRC Exam None Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Comments:

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 73 of 150 Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 012 K2.01 Importance Rating 3.3 K/A Statement K2.01: Knowledge of bus power supplies to the following: RPS channels, components, and interconnections.

Proposed Question: RO 37 Rev: 0 Given: The plant is at 100% power A loss of SUPS MD occurs Reactor Trip Switchgear breakers _____(1)_____ open and a reactor trip _______(2)______. (1) (2) A. 3, 4, 7, and 8 occurs B. 3, 4, 7, and 8 does not occur C. 1, 2, 5, and 6 occurs D. 1, 2, 5, and 6 does not occur

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 74 of 150 Proposed Answer: B Explanation: (Optional)

A. Incorrect. Correct breakers but reactor trip does not occur, as power to CEDMCS is not interrupted with these reactor trip breakers open B. CORRECT: On al loss of SUPS MD, Reactor trip breakers 3, 4, 7, and 8 open but power is still available to CEDMCS. C. Incorrect. Reactor trip breakers 1, 2, 5, and 6 are A train powered and are not affected. A reactor trip would still not occur. D. Incorrect. Reactor trip breakers 1, 2, 5, and 6 are A train powered and are not affected. A reactor trip would still not occur.

Technical Reference(s): OP-901-312 pg. 18 revision 307 (Attach if not previously provided) SD-CED Fig. 8 revision 11 (including version/revision number)

Proposed references to be provided to applicants during examination:

None Learning Objective: WLP-OPS-PPO50 obj. 3 (As available)

Question Source: Bank #

X Question #10 Modified Bank #

(Note changes or attach parent)

New Question History: Last NRC Exam 2007 RO makeup Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 6 55.43 Comments:

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 75 of 150 Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 012 2.4.45 Importance Rating 4.5 K/A Statement 2.4.45: Ability to prioritize and interpret the significance of each annunciator or alarm. Proposed Question: RO 38 Rev: 0 Given: Plant is at 90% power Reactor Cutback is in service CRS entered OP-901-201, Steam Generator (SG) level control malfunction due to a problem with SG level control. The following alarms are present in the Control Room:

SG1 STEAM/FW FLOW SIGNAL DEV SG2 STEAM/FW FLOW SIGNAL DEV SG1 LEVEL HI/LO SG2 LEVEL HI/LO FWPT A FLOW LO SG1 LEVEL LO PRETRIP A/C SG1 LEVEL LO PRETRIP B/D SG1 level is 26% NR on all indications and stable. SG2 level is 30% NR on all indications and stable. The operator reports FWPT A speed is 3900 rpm and FWPT B speed is 5200 rpm.

The crew will prioritize the ____(1)____ alarm(s) and take action to _____(2)_____. (1) (2) A. SG1 LEVEL LO PRETRIP manually trip FWPT A B. FWPT A FLOW LO manually trip FWPT A C. SG1 LEVEL LO PRETRIP manually trip the reactor D. FWPT A FLOW LO manually trip the reactor

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 76 of 150 Proposed Answer: C Explanation: (Optional)

A. Incorrect: First part is correct. OP-901-201, SG Level Control Malfunction, directs tripping the FWPT with a speed controller malfunction (in this case 3900 rpm indicates a speed controller malfunction). This will initiate a reactor cutback and allow the remaining FWPT to restore SG levels. However, in this case the low SG level RPS setpoint has been exceeded and the reactor must be manually tripped. B. Incorrect: The pre-trip alarms should be prioritized because these alarms signify SG level is close to a reactor trip setpoint. OP-901-201, SG Level Control Malfunction, directs tripping the FWPT with a speed controller malfunction. This will initiate a reactor cutback and allow the remaining FWPT to restore SG levels. However, in this case the low SG level RPS setpoint has been exceeded and the reactor must be manually tripped.

C. CORRECT: The pre-trip alarms should be prioritized because these alarms signify SG level is close to a reactor trip setpoint. SG level of 26% NR means the reactor should have automatically tripped (ATWS). The correct action is to manually trip the reactor. D. Incorrect: FWPT low flow alarm is an important alarm and it may prompt action to trip the reactor, however, unlike the SG Lo level pre-trip alarms, it is not a direct indication of approaching an RPS trip setpoint. Second part is correct.

Technical Reference(s): OP-901-201 pg. 3,5,10 rev6 (Attach if not previously provided) EN-OP-115 pg. 8 rev15 (including version/revision number)

Proposed references to be provided to applicants during examination:

None Learning Objective: WLP-OPS-PPO20 obj.1 (As available)

Question Source: Bank #

Modified Bank #

(Note changes or attach parent)

New X Question History: Last NRC Exam None Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 4,7 55.43 Comments:

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 77 of 150 Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 013 K5.01 Importance Rating 2.8 K/A Statement K5.01: Knowledge of the operational implications of the following concepts as they apply to the ESFAS: Definitions of safety train and ESF channel.

Proposed Question: RO 39 Rev: 0 Manual actuation of a Containment Spray Actuation Signal (CSAS) from the Control Room requires _____(1)_____ two of the four ESFAS channels and will start _____(2)_____ train(s) of Containment Spray.

(1) (2) A. a specific one B. a specific both C. any one D. any both 2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 78 of 150 Proposed Answer: B Explanation: (Optional)

A. Incorrect: First part is correct. Meeting the trip logic will start both trains of Containment Spray. B. CORRECT: ESFAS logic is designed with a selective 2 out of 4 trip strategy. The correct 2 out of 4 channels must be actuated to meet the trip logic. Once the trip logic is met, both trains of Containment Spray will start. C. Incorrect: The correct 2 out of 4 channels must be actuated to meet the trip logic not just any two. Once the trip logic is met, both trains of Containment Spray will start. D. Incorrect: The correct 2 out of 4 channels must be actuated to meet the trip logic not just any two. Second part is correct.

Technical Reference(s): SD-PPS pg. 43 & Fig. 40 (rev16) (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. 5,6 (As available)

Question Source: Bank #

Modified Bank #

(Note changes or attach parent)

New X Question History: Last NRC Exam None Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 Comments:

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 79 of 150 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 40 Rev: 0 Given: Plant has experienced an Excess Steam Demand event RCS Pressure is 1750 PSIA Containment Pressure is 16.9 PSIA Steam Generator 1 pressure is 650 PSIA Steam Generator 2 pressure is 600 PSIA Steam Generator #1 level is 5% NR and dropping Steam Generator #2 level is 4% NR and dropping What is the current status of EFAS 1 and EFAS 2?

A. Only EFAS 1 is initiated B. Only EFAS 2 is initiated C. Both EFAS 1 and EFAS 2 are initiated D. Neither EFAS 1 nor EFAS 2 are initiated

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 80 of 150 Proposed Answer: D Explanation: (Optional)

A. Incorrect. An EFAS-1 is not present because SG pressure has dropped below 666 PSIA and the required DP of 123 PSID has not been established. B. Incorrect. An EFAS-2 is not present because SG pressure has dropped below 666 PSIA and the required DP of 123 PSID has not been established. C. Incorrect. An EFAS-1 or EFAS-2 is not present because SG pressure has dropped below 666 PSIA and the required DP of 123 PSID has not been established.

D. CORRECT: An EFAS is generator for the respective generator if one of the following conditions occur: 1). Low SG level (27.4%NR) coincident with no low SG pressure (666 PSIA) or 2) Low SG level (27.4%NR) coincident with a generator DP of 123 PSID to feed the SG with the higher pressure. In this case, both SG levels have reached the EFAS setpoint but the 123 PSID setpoint has not been achieved in either SG. Therefore, no EFAS-1 nor EFAS-2 has been generated.

Technical Reference(s): SD-EFW page 35 rev 12 (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 None Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Comments:

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 81 of 150 Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 022 K4.02 Importance Rating 3.1 K/A Statement K4.02: Knowledge of CCS design feature(s) and/or interlock(s) which provide for the following: Correlation of fan speed and flowpath changes with containment pressure. Proposed Question: RO 41 Rev: 0 Given: The plant is operating at 100% power A LOCA occurs Containment pressure is 17.3 PSIA and rising Pressurizer pressure is 1700 PSIA and lowering The containment fan coolers will be operating in _____(1)_____ speed. Containment Fan Cooler discharge will be to the containment cooling ______(2)_____. (1) (2) A. slow system ring header B. slow safety discharge dampers C. fast safety discharge dampers D. fast system ring header

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 82 of 150 Proposed Answer: B Explanation: (Optional)

A. Incorrect. Part 1 is correct. The discharge of the containment fan coolers will swap from the CCS ring header to CCS-102A and CCS-102B. These dampers fail open on a SIAS. B. CORRECT: SIAS is present. PZR pressure does not meet the setpoint for a SIAS but containment pressure does. On a SIAS, all containment fan coolers will swap to slow speed. The discharge of the containment fan coolers will swap from the CCS ring header to CCS-102A and CCS-102B. These dampers fail open on a SIAS. C. Incorrect. On a SIAS, all containment fan coolers will swap to slow speed. Part 2 is correct. D. Incorrect. SIAS is present. PZR pressure does not meet the setpoint for a SIAS but containment pressure does. On a SIAS, all containment fan coolers will swap to slow speed. The discharge of the containment fan coolers will swap from the CCS ring header to CCS-102A and CCS-102B. These dampers fail open on a SIAS.

Technical Reference(s): OP-008-003 revision 302 (Attach if not previously provided) SD-CCS pg. 8,10 rev 7 (including version/revision number)

Proposed references to be provided to applicants during examination:

None Learning Objective: WLP-OPS-CCS00 (As available)

Question Source: Bank #

Modified Bank #

(Note changes or attach parent)

New X Question History: Last NRC Exam None Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7,8 55.43 Comments:

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 83 of 150 Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 026 A4.05 Importance Rating 3.5 K/A Statement A4.05: Ability to manually operate and/or monitor in the control room: Containment spray reset switches.

Proposed Question: RO 42 Rev: 0 Given: A Containment Spray Actuation Signal (CSAS) was inadvertently actuated The CRS directs the BOP operator to reset the CSAS The CSAS initiation relays are reset at ___(1)___ and the actuation relays are reset at ___(2)___.

(1) (2) A. CP-10 CP-33 B. CP-10 CP-10 C. CP-33 CP-33 D. CP-33 CP-10 2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 84 of 150 Proposed Answer: A Explanation: (Optional)

A. CORRECT: Per OP-901-504. Initiation relays are reset at CP-10 and actuation relays are reset at CP-33. B. Incorrect. Correct initiation relay reset location. Wrong actuation relay reset location. C. Incorrect. Wrong initiation relay reset location. Correct actuation relay reset location. D. Incorrect. Wrong initiation relay reset location. Wrong actuation relay reset location.

Technical Reference(s): OP-901-504, E2 Steps 9 and 10, rev 9 (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 #

Question #38 Modified Bank #

X (Note changes or attach parent)

New Question History: Last NRC Exam 2011 RO NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 Comments:

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 85 of 150 Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 039 K5.05 Importance Rating 2.7 K/A Statement K5.05: Knowledge of the operational implications of the following concepts as the apply to the MRSS: Bases for RCS cooldown limits.

Proposed Question: RO 43 Rev: 0 Given: Reactor Coolant System Tcold is 500F and lowering Reactor Coolant Pumps 1B and 2B are running 15 minutes ago the Reactor Coolant System Tc was 515F when a controlled cool down was established The cool down is being controlled using MS-319A, Main Steam Bypass 1A, at 15% open IF MS-319A failed open, the maximum allowed cool down rate of ___(1)____ which protects the ___(2)__________ under all conditions, would be exceeded.

(1) (2) A. 60F/Hr most limiting component in Reactor Coolant System B. 60F/Hr SG tube sheet from cyclic stress, which is the most limiting component C. 100F/Hr SG tube sheet from cyclic stress, which is the most limiting component D. 100F/Hr most limiting component in Reactor Coolant System

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 86 of 150 Proposed Answer: D Explanation: (Optional)

A. Incorrect. 60F/Hr is heat up rate. B. Incorrect. 60F/Hr is heat up rate and the tube sheet is not the most limiting component C. Incorrect. Tube sheet is not the most limiting component under all conditions D. CORRECT: 100F/Hr is the maximum allowed cool down rate and this protects the most limiting component from cyclic stress under all conditions Technical Reference(s): TS basis 3.4.8 (Attach if not previously provided) (including version/revision number)

Proposed references to be provided to applicants during examination:

None Learning Objective: WLP-OPS-RCS00 obj. 8 (As available)

Question Source: Bank #

X Question #43 Modified Bank #

(Note changes or attach parent)

New Question History: Last NRC Exam 2011 NRC RO Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 3 55.43 Comments:

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 87 of 150 Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 059 K4.16 Importance Rating 3.1 K/A Statement K4.16: Knowledge of MFW design feature(s) and/or interlock(s) which provide for the following: Automatic trips for MFW pumps.

Proposed Question: RO 44 Rev: 0 Given: The plant is at 42% power Condensate Pumps A, B and C are in operation Both Main Feedwater Pumps are operating Condensate Pumps A and B trip.

The Main Feedwater Pump Turbine Condensate Pump Interlock will trip Main Feedwater Pump _____(1)_____. This trip is active if the Main Feedwater Pump Turbine High Pressure (HP) governor valves are ______(2)_____. (1) (2) A. A not fully closed B. A fully closed C. B fully closed D. B not fully closed

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 88 of 150 Proposed Answer: D Explanation: (Optional)

A. Incorrect. If Condensate Pump C is the only pump left running, then Main Feedwater Pump B will trip. Part 2 is correct B. Incorrect. If Condensate Pump C is the only pump left running, then Main Feedwater Pump B will trip. both HP governor valves must be not closed for the Main Feedwater Pump turbine condensate pump interlock to be active C. Incorrect. If Condensate Pump C is the only pump left running, then Main Feedwater Pump B will trip. both HP governor valves must be not closed for the Main Feedwater Pump turbine condensate pump interlock to be active D. CORRECT: Per OP-003-033 automatic action 9.1.7 and limitation 3.2.3, both HP governor valves must be not closed for the Main Feedwater Pump turbine condensate pump interlock to be active. If Condensate Pump C is the only pump left running, then Main Feedwater Pump B will trip.

Technical Reference(s): OP-003-033, automatic action 9.1.15 and limitation 3.2.3, revision 320 (Attach if not previously provided) SD-FWP pg. 10, revision 4 (including version/revision number)

Proposed references to be provided to applicants during examination:

None Learning Objective: WLP-OPS-FWP00 obj. 2 (As available)

Question Source: Bank #

Modified Bank #

(Note changes or attach parent)

New X Question History: Last NRC Exam None Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 Comments:

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 89 of 150 Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 061 K6.01 Importance Rating 2.5 K/A Statement K6.01: Knowledge of the effect of a loss or malfunction of the following will have on the AFW components: Controllers and positioners.

Proposed Question: RO 45 Rev: 0 Given: The plant has a experienced a Loss of Main Feedwater Event An EFAS-1 has occurred Steam Generator #1 level is 40% WR and dropping Steam Generator #1 EFW flow transmitter has failed high The Steam Generator #1 Primary Flow Control Valve is ______(1)______. The Steam Generator #1 Backup Flow Control Valve is _______(2)______. (1) (2) A. closed open to 400 gpm flow value B. closed closed C. open to a preset valve position closed D. open to a preset valve position open to 400 gpm flow value

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 90 of 150 Proposed Answer: C Explanation: (Optional)

A. Incorrect: The EFW primary FCV will open to a preset valve position at 55% WR regardless of what EFW flow indicates. The EFW B/U FCV will open to maintain 400 gpm at 45% WR level. In this instance, the EFW flow transmitter is failed high, therefore, the B/U FCV input is telling it that no response is required, the B/U FCV remains closed. B. Incorrect. The EFW primary FCV will open to a preset valve position at 55% WR regardless of what EFW flow indicates. Part 2 is correct.

C. CORRECT: The EFW primary FCV will open to a preset valve position at 55% WR regardless of what EFW flow indicates. The EFW B/U FCV will open to maintain 400 gpm at 45% WR level. In this instance, the EFW flow transmitter is failed high, therefore, the B/U FCV input is telling it that no response is required, the B/U FCV remains closed. D. Incorrect. Part 1 is correct. The EFW B/U FCV will open to maintain 400 gpm at 45% WR level. In this instance, the EFW flow transmitter is failed high, therefore, the B/U FCV input is telling it that no response is required, the B/U FCV remains closed.

Technical Reference(s): SD-EFW pg. 26, rev12 (Attach if not previously provided) WLP-OPS-EFW00 slide 113 rev 36 (including version/revision number)

Proposed references to be provided to applicants during examination:

None Learning Objective: WLP-OPS-EFW00 obj. 5 (As available)

Question Source: Bank #

Modified Bank #

(Note changes or attach parent)

New X Question History: Last NRC Exam None Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Comments:

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 91 of 150 Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 062 A2.09 Importance Rating 2.7 K/A Statement A2.09: 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: Consequences of exceeding current limitations.

Proposed Question: RO 46 Rev: 0 Given: Power is 100% CEDMCS Motor Generator Tie Breaker (TCB 9) is removed for maintenance An equipment problem caused an overcurrent condition on the 3A-to-2A bus tie (BT). Overcurrent on the bus tie will trip the ____(1)____ . The crew will enter ______(2)______ to mitigate the event.

(1) (2) A. 3A-to-2A BT breaker only OP-901-310, Loss of Train A Safety Bus and OP-902-000, Standard Post Trip Actions concurrently B. 3A-to-2A BT breaker only OP-901-310, Loss of Train A Safety Bus only C. 3A-to-2A and the 2A-to-3A BT breakers OP-901-310, Loss of Train A Safety Bus and OP-902-000, Standard Post Trip Actions concurrently D. 3A-to-2A and the 2A-to-3A BT breakers OP-901-310, Loss of Train A Safety Bus only

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 92 of 150 Proposed Answer: D Explanation: (Optional)

A. Incorrect: First part is plausible because an undervoltage on the BT will only open the 3A-to-2A BT breaker but an overcurrent condition will open both BT breakers. Second part is plausible because given that TCB 9 is removed; a loss on MG A (upon the loss of 3A safety bus) appears to affect CEDMCS power and cause a reactor trip, but it does not. MG B will continue to supply power to all CEAs. B. Incorrect: First part is plausible because an undervoltage on the BT will only open the 3A-to-2A BT breaker but an overcurrent condition will open both BT breakers. Second part is correct. C. Incorrect: First part is correct. Second part is plausible because given that TCB 9 is removed; a loss on MG A (upon the loss of 3A safety bus) appears to affect CEDMCS power and cause a reactor trip, but it does not. MG B will continue to supply power to all CEAs.

D. CORRECT: An overcurrent condition will open both BT breakers. Loss of one safety bus will not directly cause a reactor trip. OP-901-310 provides adequate guidance to mitigate this event.

Technical Reference(s): SD-4KV pg.20-21 (rev 6) (Attach if not previously provided) OP-901-310 pp. 2,3 (rev 309) (including version/revision number)

Proposed references to be provided to applicants during examination:

None Learning Objective: WLP-OPS-ED00 obj.4 (rev20) (As available)

Question Source: Bank #

Modified Bank #

(Note changes or attach parent)

New X Question History: Last NRC Exam None Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Comments:

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 93 of 150 Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 062 2.2.36 Importance Rating 3.1 K/A Statement 2.2.36: Ability to analyze the effect of maintenance activities, such as degraded power sources, on the status of limiting conditions for operations.

Proposed Question: RO 47 Rev: 0 Given: Power is 100% W3 System Engineer and Transmission & Distribution workers are investigating an alarm in the 230KV yard The auxiliary operator reports a problem with EDG B during his rounds and the Control Room staff declares EDG B inoperable and enters the appropriate Tech Specs.

Both 230KV West Bus Feeder breakers trip. Both 230KV East Bus Feeder breakers remain closed.

Select the correct action:

A. Both Offsite Trains remain operable. Restore EDG B within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

B. Offsite Train A is inoperable. Restore EDG B or Offsite Train A within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. C. Both Offsite Trains are inoperable. Restore EDG B or either Offsite Train within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

D. Both Offsite Trains are inoperable. Enter Tech Spec 3.0.3 and prepare to commence a plant shut down in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 94 of 150 Proposed Answer: A Explanation: (Optional)

A. CORRECT: OP-903-066, Electrical Breaker Alignment Check, allows for both Offsite electrical trains to be supplied by the same 230KV bus (i.e. East or West) and still maintain electrical separation and operability. EDG B would be the only inoperable power source. TS 3.8.1.1.b requires restoration within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. B. Incorrect. Plausible because west trains are normally associated with safety train A and TS 3.8.1.1.c requires restoration of either power source within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. C. Incorrect. Plausible because the applicant may determine that train separation does not exist and therefore both offsite electrical circuits are inoperable. TS 3.8.1.1.d has an action to restore an inoperable EDG within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. This action is preceded by "with one EDG inoperable, in addition to Action b or c above". The applicant may determine that action d covers inoperability in addition to action c. D. Incorrect. Plausible because the applicant may determine that train separation does not exist and therefore both offsite electrical circuits are inoperable. The applicant may determine that TS 3.8.1.1 does not address one EDG and both offsite circuits being inoperable simultaneously and therefore TS 3.0.3 would apply.

Technical Reference(s): TS 3.8.1.1 (Attach if not previously provided) OP-903-066 (rev 302) (including version/revision number)

Proposed references to be provided to applicants during examination: TS 3.8.1.1; Att. 10.1 & 10.2 of OP-903-066 Learning Objective: WLP-OPS-TS04 obj. 1 (As available)

WLP-OPS-ED00 obj. 8 Question Source: Bank #

Modified Bank #

(Note changes or attach parent)

New X Question History: Last NRC Exam None Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 Comments:

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 95 of 150 Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 063 K3.01 Importance Rating 3.7 K/A Statement K3.01: Knowledge of the effect that a loss or malfunction of the DC electrical system will have on the following: ED/G. Proposed Question: RO 48 Rev: 0 Given: Emergency Diesel Generator 'A' is running loaded A loss of the 125VDC A Bus occurs Which of the following describes the effect of the loss of DC control power on the EDG and its auxiliaries?

A. The EDG must be secured locally by pulling the overspeed trip.

B. The EDG CCW flow control valve, CC-413A, will fail full open.

C. Fuel oil transfer pump starts and must be secured to prevent overfilling the feed tank. D. Jacket cooling water valves fail open and the jacket water heater loses power.

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 96 of 150 Proposed Answer: C Explanation: (Optional)

A. Incorrect. The EDG trips on a loss of DC power. The 20SD solenoid opens to shutdown the engine. B. Incorrect. CC-413 A does fail open on a loss of power but is powered from PDP-90A, which is powered from the SUPS.

C. CORRECT: On a loss of DC power, the fuel oil transfer pump starts and must be secured by opening the breaker to it. D. Incorrect. Jacket water heater is AC powered.

Technical Reference(s): OP-901-313 section E1 step 5 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-EDG00 obj. 2 (As available)

Question Source: Bank #

X Question 52 Modified Bank #

(Note changes or attach parent)

New Question History: Last NRC Exam 2007 RO Makeup Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 8 55.43 Comments:

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 97 of 150 Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 064 A2.09 Importance Rating 3.1 K/A Statement A2.09: 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: Synchronization of the ED/G with other electric power supplies. Proposed Question: RO 49 Rev: 0 Given: EDG B is being synchronized to the grid for a retest following maintenance The synchroscope is slowly rotating in the counter-clockwise (Slow) direction. EDG voltage is approximately 10 volts lower than Bus voltage IF the EDG output breaker is closed under these conditions, the potential effect of an (a) ____(1)____ trip exists.

Taking action to raise EDG ______(2)______ PRIOR to closing the EDG output breaker will prevent the effect.

(1) (2) A. overcurrent frequency B. overcurrent voltage C. reverse power frequency D. reverse power voltage 2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 98 of 150 Proposed Answer: C Explanation: (Optional)

A. Incorrect: Overcurrent is not a concern with EDG voltage 10 volts below bus voltage. Plausible because EDG voltage should be match or slightly higher. Second part is correct. B. Incorrect: Overcurrent is not a concern with EDG voltage 10 volts below bus voltage. Plausible because EDG voltage should be match or slightly higher. Second part - raising voltage will not correct the synchroscope rotation.

C. CORRECT: Synchroscope rotating in the counter-clockwise (Slow) direction has the potential of tripping the EDG output breaker on reverse power. Raising EDG speed (frequency) will correct the rotation. D. Incorrect: First part is correct. Second part - raising voltage will not correct the synchroscope rotation.

Technical Reference(s): OP-009-002, Precaution 3.1.7 (rev 325) (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. 2 & 8 (As available)

Question Source: Bank #

X Question #22 Modified Bank #

(Note changes or attach parent)

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

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 99 of 150 Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 064 2.1.27 Importance Rating 3.9 K/A Statement 2.1.27: Knowledge of system purpose and/or function.

Proposed Question: RO 50 Rev: 0 Given: The crew is performing OP-903-068, Emergency Diesel Generator Operability and Subgroup Relay Operability Verification, for Emergency Diesel Generator B The crew has started Emergency Diesel Generator B in accordance with OP-009-002, Emergency Diesel Generator OP-009-002 Attachment 11.8, Emergency Diesel Generator Start Evaluation, states that a successful start of Emergency Diesel Generator is defined as required voltage and frequency is ____(1)____ within ______(2)______ of taking the EDG B control switch to start. (1) (2) A. attained 5 seconds B. stabilized 5 seconds C. stabilized 10 seconds D. attained 10 seconds

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 100 of 150 Proposed Answer: D Explanation: (Optional)

A. Incorrect. Part 1 is correct. Five seconds is listed as precaution 3.1.7 in the same procedure, but has to do with the time limit for loading to prevent a diesel trip. B. Incorrect. The start evaluation does require the crew to obtain the time for voltage and frequency to stabilize but a note at the bottom of the attachment states that this time is not required for operability. Five seconds is listed as precaution 3.1.7 in the same procedure, but has to do with the time limit for loading to prevent a diesel trip. C. Incorrect. The start evaluation does require the crew to obtain the time for voltage and frequency to stabilize but a note at the bottom of the attachment states that this time is not required for operability. Part 2 is correct D. CORRECT: The definition of a successful start of a EDG is defined in OP-009-002 attachment 11.8, Emergency Diesel Generator Start Evaluation. It is defined as any start where the required voltage and frequency is attained within 10 seconds.

Technical Reference(s): OP-009-002, Att.11.8 revision 325 (Attach if not previously provided) (including version/revision number)

Proposed references to be provided to applicants during examination:

None Learning Objective: WLP-OPS-EDG obj. 8 (As available)

Question Source: Bank #

Modified Bank #

(Note changes or attach parent)

New X Question History: Last NRC Exam None Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 Comments:

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 101 of 150 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 51 Rev: 0 Given: The plant is at 40% power Blowdown discharge to Circ Water is in progress BD-303, Blowdown to Circ Water and Metal Waste Pond Isolation Valve, closes Which ONE of the following caused the valve to close?

A. High radiation on the Blowdown Radiation Monitor B. High Radiation on the Circ Water Radiation Monitor C. Trip of the running Blowdown Pump D. Hi pH alarm on Blowdown Proportional Sampler

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 102 of 150 Proposed Answer: B Explanation: (Optional)

A. Incorrect. The blowdown radiation monitor does not send signals for any components to actuate. Alarm only. B. CORRECT: High Radiation on the Circ Water Radiation Monitor will automatically close BD-303 isolating blowdown flow to the circ water system. C. Incorrect. Blowdown pump tripping does not send a signal to close BD-303. D. Incorrect. Hi pH alarm on Blowdown Proportional Sampler is an actual alarm but does not cause closure of BD-303.

Technical Reference(s): OP-003-010 step 9.2.5 revision 28 (Attach if not previously provided) (including version/revision number)

Proposed references to be provided to applicants during examination:

None Learning Objective: WLP-OPS-BD00 obj. 1 (As available)

Question Source: Bank #

X Question #25 Modified Bank #

(Note changes or attach parent)

New Question History: Last NRC Exam 2007 NRC RO Makeup Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7,11 55.43 Comments:

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 103 of 150 Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 073 A4.03 Importance Rating 3.1 K/A Statement A4.03: Ability to manually operate and/or monitor in the control room:

Check source for operability demonstration.

Proposed Question: RO 52 Rev: 0 Given: The crew is preparing to discharge Waste Condensate Tank A to Circ Water per OP-007-004, Liquid Waste Management System The crew can perform a source check for LWM Radiation Monitor, PRM-IRE-0647 from ____(1)____. A successful source check will result in a _____(2)_____ for the channel under test.

(1) (2) A. RM-11 at CP-6 only green light B. RM-11 at CP-6 or locally at the RM-80 green light C. RM-11 at CP-6 or locally at the RM-80 medium blue light D. RM-11 at CP-6 only medium blue light

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 104 of 150 Proposed Answer: B Explanation: (Optional)

A. Incorrect: OP-007-004 section 6.10 states that the source check can be completed at the RM-11 on CP-6 are locally at the RM-80. Part 2 is correct. B. CORRECT: OP-007-004 section 6.10 states that the source check can be completed at the RM-11 on CP-6 are locally at the RM-80. The note prior to Step 6.3.3.5 of OP-004-001 states that a successful source check will indicate green while a failed source check will indicate medium blue. C. Incorrect. Part 1 is correct. The note prior to Step 6.3.3.5 of OP-004-001 states that a successful source check will indicate green while a failed source check will indicate medium blue. D. Incorrect. OP-007-004 section 6.10 states that the source check can be completed at the RM-11 on CP-6 are locally at the RM-80. The note prior to Step 6.3.3.5 of OP-004-001 states that a successful source check will indicate green while a failed source check will indicate medium blue.

Technical Reference(s): OP-004-001 step 6.3.3.5 revision 305 (Attach if not previously provided) OP-007-004 step 6.10.5 revision 308 (including version/revision number)

Proposed references to be provided to applicants during examination:

None Learning Objective: WLP-OPS-RMS00 obj. 6 (As available)

Question Source: Bank #

Modified Bank #

(Note changes or attach parent)

New X Question History: Last NRC Exam None Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 11 55.43 Comments:

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 105 of 150 Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 076 K4.02 Importance Rating 2.9 K/A Statement K4.02: Knowledge of SWS design feature(s) and/or interlock(s) which provide for the following: Automatic start features associated with SWS pump controls.

Proposed Question: RO 53 Rev: 0 Which of the following conditions will send an automatic start signal to ACCW Pump A?

A. Low ACCW system pressure B. High ACCW system temperature C. Dry Cooling Tower A bypass opens D. Low Component Cooling Water flow

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 106 of 150 Proposed Answer: A Explanation: (Optional)

A. CORRECT: ACCW pressure of 5 psig or less at the CCW HX will automatically start the associated ACCW pump. B. Incorrect: High CCW (not ACCW) system temperature will auto start ACCW pump. C. Incorrect: Dry Cooling Tower bypass valve opening would cause CCW temperature to slowly trend up and may eventually cause an auto start of the ACCW pump, but bypass valve position does not directly send a start signal. D. Incorrect: Low CCW flow may also cause CCW temperature to slowly trend up, but CCW flow does not directly send a start signal.

Technical Reference(s): OP-002-001 pg. 27 (rev 307) (Attach if not previously provided) System Description SD-CC Tbl 1.28 (rev 21) (including version/revision number)

Proposed references to be provided to applicants during examination:

None Learning Objective: WLP-OPS-CC00 obj. 3 (rev 31) (As available)

Question Source: Bank #

X Question# 7 Modified Bank #

(Note changes or attach parent)

New Question History: Last NRC Exam 2010 RO NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7,8 55.43 Comments:

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 107 of 150 Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 078 A4.01 Importance Rating 3.1 K/A Statement A4.01: Ability to manually operate and/or monitor in the control room: Pressure gauges. Proposed Question: RO 54 Rev: 0 Given: Instrument Air Header pressure has lowered to 100 psig due to a leak SA-125, Station Air Backup is at its normal setpoint SA-125 is _____(1)_____ and SA-123, Air Dryer Bypass is ______(2)_____. (1) (2) A. open open B. closed open C. open closed D. closed closed 2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 108 of 150 Proposed Answer: C Explanation: (Optional)

A. Incorrect. SA-125 starts to open at 105 psig SA-123 opens at 95 psig. B. Incorrect. SA-125 starts to open at 105 psig SA-123 opens at 95 psig.

C. CORRECT. SA-125 starts to open at 105 psig SA-123 opens at 95 psig. D. Incorrect. SA-125 opens at 105 psig SA-123 opens at 95 psig.

Technical Reference(s): OP-901-511 section C 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-PPO05 obj. 3 (As available)

Question Source: Bank #

X Question #54 Modified Bank #

(Note changes or attach parent)

New Question History: Last NRC Exam 2011 NRC RO Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 4 55.43 Comments:

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 109 of 150 Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 103 K1.03 Importance Rating 3.1 K/A Statement K1.03: Knowledge of the physical connections and/or cause/effect relationships between the containment system and the following systems: Shield building vent system. Proposed Question: RO 55 Rev: 0 A plant trip and SIAS have occurred due to low Pressurizer pressure. What is the expected range of Containment Annulus to ambient differential pressure in inches of water? A. -3.0 to -8.0 B. +2.0 to -8.0 C. -3.0 to -10.0 D. +2.0 to -10.0

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 110 of 150 Proposed Answer: A Explanation: (Optional)

A. CORRECT: This is the range SBV will cycle on a SIAS. B. INCORRECT: The -8" is correct, but +2.0 inches is the maximum containment pressure that would ensure supply air to containment from the CAR System. C. INCORRECT: The -3.0" is correct but the10.0 inches is the maximum value for Containment pressure before containment purge can be initiated. D. INCORRECT: The +2.0 inches is the maximum containment pressure that would ensure supply air to containment from the CAR System and the 10.0 inches is the maximum value for Containment pressure before containment purge can be initiated.

Technical Reference(s): OP-008-008, Sect 9.0 revision 10 (Attach if not previously provided) (including version/revision number)

Proposed references to be provided to applicants during examination:

None Learning Objective: WLP-OPS-SBV00 obj 2 (As available)

Question Source: Bank #

X Question #55 Modified Bank #

(Note changes or attach parent)

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

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 111 of 150 Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 2 K/A # 002 K6.12 Importance Rating 3.0 K/A Statement K6.12: Knowledge of the effect or a loss or malfunction on the following RCS components: Code Safety valves.

Proposed Question: RO 56 Rev: 0 Which ONE of the following alarms would be the FIRST indication of a leaking Pressurizer Safety Valve?

A. Pressurizer Level High/Low B. Containment Water Leakage High C. Quench Tank Temperature High D. Reactor Drain Tank Level High/Low

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 112 of 150 Proposed Answer: C Explanation: (Optional)

A. Incorrect: For an RCS leak (within the capacity of the Charging Pumps), pressurizer level will be maintained by the Pzr Level Control System. B. Incorrect: Safety valves relieve directly to the Quench Tank (QT) not the containment atmosphere. Therefore containment water leakage high would not be expected until after the rupture disk on the QT ruptures (124 psig).

C. CORRECT: Quench Tank Temperature High at 200F indicates PZR Safety, PZR vent, or Rx Vessel Head Vent Leakage. D. Incorrect: Expected when in leakage is higher than pump capacity or pump failure. High level may occur when the QT is drained to the Reactor Drain Tank, but this is a manual action under the control of the operator.

Technical Reference(s): OP-500-008 Att 4.22 (rev 39) (Attach if not previously provided) OP-901-111 E0 Step 11 & Att. 1 pg.13 (rev 302) (including version/revision number) SD-RCS pg.27 (rev 20)

Proposed references to be provided to applicants during examination:

None Learning Objective: WLP-OPS-RCS00 Obj: 2 (As available)

Question Source: Bank #

X RO 56 Modified Bank #

(Note changes or attach parent)

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

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 113 of 150 Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 2 K/A # 011 A1.04 Importance Rating 3.1 K/A Statement A1.04: Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the PZR LCS controls including: T-ave. Proposed Question: RO 57 Rev: 0 Given: Plant is operating at 100% power TAVG LOOP SELECTOR switches in RRS Cabinets 1 and 2 are in the BOTH position RCS temperature Loop 1 Hot Leg (RC-ITI-0111-X) indicates failed low Letdown flow will initially ____(1)____. Pressurizer heaters will ______(2)_____ . (1) (2) A. lower de-energize B. lower energize C. rise energize D. rise de-energize

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 114 of 150 Proposed Answer: C Explanation: (Optional)

A. Incorrect. RCS Temperature Loop 1 Hot Leg failing low will cause the setpoint (based on Tave) for the PLCS to lower. Actual level will be greater than setpoint, therefore letdown flow will rise (not lower). Pressurizer heaters auto start on a 4% level insurge to the pressurizer. In this case, actual level will be greater than setpoint by 4%, heaters energize. The heaters would de-energize if the level indicator had failed low. B. Incorrect. RCS Temperature Loop 1 Hot Leg failing low will cause the setpoint (based on Tave) for the PLCS to lower. Actual level will be greater than setpoint, therefore letdown flow will rise (not lower). Part 2 is correct.

C. CORRECT: RCS Temperature Loop 1 Hot Leg failing low will cause the setpoint (based on Tave) for the PLCS to lower. Actual level will be greater than setpoint, therefore letdown flow will rise. Pressurizer heaters auto start on a 4% level insurge to the pressurizer. In this case, actual level will be greater than setpoint by 4%, heaters energize. D. Incorrect. Part 1 is correct. Pressurizer heaters auto start on a 4% level insurge to the pressurizer. In this case, actual level will be greater than setpoint by 4%, heaters energize. The heaters would de-energize if the level indicator had failed low.

Technical Reference(s): OP-901-110 Section E2 Revision 8 (Attach if not previously provided) (including version/revision number)

Proposed references to be provided to applicants during examination:

None Learning Objective: OP-901-PPO10 obj. 3 (As available)

Question Source: Bank #

Modified Bank #

(Note changes or attach parent)

New X Question History: Last NRC Exam None Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7,10 55.43 Comments:

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 115 of 150 Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 2 K/A # 015 A3.02 Importance Rating 3.7 K/A Statement A3.02: Ability to monitor automatic operation of the NIS, including: Annunciator and alarm signals.

Proposed Question: RO 58 Rev: 0 Given: A plant Startup to Mode 1 is in progress in accordance with OP-010-003, Plant Startup Reactor power is 10-4 % power and rising The LOGARITHMIC PWR HI BY-PASS annunciator is received The crew will _____(1)_____ on all four PPS Channels prior to exceeding ____(2)___ % power or a reactor trip will occur.

(1) (2) A. verify the High Log power trip is automatically bypassed 1.0 B. manually bypass the High Log power trip 1.0 C. manually bypass the High Log power trip 0.257 D. verify the High Log power trip is automatically bypassed 0.257 2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 116 of 150 Proposed Answer: C Explanation: (Optional)

A. Incorrect. The high log power trips are automatically placed in service during a downpower, but must be manually removed during an uppower. A reactor power of 1% is the next critical step to be performed in OP-010-003, but it is for aligning a Main Feedwater pump to service. B. Incorrect. Part 1 is correct. A reactor power of 1% is the next critical step to be performed in OP-010-003, but it is for aligning a Main Feedwater pump to service.

C. CORRECT: OP-010-003 has a caution before step 9.4.50 stating that the High Log Power Trips must be manually bypassed on all four PPS channels prior to exceeding 0.257% power or a reactor trip will occur. Step 9.4.50 directs the crew to bypass all four channels when the LOGARITHMIC PWR HI BY-PASS annunciator is received D. Incorrect. The high log power trips are automatically placed in service during a downpower, but must be manually removed during an uppower.OP-010-003 has a caution before step 9.4.50 stating that the High Log Power Trips must be manually bypassed on all four PPS channels prior to exceeding 0.257% power or a reactor trip will occur. Part 2 is correct Technical Reference(s): OP-010-003 caution and step 9.4.50 rev. 335 (Attach if not previously provided) OP-500-009 attachment 4.33 rev. 5 (including version/revision number)

Proposed references to be provided to applicants during examination:

None Learning Objective: WLP-OPS-ENI00 obj. 7 (As available)

Question Source: Bank #

Modified Bank #

(Note changes or attach parent)

New X Question History: Last NRC Exam None Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 2 55.43 Comments:

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 117 of 150 Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 2 K/A # 028 2.1.20 Importance Rating 4.6 K/A Statement 2.1.20 Ability to interpret and execute procedure steps.

Proposed Question: RO 59 Rev: 0 Given: A LOCA has occurred The crew is performing OP-902-002, Loss of Coolant Accident Recovery Which ONE of the following describes the sequence of actions required to place a Hydrogen Analyzer in service, and the hydrogen concentration indicated when the containment atmosphere reaches a flammable limit?

A. Open the H2 Analyzer containment isolation valve, then turn the power switch ON; 3%. B. Place the power switch ON, then open the H2 Analyzer containment isolation valve; 3%.

C. Open the H2 Analyzer containment isolation valve, then turn the power switch ON; 4%. D. Place the power switch ON, then open the H2 Analyzer containment isolation valve; 4%.

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 118 of 150 Proposed Answer: C Explanation: (Optional)

A. Incorrect. Sequence is correct and important because pump turns on when power turned on, if the valve is not open, there is no flowpath for pump. 3% is when the hydrogen analyzer Cntmt Hydrogen Hi annunciator is received, but flammable limit is 4%. B. Incorrect. Using this sequence, there would be no flow path for the sample pump. The sample pump starts when the analyzer is turned on. 3% is when the hydrogen analyzer Cntmt Hydrogen Hi annunciator is received, but flammable limit is 4%.

C. CORRECT: Hydrogen analyzer pump turns on when power is turned on, if the containment isolation valves are not open, there is no flowpath for the pump. D. Incorrect. Using this sequence, there would be no flow path for the sample pump. The sample pump starts when the analyzer is turned on. This is the correct concentration.

Technical Reference(s): OP-500-002 att 4.8 revision 22 (Attach if not previously provided) (including version/revision number)

Proposed references to be provided to applicants during examination:

None Learning Objective: WLP-OPS-HRA00 obj. 1,3 (As available)

Question Source: Bank #

X Question #32 Modified Bank #

(Note changes or attach parent)

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

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 119 of 150 Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 2 K/A # 035 A4.01 Importance Rating 3.7 K/A Statement A4.01: Ability to manually operate and/or monitor in the control room: Shift of S/G controls between manual and automatic control, by bumpless transfer.

Proposed Question: RO 60 Rev: 0 Given: Plant is at 55% power Steam Generator Feed Pump (SGFP) B has been started in accordance with OP-003-033, Main Feedwater Local speed balance meter is balanced and local controls are in Auto Main Feed Pump B speed is 3900 rpm and the red light on the Speed Governor Control Switch for Main Feed Pump B is illuminated The CRS has directed the BOP to align SGFP B for automatic operation The BOP will manually adjust SGFP B speed using the ____(1)____ . Before placing SGFP B in auto on CP-1, SGFP B output will be adjusted to match _____(2)_____ to assure minimal feedwater flow perturbation.

(1) (2) A. speed controller M/A station SGFP A output B. speed governor control switch SGFP A output C. speed controller M/A station SGFP B input D. speed governor control switch SGFP B input

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 120 of 150 Proposed Answer: C Explanation: (Optional)

A. Incorrect. Part 1 is correct. Matching SGFP A's output is plausible since this would match both pump discharge pressures, but as soon as, SGFP B controller is placed in auto the pump would drive itself to its own auto input signal. B. Incorrect. The speed governor control switch is initially used to manually raise SGFP speed to 3900 rpm. Once the local controls are placed in auto, speed control is transferred to the speed controller (the control switch no longer has any effect on speed). Matching SGFP A's output is plausible since this would match both pump discharge pressures, but as soon as, SGFP B controller is placed in auto the pump would drive itself to its own auto input signal.

C. CORRECT: Once the local speed controls are placed in auto, speed control is transferred to the speed controller (Manual/Auto station) and the controller is used to manually adjust speed. A note prior to step 6.3.24 of OP-003-033 provides three methods that can be used to minimize feedwater flow perturbations. Matching speed controller output to its own input is one of those methods. D. Incorrect. Once the local speed controls are placed in auto, speed control is transferred to the speed controller and the controller is used to manually adjust speed. Part 2 is correct.

Technical Reference(s): OP-003-033 note prior to step 6.3.24 revision 319 (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. 11 (As available)

Question Source: Bank #

Modified Bank #

(Note changes or attach parent)

New X Question History: Last NRC Exam None Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Comments:

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 121 of 150 Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 2 K/A # 016 K4.03 Importance Rating 2.8 K/A Statement K4.03: Knowledge of NNIS design feature(s) and/or interlock(s) which provide for the following: Input to control systems.

Proposed Question: RO 61 Rev: 0 RCS Reference Temperature (Tref) signal is generated in the ____(1)____ system as a function of the _____(2)_____ input.

(1) (2) A. Reactor Regulating Main Steam Crossover Header pressure B. Steam Bypass Control Turbine First Stage pressure C. Reactor Regulating Turbine First Stage pressure D. Steam Bypass Control Main Steam Crossover Header pressure 2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 122 of 150 Proposed Answer: C Explanation: (Optional)

A. Incorrect: Reactor Regulating System takes input from Turbine First Stage Pressure (PTFS) to calculate the current Tref but does not use MS Crossover Header Pressure. B. Incorrect: Steam Bypass Control uses Turbine First Stage Pressure as an input to the AMI circuitry but it does not calculate Tref in Steam Bypass Control.

C. CORRECT: Reactor Regulating System takes input from Turbine First Stage Pressure (PTFS) to calculate the current Tref. D. Incorrect: Steam Bypass Control input from Main Steam Crossover Header pressure for Master Controller permissive input is not part of the Reactor Regulating System.

Technical Reference(s): SD-RR Fig 02 (rev 6) (Attach if not previously provided) SD-SBC Fig 02 (rev 10) (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 #

X Question #30 Modified Bank #

(Note changes or attach parent)

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

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 123 of 150 Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 2 K/A # 055 K1.06 Importance Rating 2.6 K/A Statement K1.06: Knowledge of the physical connections and/or cause/effect relationships between the CARS and the following systems: PRM system.

Proposed Question: RO 62 Rev: 0 OP-003-001, Condenser Air Evacuation System, requires Condenser Vacuum Pumps (1) be operating to provide a suction path for the (2) (1) (2) A. A or B Condenser Air Evacuation WRGM B. A or B Condenser Air Evacuation PIG Rad Monitor C. B or C Condenser Air Evacuation WRGM D. B or C Condenser Air Evacuation PIG Rad Monitor

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 124 of 150 Proposed Answer: D Explanation: (Optional)

A. Incorrect. Condenser Air Evacuation (WRGM) gaseous detector monitors non-condensable gas exhaust in the common header and discharges to atmosphere. B. Incorrect. Air Evacuation Discharge (Gaseous) PIG detector (PRM-IR-0004) monitors discharge of the common header. C. Incorrect. Condenser Air Evacuation (WRGM) gaseous detector monitors non-condensable gas exhaust in the common header and discharges to atmosphere.

D. CORRECT: Air Evacuation Discharge (Gaseous) PIG detector (PRM-IR-0004) monitors discharge of the common header.

Technical Reference(s): OP-003-001 note on page 10 revision 18 (Attach if not previously provided) (including version/revision number)

Proposed references to be provided to applicants during examination:

None Learning Objective: WLP-OPS-AE00 obj. 1 (As available)

Question Source: Bank #

X Question #36 Modified Bank #

(Note changes or attach parent)

New Question History: Last NRC Exam 2010 NRC RO Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 11 55.43 Comments:

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 125 of 150 Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 2 K/A # 001 K5.08 Importance Rating 3.9 K/A Statement K5.08 Knowledge of the operational implication of the following concepts as they apply to the CRDS: Reason for rod insertion limits and their effect on shutdown margin.

Proposed Question: RO 63 Rev: 0 In Modes 1 and 2, the Transient Insertion Limits of TS 3.1.3.6 (Regulating and Group P CEA Insertion Limits) and TS 3.1.3.5 (Shutdown CEA Insertion Limits) ensures that____(1)____. If these limits cannot be maintained, the crew is required to ______(2)_____ . (1) (2) A. radial xenon redistribution effects are maintained within limits emergency borate B. radial xenon redistribution effects are maintained within limits restore the CEA to within limits within two hours C. minimum shutdown margin requirements are maintained restore the CEA to within limits within two hours D. minimum shutdown margin requirements are maintained emergency borate

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 126 of 150 Proposed Answer: D Explanation: (Optional)

A. Incorrect. radial xenon redistribution effects are maintained within limits is the basis for maintaining CEAs above the long term and short term insertion limits. Part 2 is correct. B. Incorrect. radial xenon redistribution effects are maintained within limits is the basis for maintaining CEAs above the long term and short term insertion limits. The crew is allowed two hours to restore CEAs above the transient insertion limit only if a reactor power cutback has occurred or surveillance testing. Neither are occurring in this instance. C. Incorrect. Part 1 is correct. The crew is allowed two hours to restore CEAs above the transient insertion limit only if a reactor power cutback has occurred or surveillance testing. Neither are occurring in this instance D. CORRECT: The basis for TS 3.1.3.5 and TS 3.1.3.6 defines the reasons for the transient insertion limits (ensures adequate shutdown margin). If shutdown margin cannot be maintained the action is to commence emergency boration.

Technical Reference(s): TS 3.1.3 basis (Attach if not previously provided) TS 3.1.3.6 (including version/revision number) OP-901-103 revision 3 Proposed references to be provided to applicants during examination:

None Learning Objective: WLP-OPS-CED00 obj. 8 (As available)

Question Source: Bank #

Modified Bank #

(Note changes or attach parent)

New X Question History: Last NRC Exam None Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 5,10 55.43 Comments:

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 127 of 150 Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 2 K/A # 071 K3.05 Importance Rating 3.2 K/A Statement K3.05: Knowledge of the effect that a loss or malfunction of the Waste Gas Disposal System will have on the following: ARM and PRM systems.

Proposed Question: RO 64 Rev: 0 Which ONE of the following radiation monitors will cause an AUTO closure of Waste Gas Discharge FCV (GWM-309) valve during a Gaseous Waste Discharge evolution?

A. Waste Gas Discharge Rad Monitor B. Plant Stack PIG Monitor A or B C. Plant Stack WRGM Monitor D. RAB HVAC PIG Monitor B

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 128 of 150 Proposed Answer: A Explanation: (Optional)

A. CORRECT: Hi-Hi activity on the Waste Gas Discharge Rad Monitor will isolate the discharge by closing GWM-309. There is a malfunction of the GWM system if high activity occurs (GDT inlet valves leaking by, GST isolations leaking by, etc.) B. Incorrect. The Plant Stack PIGs are mentioned in the Waste Gas Discharge offnormal to determine length and severity of the discharge. Their auto function is to secure containment purge. C. Incorrect. Hi-Hi activity on the Plant Stack WRGM Rad Monitor will not isolate the discharge. Although, it will see the rise in activity. D. Incorrect. RAB HVAC PIG Monitor B will monitor the area of the GWM Radiation monitor but it does not have an auto function.

Technical Reference(s): OP-007-003 section 9 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-RMS00 obj.3 (As available)

Question Source: Bank #

X Question #64 Modified Bank #

(Note changes or attach parent)

New Question History: Last NRC Exam 2009 NRC RO Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 11 55.43 Comments:

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 129 of 150 Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 2 K/A # 075 K2.03 Importance Rating 2.6 K/A Statement K2.03: Knowledge of bus power supplies to the following: Emergency/essential SWS pumps.

Proposed Question: RO 65 Rev: 0 Turbine Closed Cooling Water (TCCW) Pump B is powered from which bus? A. SWGR 21B B. SWGR 3B C. SWGR 1B D. SWGR 2B 2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 130 of 150 Proposed Answer: D Explanation: (Optional)

A. Incorrect: TCCW Pump B is powered from the 2B bus. The 21B is a 480 V safety bus fed from the 2B bus. B. Incorrect: TCCW Pump B is powered from the 2B bus. The 3B bus is a 4.16KV safety bus C. Incorrect: TCCW Pump B is powered from the 2B bus. The 1B bus is a 6.9 Kv which carries larger loads.

D. CORRECT: TCCW Pump B is powered from the 2B bus. The 2B bus is a 4.16 Kv bus. Technical Reference(s): OP-003-027 Attachment 11.2 Revision 15 (Attach if not previously provided) (including version/revision number)

Proposed references to be provided to applicants during examination:

None Learning Objective: WLP-OPS-TC00 obj. 3 (As available)

Question Source: Bank #

Question 64 Modified Bank #

(Note changes or attach parent)

New X Question History: Last NRC Exam None Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 4 55.43 Comments:

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 131 of 150 Examination Outline Cross-

Reference:

Level RO SRO Tier # 3 Group # 1 K/A # 2.1.1 Importance Rating 3.8 K/A Statement 2.1.1 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 permission 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.

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 132 of 150 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, step 5.2 Revision 15 (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 #

X Question 66 Modified Bank #

(Note changes or attach parent)

New Question History: Last NRC Exam 2012 RO Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Comments:

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 133 of 150 Examination Outline Cross-

Reference:

Level RO SRO Tier # 3 Group # 1 K/A # 2.1.42 Importance Rating 2.5 K/A Statement 2.1.42 Knowledge of new and spent fuel movement procedures.

Proposed Question: RO 67 Rev: 0 Per TS 3.9.10.1 Water Level-Reactor Vessel and RF-005-001 Fuel Movement, at least 23 feet of water shall be maintained over the top of the ____(1)_____ during movements of fuel within the reactor pressure vessel. This is equivalent to a ____(2)____ foot level in the Refuel Cavity.

(1) (2) A. reactor vessel flange 32 B. reactor vessel flange 43 C. irradiated fuel assemblies seated in the reactor vessel 43 D. irradiated fuel assemblies seated in the reactor vessel 32 2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 134 of 150 Proposed Answer: B Explanation: (Optional)

A. Incorrect. Part 1 is correct. The vessel flange is at the 20 foot level. Therefore, the refuel cavity must be at 43 feet. The top of the fuel assemblies in the vessel is 9 feet. RF-005-01, Fuel Movement, requires 23 feet above of the assemblies to move CEAs in the vessel. This would be equivalent to 32 feet in the refuel cavity. B. CORRECT: Per TS 3.9.10.1 Water Level-Reactor Vessel and RF-005-001 Fuel Movement, at least 23 feet of water shall be maintained over the top of the reactor pressure vessel flange The vessel flange is at the 20 foot level. Therefore, the refuel cavity must be at 43 feet. C. Incorrect. The top of the fuel assemblies in the vessel is 9 feet. RF-005-01, Fuel Movement, requires 23 feet above of the assemblies to move CEAs in the vessel. Part 2 is correct. D. Incorrect. The top of the fuel assemblies in the vessel is 9 feet. RF-005-01, Fuel Movement, requires 23 feet above of the assemblies to move CEAs in the vessel. This would be equivalent to 32 feet in the refuel cavity. The question is asking required level to move fuel assemblies, not CEAs.

Technical Reference(s): RF-005-001 limitation 3.1.6 revision 313 (Attach if not previously provided) TS 3.9.10.1 (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 #

Modified Bank #

(Note changes or attach parent)

New X Question History: Last NRC Exam None Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Comments:

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 135 of 150 Examination Outline Cross-

Reference:

Level RO SRO Tier # 3 Group # 1 K/A # 2.1.45 Importance Rating 4.3 K/A Statement 2.1.45 Ability to identify and interpret diverse indications to validate the response of another indication.

Proposed Question: RO 68 Rev: 0 Given: A reactor trip has occurred The crew is performing OP-902-000, Standard Post Trip Actions For the RCS Inventory Control Safety Function, the ATC will verify that RCS subcooling is greater than or equal to ____(1)____ °F. RCS subcooling greater than or equal to the minimum RCS subcooling requirements for RCS inventory control will ______(2)_____ . (1) (2) A. 28 validate pressurizer level indication as being representative of total RCS inventory B. 30 validate pressurizer level indication as being representative of total RCS inventory C. 28 ensure that Reactor Coolant Pumps are within the acceptable range of the RCP operating curves D. 30 ensure that Reactor Coolant Pumps are within the acceptable range of the RCP operating curves 2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 136 of 150 Proposed Answer: A Explanation: (Optional)

A. CORRECT: RCS inventory control safety function requires the RCS to be >

28°F subcooled. The tech guide for standard post trip actions states that this verification for subcooling is to validate pressurizer level indication as being representative of total RCS inventory. (no voiding in the head). This question is TIER 3 because the inventory control SFSC and immediate action is required for any transient involving a reactor trip. This question is comprehensive since the applicant is required to know the reason for subcooled margin requirement in RCS inventory control, which can only be found in the Tech guide. B. Incorrect. 30°F is plausible since this is the setpoint of the subcooled margin Lo annunciator on CP-8. Part 2 is correct. C. Incorrect. Part 1 is correct. The crews will use the RCP operating curves to determine if RCPs are required to be secured. Although, the subcooled margin Lo annunciators are used to clue them that the RCP operating curves have been exceeded. D. Incorrect. 30°F is plausible since this is the setpoint of the subcooled margin Lo annunciator on CP-8. The crews will use the RCP operating curves to determine if RCPs are required to be secured. Although, the subcooled margin Lo annunciators are used to clue them that the RCP operating curves have been exceeded.

Technical Reference(s): TGOP-902-000 step number 3 revision 303 (Attach if not previously provided) OP-902-000 revision 15 step 3 (including version/revision number) OP-500-011 M-7 revision 34 Proposed references to be provided to applicants during examination:

None Learning Objective: WLP-OPS-PPE01 obj. 11 (As available)

Question Source: Bank #

Modified Bank #

(Note changes or attach parent)

New X Question History: Last NRC Exam None Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 Comments:

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 137 of 150 Examination Outline Cross-

Reference:

Level RO SRO Tier # 3 Group # 2 K/A # 2.2.21 Importance Rating 2.9 K/A Statement 2.2.21 Knowledge of pre- and post-maintenance operability requirements.

Proposed Question: RO 69 Rev: 0 Given: Plant is at 100% Power At 0100, the BOP operator bypasses the Channel C High LPD and Low DNBR trip bistables for a scheduled two hour I&C surveillance on Channel C Core Protection Calculator The I&C technician informs the CRS that the CPC has failed the surveillance and will require a card replacement Operability will be tracked by ___(1)___ when the work begins. The Shift Manager will authorize operability following retest using ___(2)___. (1) (2) A. the work package the CR operability tab in PCRS B. OP-100-010, Att. 7.1, TS/TRM Entry Guidelines OP-100-010, Att. 7.2, EOS Checklist C. the work package OP-100-010, Att. 7.2, EOS Checklist D. OP-100-010, Att. 7.1, TS/TRM Entry Guidelines the CR operability tab in PCRS

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 138 of 150 Proposed Answer: B Explanation: (Optional)

A. Incorrect: The work package is not used to track operability. OP-100-010 ATT 7.2 is used when a retest (other than a Channel Check) is required for operability. A CR is initiated in PCRS for any TS component that fails a surveillance, but PCRS is not used to authorize operability of equipment. B. CORRECT: IAW OP-100-010 the initial work is performed using OP-100-010 Att. 7.1 TS/TRM Entry guidelines. When a Tech Spec component fails a surveillance and a retest is required for operability, OP-100-010 ATT 7.2 EOS checklist is required. Tier 3 because any TS component is handled the same way. C. Incorrect: The work package is not used to track operability. Second part is correct. D. Incorrect: First part is correct. A CR is initiated in PCRS for any TS component that fails a surveillance, but PCRS is not used to authorize operability of equipment.

Technical Reference(s): OP-100-010 , steps 5.1.1 and 5.1.2 revision 311 (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 #

X Question 70 Modified Bank #

(Note changes or attach parent)

New Question History: Last NRC Exam 2014 NRC RO Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Comments: revised per Facility Reviewer comments

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 139 of 150 Examination Outline Cross-

Reference:

Level RO SRO Tier # 3 Group # 2 K/A # 2.2.40 Importance Rating 3.4 K/A Statement 2.2.40 Ability to apply Technical Specifications for a system.

Proposed Question: RO 70 Rev: 0 Per TS 3.6.3 basis, a locked closed containment isolation valve can be opened on an intermittent basis under administrative controls.

TS 3.6.3 basis states that the administrative controls will include assigning an operator in constant communication with the control room ____(1)____. The list of containment isolation valves can be found in _____(2)____ . (1) (2) A. to close the valve within four hours of an accident situation TS 3.6.3 B. to be stationed at the valve controls TS 3.6.3 C. to close the valve within four hours of an accident situation TRM 3.6.3 D. to be stationed at the valve controls TRM 3.6.3

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 140 of 150 Proposed Answer: D Explanation: (Optional)

A. Incorrect. Four hours to close the containment isolation valve is plausible because this is the limiting time per TS 3.6.3 to close the CIV if the loop is an open system. (TS 3.6.3e). TS 3.6.3 is a plausible distractor because the list of Containment Isolation Valves were listed here at one time before being moved to TRM 3.6.3. B. Incorrect. Part 1 is correct. TS 3.6.3 is a plausible distractor because the list of Containment Isolation Valves were listed here at one time before being moved to TRM 3.6.3. C. Incorrect. Four hours to close the containment isolation valve is plausible because this is the limiting time per TS 3.6.3 to close the CIV if the loop is an open system. (TS 3.6.3e). Part 2 is correct.

D. CORRECT: TS 3.6.3 basis list the requirements for administrative controls when opening a locked closed containment isolation valve. Stationing an operator at the valve controls while in constant communication with the control room is included in this list. TS 3.6.3 basis also identifies that the list of containment isolation valves is located in TRM 3.6.3.

Technical Reference(s): TS 3.6.3 basis (Attach if not previously provided) (including version/revision number)

Proposed references to be provided to applicants during examination:

None Learning Objective: WLP-OPS-TS02 obj. 11 (As available)

Question Source: Bank #

Modified Bank #

(Note changes or attach parent)

New X Question History: Last NRC Exam None Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 3 55.43 Comments:

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 141 of 150 Examination Outline Cross-

Reference:

Level RO SRO Tier # 3 Group # 3 K/A # 2.3.11 Importance Rating 3.8 K/A Statement 2.3.11 Ability to control radiation releases.

Proposed Question: RO 71 Rev: 0 Given: A Steam Generator Tube Leak is in progress The crew is performing actions in accordance with OP-901-202, Steam Generator Tube Leakage or High Activity The condenser is not available The crew will commence a normal plant cooldown to obtain a Hot leg temperature of less than or equal to 520°F using _____(1)_____. The discharge through the Atmospheric Dump Valve(s) is a(an) ______(2)_____ release.

(1) (2) A. the atmospheric dump valve on the S/G with the lowest indicated activity monitored B. the atmospheric dump valve on the S/G with the lowest indicated activity unmonitored C. both atmospheric dump valves monitored D. both atmospheric dump valves unmonitored

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 142 of 150 Proposed Answer: B Explanation: (Optional)

A. Incorrect. Part 1 is correct. OP-901-202 step 20.5 reminds the crew that this path is an unmonitored release. B. CORRECT: OP-901-202 step 20.5 directs the crew to reduce Th to < 520°F using the atmospheric dump valve on the S/G with the lowest activity. This is performed if the condenser is unavailable. This same step reminds the crew that this path is an unmonitored release. C. Incorrect. OP-901-202 step 20.5 directs the crew to reduce Th to < 520°F using the atmospheric dump valve on the S/G with the lowest activity. This is performed if the condenser is unavailable. OP-902-007, S/G Tube rupture directs the crew to cooldown to < 520°F using both ADVs in manual. (that is not the case for a tube leak) This same step reminds the crew that this path is an unmonitored release. D. Incorrect. OP-901-202 step 20.5 directs the crew to reduce Th to < 520°F using the atmospheric dump valve on the S/G with the lowest activity. This is performed if the condenser is unavailable. OP-902-007, S/G Tube rupture directs the crew to cooldown to < 520°F using both ADVs in manual. (that is not the case for a tube leak). Part 2 is correct.

Technical Reference(s): OP-901-202, step 20.5 revision 15 (Attach if not previously provided) (including version/revision number)

Proposed references to be provided to applicants during examination:

None Learning Objective: WLP-OPS-PPO20 obj. 3 (As available)

Question Source: Bank #

Modified Bank #

(Note changes or attach parent)

New X Question History: Last NRC Exam None Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Comments:

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 143 of 150 Examination Outline Cross-

Reference:

Level RO SRO Tier # 3 Group # 3 K/A # 2.3.14 Importance Rating 3.4 K/A Statement 2.3.14 Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or emergency conditions or activities.

Proposed Question: RO 72 Rev: 0 As an Emergency Team member following a loss of reactor cavity level that has resulted in uncovering irradiated fuel, you will enter the Containment in order to rescue an injured person. The maximum exposure limit (TEDE) allowed for these conditions is _____(1)_____. Authorization ______(2)_____ require that dose rates in the area be known.

(1) (2) A. 10 rem does B. 10 rem does not C. 25 rem does D. 25 rem does not 2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 144 of 150 Proposed Answer: C Explanation: (Optional)

A. Incorrect: Procedural exposure limit is 25 rem TEDE for life saving activities. Second part is correct. B. Incorrect: Procedural exposure limit is 25 rem TEDE for life saving activities. Authorization requires that dose rates in the area be known/measurable.

C. CORRECT: For the purposes of life saving activities, procedural exposure limit is 25 rem (TEDE). Known/measurable dose rates is one of the conditions in EP-002-030, Att. 7.1, Emergency Exposure Authorization Form (step 5). D. Incorrect: First part is correct. Second part is incorrect because authorization requires that dose rates in the area be known/measurable.

Technical Reference(s): EP-002-030 step 5.2 & Att. 7.1 step 5 revision 10 (Attach if not previously provided) (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 None Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Comments:

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 145 of 150 Examination Outline Cross-

Reference:

Level RO SRO Tier # 3 Group # 4 K/A # 2.4.12 Importance Rating 4.0 K/A Statement 2.4.12 Knowledge of general operating crew responsibilities during emergency operations Proposed Question: RO 73 Rev: 0 In accordance with OI-038-000, Emergency Operating Procedures Operations Expectations/Guidance, the operating crew will monitor the safety function status checklist ____(1)____. The STA is required to monitor the safety function status checklist every ______(2)_____ minutes.

(1) (2) A. continuously 30 B. every 30 minutes 15 C. every 30 minutes 30 D. continuously 15 2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 146 of 150 Proposed Answer: D Explanation: (Optional)

A. Incorrect. OI-038-000, Step 5.4.63, states that the STA performs the initial assessment of the SFSC and then every 15 minutes. A recent procedure change changed the STA requirement from 30 minutes to 15 minutes. The operating crew should continuously monitor the SFSC. B. Incorrect. OI-038-000, Step 5.4.63, states that the STA performs the initial assessment of the SFSC and then every 15 minutes. The operating crew should continuously monitor the SFSC. C. Incorrect. OI-038-000, Step 5.4.63, states that the STA performs the initial assessment of the SFSC and then every 15 minutes. A recent procedure change changed the STA requirement from 30 minutes to 15 minutes. The operating crew should continuously monitor the SFSC.

D. CORRECT: OI-038-000, Step 5.4.63, states that the STA performs the initial assessment of the SFSC and then every 15 minutes. The operating crew should continuously monitor the SFSC.

Technical Reference(s): OI-038-000 step 5.4.63 revision 10 (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. 4 (As available)

Question Source: Bank #

Modified Bank #

(Note changes or attach parent)

New X Question History: Last NRC Exam None Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Comments:

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 147 of 150 Examination Outline Cross-

Reference:

Level RO SRO Tier # 3 Group # 4 K/A # 2.4.42 Importance Rating 2.6 K/A Statement 2.4.42 Knowledge of emergency response facilities.

Proposed Question: RO 74 Rev: 0 The lowest emergency classification that the Technical Support Center (TSC) and the Emergency Operations Facility (EOF) shall be activated is an ____(1)____. Continuous communications with the NRC will be transferred from the Control Room to the ____(2)_____. (1) (2) A. unusual event EOF B. unusual event TSC C. alert TSC D. alert EOF 2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 148 of 150 Proposed Answer: C Explanation: (Optional)

A. Incorrect The TSC and EOF can be activated any time a decision is made to activate them, including an Unusual Event. But, step 4.2 of both EP-002-100 and EP-002-102 states that upon an Alert, both facilities shall be activated. Continuous communications with the NRC is transferred to the TSC, communications with all other facilities is transferred to the EOF. B. Incorrect. The TSC and EOF can be activated any time a decision is made to activate them, including an Unusual Event. But, step 4.2 of both EP-002-100 and EP-002-102 states that upon an Alert, both facilities shall be activated. Part 2 is correct.

C. CORRECT: Per EP-002-100 and EP-002-102, both the EOF and TSC shall be activated at an alert, but could be activated at an Unusual Event. Continuous communications with the NRC is transferred to the TSC. D. Incorrect. Part 1 is correct. Continuous communications with the NRC is transferred to the TSC, communications with all other facilities is transferred to the EOF.

Technical Reference(s): EP-002-100 Revision 43 (Attach if not previously provided) EP-002-102 Revision 306 (including version/revision number)

Proposed references to be provided to applicants during examination:

None Learning Objective: WLP-OPS-EP02 obj.5, 9 (As available)

Question Source: Bank #

Modified Bank #

(Note changes or attach parent)

New X Question History: Last NRC Exam None Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Comments:

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 149 of 150 Examination Outline Cross-

Reference:

Level RO SRO Tier # 3 Group # 4 K/A # 2.4.49 Importance Rating 4.6 K/A Statement 2.4.49 Ability to perform without reference to procedures those actions that require immediate operation of system components and controls.

Proposed Question: RO 75 Rev: 0 Given: A reactor trip has occurred The crew is performing OP-902-000, Standard Post Trip Actions The minimum requirements to verify the main turbine is tripped is to check ____(1)____ are closed. If the main turbine did not trip automatically, the first contingency action to secure steam admission to the main turbine is to ______(2)_____. (1) (2) A. governor and throttle valves depress the turbine trip and think pushbuttons B. governor valves close both MSIVs C. governor and throttle valves close both MSIVs D. governor valves depress the turbine trip and think pushbuttons

2015 NRC Exam RO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 150 of 150 Proposed Answer: A Explanation: (Optional)

A. CORRECT: The BOP must check the main turbine tripped by verifying both governor valves and throttle valves are closed. OI-038-000, EOP Operations Expectations/Guidance procedure, states that "the sub steps listed for stopping steam admission to main turbine are listed in a preferred order. If one of the contingency actions results in tripping the main turbine, it is not necessary or desired to perform the remaining contingency steps." Tripping the main turbine using the turbine trip and think pushbuttons is listed before the contingency step for closing both MSIVs. This question is TIER 3 because this step is performed before any optimal recovery procedure entry and the guidance to answer the second portion correctly is contained in the EOP Operations Expectations/Guidance procedure. B. Incorrect. Verifying only the governor valves are closed does not comply with the minimum requirements of OP-902-000 to check the main turbine is tripped. The sub steps listed for stopping steam admission to the main turbine are listed in a preferred order as stated in OI-038-000. An attempt must be made to trip the main turbine first by using the trip pushbuttons. C. Incorrect. Part 1 is correct. The sub steps listed for stopping steam admission to the main turbine are listed in a preferred order as stated in OI-038-000. An attempt must be made to trip the main turbine first by using the trip pushbuttons before closing both MSIVs..

D. Incorrect. Verifying only the governor valves are closed does not comply with the minimum requirements of OP-902-000 to check the main turbine is tripped. Part 2 is correct Technical Reference(s): OI-038-000 step 5.2.3 revision 10 (Attach if not previously provided) OP-902-000 step 2 revision 15 (including version/revision number)

Proposed references to be provided to applicants during examination:

None Learning Objective: WLP-OPS-PPE01 obj. 4 (As available)

Question Source: Bank #

Modified Bank #

(Note changes or attach parent)

New X Question History: Last NRC Exam None Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Comments:

2015 NRC Exam SRO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 1 of 50 Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 000009 2.2.38 Importance Rating 4.5 K/A Statement 2.2.38: Knowledge of conditions and limitations in the facility license.

Proposed Question:

SRO 1 Rev: 0 Given: The plant is at 100 % power Atmospheric Dump Valve #1 (MS-116A) controller has been placed in manual in an attempt to more fully seat the valve The crew is evaluating Tech Spec 3.7.1.7, Atmospheric Dump Valve, requirements The automatic actuation channel for Atmospheric Dump Valve # 1 is ___(1)____ For the given power level, TS 3.7.1.7 basis states that the automatic operation of one Atmospheric Dump Valve along with one train of High Pressure Safety Injection is required for mitigation of the _____(2)_____ event.

(1) (2) A. inoperable steam generator tube rupture B. operable steam generator tube rupture C. inoperable small Break LOCA D. operable small Break LOCA

2015 NRC Exam SRO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 2 of 50 Proposed Answer: C Explanation: (Optional)

A. Incorrect. Part 1 is correct. A steam generator tube rupture event is plausible since it is an event that the RCS will lose inventory and both the ADV and HPSI is procedurally operated. B. Incorrect. Determining that the automatic actuation channel is operable is plausible since the operator could restore the ADV auto function simply by placing the controller back to auto. In this instance, nothing is wrong with the auto actuation inputs. A steam generator tube rupture event is plausible since it is an event that the RCS will lose inventory and both the ADV and HPSI is procedurally operated.

C. CORRECT: The basis for TS 3.7.1.7 states that "At greater than 70% rated thermal power, one high pressure safety injection train and one ADV, in automatic, are capable of mitigating the SBLOCA event. The basis also states that the action for automatic control of an ADV must be entered if the automatic controls for auto actuation are placed in manual. D. Incorrect. Determining that the automatic actuation channel is operable is plausible since the operator could restore the ADV auto function simply by placing the controller back to auto. In this instance, nothing is wrong with the auto actuation inputs. Part 2 is correct.

Technical Reference(s): TS 3.7.1.7 basis (Attach if not previously provided) (including version/revision number)

Proposed references to be provided to applicants during examination:

None Learning Objective: WLP-OPS-MS00 obj. 6 (As available)

Question Source: Bank #

Modified Bank #

(Note changes or attach parent)

New X Question History: Last NRC Exam None Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 2 Comments:

2015 NRC Exam SRO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 3 of 50 Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 000011 2.4.9 Importance Rating 4.2 K/A Statement 2.4.9: 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 2 Rev: 0 Given: Large Break LOCA has occurred The crew is performing the required actions of OP-902-002, LOCA Recovery Procedure The CRS will determine if simultaneous Hot and Cold Injection is required based on ___(1)____. If conditions are met, hot and cold leg injection must be aligned within the maximum allowable time of ____(2)_____ hours from the start of the event to ensure boric acid precipitation in the core does not occur.

(1) (2) A. HPSI and LPSI flows within acceptable region of the SI flow curves two B. Pressurizer level, Reactor vessel level and RCS subcooling two C. Pressurizer level, Reactor vessel level and RCS subcooling three D. HPSI and LPSI flows within the acceptable region of the SI flow curves three 2015 NRC Exam SRO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 4 of 50 Proposed Answer: C Explanation: (Optional)

A. Incorrect. This step is performed to eliminate boron precipitation in the vessel, the applicant could determine that SI flows are the initiating parameters since they would be affected by boron precipitation. Two hours from the event is plausible because two hours is the minimum time required to align hot and cold leg injection. The two hour requirement ensures that borated water injected to the hot leg is not entrained in the steam being released through the break. B. Incorrect. Part 1 is correct. Two hours from the event is plausible because two hours is the minimum time required to align hot and cold leg injection. The two hour requirement ensures that borated water injected to the hot leg is not entrained in the steam being released through the break. C. CORRECT: Step 50 of OP-902-002 requires the crew to evaluate the need for hot and cold leg injection based on Pressurizer level, Reactor vessel level and RCS subcooling. OP-902-002 Tech guide states that hot and cold injection must be performed within three hours of the event to prevent boric acid precipitation in the core.

D. Incorrect. Step 50 of OP-902-002 requires the crew to evaluate the need for hot and cold leg injection based on Pressurizer level, Reactor vessel level and RCS subcooling. Since this step is performed to eliminate boron precipitation in the vessel, the applicant could determine that SI flows are the initiating parameters. Part 2 is correct.

Technical Reference(s): OP-902-002 step 50 revision 19 (Attach if not previously provided) TGOP-902-002 step 50 revision 18 (including version/revision number)

Proposed references to be provided to applicants during examination:

None Learning Objective: WLP-OPS-PPE02 obj. 17 (As available)

Question Source: Bank #

Modified Bank #

(Note changes or attach parent)

New X Question History: Last NRC Exam None Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 5 Comments:

2015 NRC Exam SRO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 5 of 50 Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 000026 2.2.12 Importance Rating 4.1 K/A Statement 2.2.12: Knowledge of surveillance procedures.

Proposed Question:

SRO 3 Rev: 0 Given: Component Cooling Water (CCW) Pump A has tripped on overcurrent.

The crew will declare CCW Train A inoperable along with the affected systems listed in ___(1)____. The crew is required to demonstrate the operability of the Train B offsite AC circuit by performing surveillance procedure____(2)____ within one hour and at least once per eight hours thereafter.

(1) (2) A. OP-100-014, Technical Specification and Technical Requirements Compliance OP-903-001, Attachment 11.14, Electrical Distribution Operability Check B. OP-100-010, Equipment Out of Service OP-903-001, Attachment 11.14, Electrical Distribution Operability Check C. OP-100-014, Technical Specification and Technical Requirements Compliance OP-903-066, Electrical Breaker Alignment Check D. OP-100-010, Equipment Out of Service OP-903-066, Electrical Breaker Alignment Check

2015 NRC Exam SRO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 6 of 50 Proposed Answer: C Explanation: (Optional)

A. Incorrect. Part 1 is correct. OP-903-001 Attachment 11.14, Electrical Distribution Operability Check is performed once every 7 days to meet the requirements of TS surveillance 4.8.1.1.1a. But, OP-100-014 directs the crew to perform the OP-903-066 surveillance upon TS 3.8.1.1b entry due to a loss of a CCW train. B. Incorrect. OP-100-010, Equipment Out of Service will be performed but will refer the crew OP-100-014 to obtain the affected systems and tech specs to enter upon a loss of a CCW train. OP-903-001 Attachment 11.14, Electrical Distribution Operability Check is performed once every 7 days to meet the requirements of TS surveillance 4.8.1.1.1a. But, OP-100-014 directs the crew to perform the OP-903-066 surveillance upon TS 3.8.1.1b entry due to a loss of a CCW train.

C. CORRECT: OP-100-014, Technical Specification and Technical Requirements Compliance lists the affected systems and tech specs to enter upon a loss of a CCW train (Cascading Tech Specs). W3 uses surveillance procedure OP-903-066, Electrical Breaker Alignment Check to verify surveillance requirements of 4.8.1.1.1a are met upon a loss of a CCW train (CCW train inoperable renders the associated EDG inoperable). D. Incorrect. OP-100-010, Equipment Out of Service will be performed but will refer the crew OP-100-014 to obtain the affected systems and tech specs to enter upon a loss of a CCW train. Part 2 is correct.

Technical Reference(s): OP-100-014 attachment 6.6 revision 328 (Attach if not previously provided) OP-903-066 revision 302 (including version/revision number) TS 3.8.1.1 action a 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 None Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 2 Comments:

2015 NRC Exam SRO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 7 of 50 Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 000038 EA2.15 Importance Rating 4.4 K/A Statement EA2.15: Ability to determine or interpret the following as they apply to a SGTR: Pressure at which to maintain RCS during S/G cooldown.

Proposed Question:

SRO 4 Rev: 0 Given: A plant shutdown in accordance with OP-010-005, Plant Shutdown, is in progress due to primary to secondary leakage The control room has entered OP-901-202, Steam Generator Tube Leakage (SGTL) or High Activity Charging pump AB is tagged out Primary to secondary leakage increases to 95 gpm.

The crew will ___(1)____. After the initial plant cooldown to 520 °F, the CRS will give the order to de-pressurize the RCS to _____(2)______. Note: OP-901-212: Rapid Plant Power Reduction (1) (2) A. transition to OP-901-212, concurrently with OP-901-202, SGTL or High Activity 930 psia to minimize leakage B. trip the reactor and diagnose into OP-902-007, Steam Generator Tube Rupture Recovery 930 psia to minimize leakage C. transition to OP-901-212, concurrently with OP-901-202, SGTL or High Activity 1150 psia to maintain RCP NPSH D. trip the reactor and diagnose into OP-902-007, Steam Generator Tube Rupture Recovery 1150 psia to maintain RCP NPSH

2015 NRC Exam SRO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 8 of 50 Proposed Answer: D Explanation: (Optional)

A. Incorrect. Transitioning to OP-901-212 (Rapid Down Power) is plausible because leakage has increased. If primary to secondary leakage is small and pressurizer level can be maintained the crew is expected to remain in OP-901-202. However, in this case, leakage is in excess of Charging pump capacity and EOP entry is warranted. Part 2 is plausible because the step to de-pressurize the RCS lists several criteria; <930 psia is one of those criteria and the reason to minimize leakage is correct. However, TGOP-902-007 pg. 29 says that the requirement to maintain RCP NPSH takes precedence over the strategy to equalize primary and secondary pressures. B. Incorrect. Part 1 is correct. Maintaining RCP NPSH takes precedence over the strategy to equalize primary and secondary pressures (TGOP-902-007 pg. 29). C. Incorrect. Leakage is in excess of Charging pump capacity and EOP entry is warranted. Part 2 is correct.

D. CORRECT: A leak rate of 95 gpm is in excess of Charging pump capacity. EOP entry would be warranted and OP-902-007 would be diagnosed. Both OP-902-007 and OP-901-202 have similar steps to cooldown and de-pressurize the RCS. No indication is given that the plant is on natural circulation therefore RCP operation must be assumed and depressurizing to 1150 psia would maintain RCP NPSH requirements. Maintaining RCP NPSH takes precedence over the strategy to equalize primary and secondary pressures (TGOP-902-007 pg. 29).

Technical Reference(s): OP-902-007 step 12 (rev16); TGOP-902-007 pg. 29 (rev 306) (Attach if not previously provided) OP-901-202 pg.8 (rev15) (including version/revision number) OI-902-009 Att. 2-A (rev310)

Proposed references to be provided to applicants during examination: OP-902-009, Att. 2-A RCS Press/Temp Limits (non-harsh) (rev 310)

Learning Objective: WLP-OPS-PPE07 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 55.43 5 Comments:

2015 NRC Exam SRO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 9 of 50 Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 000057 AA2.19 Importance Rating 3.1 K/A Statement AA2.19: Ability to determine and interpret the following as they apply to the Loss of Vital AC Instrument Bus: The plant automatic actions that will occur on the loss of a vital ac electrical instrument bus.

Proposed Question:

SRO 5 Rev: 0 The plant is operating at 100% power when the SUPS power supply to PDP-90A is lost.

The crew will enter section ___(1)____ of OP-901-312, Loss of Vital Instrument Bus. The CRS will direct the crew to ______(2)_____. (1) (2) A. E8 Loss of SUPS A monitor primary plant parameters on Safety Channels B, C, and D B. E1 Loss of SUPS MA verify Component Cooling Water supplying the AB loop from the B train of CCW C. E8 Loss of SUPS A verify Component Cooling Water supplying the AB loop from the B train of CCW D. E1 Loss of SUPS MA monitor primary plant parameters on Safety Channels B, C, and D 2015 NRC Exam SRO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 10 of 50 Proposed Answer: C Explanation: (Optional)

A. Incorrect. Part 1 is correct. Monitoring primary plant parameters on Safety Channels B, C, and D are the first step to be performed on a loss of SUPS MA. B. Incorrect. PDP-90A is fed from SUPS A. Part 2 is correct.

C. CORRECT: PDP-90A is fed from SUPS A. All components fed for PDP-90A are lost when SUPS A fails. CC-200A and CC-727 are powered from PDP-90A and will go closed on a loss of power. OP-901-312, E8 Loss of SUPS A, directs the crew to verify Component Cooling Water supplying the AB loop from the B side. D. Incorrect. PDP-90A is fed from SUPS A. Monitoring primary plant parameters on Safety Channels B, C, and D are the first step to be performed on a loss of SUPS MA. Technical Reference(s): OP-901-312 section E8, 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-PPO30 obj. 1 and 3 (As available)

Question Source: Bank #

Modified Bank #

(Note changes or attach parent)

New X Question History: Last NRC Exam None Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 b5 Comments:

2015 NRC Exam SRO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 11 of 50 Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 000065 AA2.07 Importance Rating 3.2 K/A Statement AA2.07: Ability to determine and interpret the following as they apply to the Loss of Instrument Air: Whether backup nitrogen supply is controlling valve position.

Proposed Question:

SRO 6 Rev: 0 Given: The plant experienced an uncomplicated reactor trip. The crew transitioned to OP-010-005, Plant Shutdown, and is holding mode 3 conditions. An unisolable Instrument Air leak occurred when the plant tripped. Instrument Air pressure has been lost for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> due to the leak and leak repair should be complete in 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />. Reactor Coolant Pumps 1B and 2B are operating. The following annunciators are illuminated on CP-2 o RCP 1A CCW FLOW LO o RCP 2A CCW FLOW LO o RCP 1B CCW FLOW LO o RCP 2B CCW FLOW LO The following annunciators are illuminated on CP-18:

o RCP 1A CCW FLOW LOST o RCP 2A CCW FLOW LOST The CRS will direct the crew to ___(1)____ in accordance with the guidance in _________(2)_________.

(1) (2) A. trip Reactor Coolant Pumps 1B and 2B OP-901-511, Loss of Instrument Air B. gag open CC-200A and CC-727, CCW Suct & Discharge Header Tie Valves A to AB OP-901-511, Loss of Instrument Air C. gag open CC-200A and CC-727, CCW Suct & Discharge Header Tie Valves A to AB OP-901-510, Component Cooling Water Malfunction D. trip Reactor Coolant Pumps 1B and 2B OP-901-510, Component Cooling Water Malfunction

2015 NRC Exam SRO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 12 of 50 Proposed Answer: B Explanation: (Optional)

A. Incorrect. Plausible because this is the desired response if all CCW is lost to the RCPs. The applicant could determine that CC-200A and 727 are closed and there is not enough time to gag open CC-200A and CC-727 within 3 minutes. Part 2 is correct. B. CORRECT: Per step 18 of OP-901-511, Loss of instrument Air: If the loss of instrument air is expected to last longer than 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />, then actions must be taken within 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />. The applicant will determine that since the flow lost annunciator is clear to the RCPs that are running, CCW flow must still exist. By procedure, the CRS will direct gagging open CC-200A and CC-727 before they fail closed. C. Incorrect. Part 1 is correct. OP-901-510 deals with CCW malfunctions but does not provide guidance for a loss of instrument air and nitrogen accumulators to CCW valves.

D. Incorrect. Plausible because this is the desired response if all CCW is lost to the RCPs. The applicant could determine that CC-200A and CC-727 are closed and there is not enough time to gag them open within 3 minutes. The guidance to trip the Reactor Coolant Pumps on a loss of CCW is located in OP-901-510 but is the wrong response.

Technical Reference(s): OP-901-511 step 18 revision 14 (Attach if not previously provided) OP-500-008 att. 4.75 revision 26 (including version/revision number) OP-500-013 att. 4.1 revision 20 Proposed references to be provided to applicants during examination:

None Learning Objective: WLP-OP-PPO50 obj. 3 (As available)

Question Source: Bank #

Modified Bank #

(Note changes or attach parent)

New X Question History: Last NRC Exam None Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 b5 Comments:

2015 NRC Exam SRO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 13 of 50 Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 2 K/A # 000003 AA2.01 Importance Rating 3.9 K/A Statement AA2.01: Ability to determine and interpret the following as they apply to the Dropped Control Rod: Rod position indication to actual rod position.

Proposed Question:

SRO 7 Rev: 0 Given: Power is 100% All CEAs fully withdrawn Five minutes later, the ATC reports the following: CEA 38 indicates 135 inches on the CP-2 CEAC CRT CEA 38 indicates 150 inches on the PMC Both the Upper and Lower Electrical Limit lights for CEA 38 are extinguished All Rod Bottom lights are extinguished Tcold is 541 ºF and slowly dropping To verify CEA position, the CRS will request indication on the ____(1)____ . The CRS will enter OP-901-102, CEA or CEDMCS Malfunction, sub-section _____(2)_____. (1) (2) A. CP-2 digital meter for CEA 38 E1, CEA Misalignment Greater than 7 inches B. CPC that CEA 38 is targeted to E5, CEA Position Indication Malfunction C. CPC that CEA 38 is targeted to E1, CEA Misalignment Greater than 7 inches D. CP-2 digital meter for CEA 38 E5, CEA Position Indication Malfunction

2015 NRC Exam SRO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 14 of 50 Proposed Answer: C Explanation: (Optional)

A. Incorrect: The CP-2 digital meter is fed by the same input as the PMC (pulse counter). In this case (slipped rod) the pulse counter will not see the change in rod position since the rod had no motion demand and it did not drop to the bottom of the core. Part 2 is correct. B. Incorrect: Part 1 is correct. PMC indication although erroneous would be expected to read 150 inches and therefore not an indication malfunction.

C. CORRECT: CPC indication is fed from safety grade RSPT indication which provides actual rod position. All indications reported by the ATC with the exception of the PMC indication are consistent with a rod misalignment. D. Incorrect: The CP-2 digital meter is fed by the same input as the PMC (pulse counter). In this case (slipped rod) the pulse counter will not see the change in rod position since the rod had no motion demand and it did not drop to the bottom of the core. PMC indication although erroneous would be expected to read 150 inches and therefore not an indication malfunction.

Technical Reference(s): SD-CPC (rev16) fig.4; SD-CED (rev 11) pg.19 (Attach if not previously provided) OP-901-102 (rev302) (including version/revision number)

Proposed references to be provided to applicants during examination:

None Learning Objective: WLP-OPS-PPO10 obj.1 (As available)

Question Source: Bank #

Modified Bank #

(Note changes or attach parent)

New X Question History: Last NRC Exam none Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 5 Comments:

2015 NRC Exam SRO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 15 of 50 Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 2 K/A # 000024 2.2.22 Importance Rating 4.7 K/A Statement 2.2.22: Knowledge of limiting conditions for operations and safety limits.

Proposed Question:

SRO 8 Rev: 0 Given: Plant is performing a cooldown for a refueling outage RCS temperature is 195 ºF RCS pressure is 350 psia EDG A is tagged out Offsite power train B is declared inoperable because of a voltage problem.

The CRS will direct performing surveillance procedure OP-903-002, Boration Flow Path Valve Lineup Verification, Attachment ____(1)____. The boration flow path will use _____(2)_____ as the operable power source.

(1) (2) A. 10.6 Boration from RWSP and HPSI pump A offsite train A B. 10.7 Boration from RWSP and HPSI pump B EDG B C. 10.4, Boration from BAM pump A and Charging pump A offsite train A D. 10.4, Boration from BAM pump B and Charging pump B EDG B 2015 NRC Exam SRO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 16 of 50 Proposed Answer: B Explanation: (Optional)

A. Incorrect: Even though HPSI pump A is operable and available to be used as a boration flow path, offsite train A cannot be credited as the operable power source. B. CORRECT: OP-903-002 is the surveillance procedure used to verify SR 4.1.2.1. TS 3.1.2.1 bases specifies that the operable power source must be an EDG. HPSI pump B is the only component that could be powered from EDG B. TS bases knowledge is required to determine the correct boration flow path. C. Incorrect: BAM pump A and Charging pump A are operable and available but offsite train A cannot be credited as the operable power source. D. Incorrect: BAM pump B is powered from safety bus A and therefore EDG B cannot be credited as the operable power source.

Technical Reference(s): OP-903-002 (rev303) (Attach if not previously provided) TS 3.1.2.1 & bases (including version/revision number)

Proposed references to be provided to applicants during examination:

None Learning Objective: WLP-OPS-CVC00 obj.8 (rev23) (As available)

Question Source: Bank #

Modified Bank #

(Note changes or attach parent)

New X Question History: Last NRC Exam none Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 2 Comments:

2015 NRC Exam SRO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 17 of 50 Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 2 K/A # 000051 AA2.02 Importance Rating 4.1 K/A Statement AA2.02: 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:

SRO 9 Rev: 0 Given: The plant is at 90% and performing a power reduction from 100% due to lowering condenser vacuum Per OP-901-220, Loss of Condenser Vacuum, the reactor will be tripped if condenser vacuum is approaching ___(1)____ Inches Hg. The crew will _________(2)_________.

(1) (2) A. 14 perform OP-902-000, Standard Post Trip Actions, concurrently with OP-901-220 B. 20 perform OP-902-000, Standard Post Trip Actions, concurrently with OP-901-220 C. 14 go to OP-902-000, Standard Post Trip Actions, exit OP-901-220 D. 20 go to OP-902-000, Standard Post Trip Actions, exit OP-901-220 2015 NRC Exam SRO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 18 of 50 Proposed Answer: B Explanation: (Optional)

A. Incorrect. 14" HG Vacuum requires taking action to protect the main condenser by isolating steam inputs to the condenser. This is done after the reactor is tripped which should occur when approaching 20" Hg Vacuum. Correct procedure implementation strategy. B. CORRECT: The Main Turbine trip setpoint for lo vacuum is 20" Hg Vacuum. OP-901-220 requires tripping the reactor at this point to prevent intentionally putting the primary through a loss of load transient which would result in potentially quick opening all 6 Steam Bypass Valves and directing high energy steam into the condenser when vacuum is already challenged. In most cases, the offnormal procedures are exited or suspended while performing OP-902-000. OP-901-220 specifically states to implement the procedure concurrently with OP-902-000. C. Incorrect. 14" HG Vacuum requires taking action to protect the main condenser by isolating steam inputs to the condenser. This is done after the reactor is tripped which should occur when approaching 20" Hg Vacuum. Incorrect procedure implementation strategy. D. Incorrect. Correct action point. Incorrect procedure implementation strategy.

Technical Reference(s): OP-901-220 pages 7 and 8 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-PPO20 obj. 3 (As available)

Question Source: Bank #

X Question 9 Modified Bank #

(Note changes or attach parent)

New Question History: Last NRC Exam 2011 NRC SRO Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 5 Comments:

2015 NRC Exam SRO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 19 of 50 Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 2 K/A # CE/A11 2.1.7 Importance Rating 4.7 K/A Statement 2.1.7: Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation.

Proposed Question: SRO 10 Rev: 0 Given: Plant was operating at 100% power with EDG A danger tagged An Excess Steam Demand occurred SIAS, CIAS, and MSIS have been initiated Pressurizer level is 0% After entering OP-902-004, Excess Steam Demand Recovery, the following conditions change: Representative CET temperature and RCS pressure start to rise. A Loss of Off Site Power occurs All components respond as designed to the event.

Based on these conditions, the CRS will ____(1)____ and ____(2)____ (1) (2) A. remain in OP-902-004 direct the ATC to stabilize RCS pressure above HPSI Pump shutoff head.

B. remain in OP-902-004 direct the ATC to stabilize RCS pressure below HPSI Pump shutoff head.

C. exit OP-902-004 and enter OP-902-008, Functional Recovery direct the ATC to stabilize RCS pressure above HPSI Pump shutoff head.

D. exit OP-902-004 and enter OP-902-008, Functional Recovery direct the ATC to stabilize RCS pressure below HPSI Pump shutoff head.

2015 NRC Exam SRO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 20 of 50 Proposed Answer: A Explanation: (Optional)

A. CORRECT: Required to stay in OP-902-004 to address stabilization of RCS pressure ABOVE the HPSI pump shutoff head. The safety function status checklist in OP-902-004 is still being met with one safety bus powered. B. INCORRECT: Required to stay in OP-902-004 to address stabilization of RCS pressure ABOVE the HPSI pump shutoff head. C. INCORRECT: OP-902-004 address the required actions. The safety function status checklist in OP-902-004 is still being met with one safety bus powered. The step to stabilize RCS pressure is contained in OP-902-008, but entry into OP-902-008 is not required. D. INCORRECT: OP-902-004 address the required actions. The safety function status checklist in OP-902-004 is still being met with one safety bus powered. The step to stabilize RCS pressure is contained in OP-902-008, but entry into OP-902-008 is not required.

Technical Reference(s): OP-902-004 pages 17, 37 revision 15 (Attach if not previously provided) (including version/revision number)

Proposed references to be provided to applicants during examination:

None Learning Objective: WLP-OPS-PPE04 obj. 7 (As available)

Question Source: Bank #

X Question #10 Modified Bank #

(Note changes or attach parent)

New Question History: Last NRC Exam 2014 SRO Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 5 Comments:

2015 NRC Exam SRO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 21 of 50 Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 004 2.2.25 Importance Rating 4.2 K/A Statement 2.2.25: Knowledge of the bases in Technical Specifications for limiting conditions for operations and safety limits.

Proposed Question: SRO 11 Rev: 0 As stated in TS surveillance requirement 4.1.2.8, each borated water source shall be demonstrated operable by verifying the Boric Acid Makeup Tank solution temperature is greater than or equal to 60°F when reactor auxiliary building air temperature is less than ______(1)_______ °F. The basis for this minimum temperature is to ensure that ______(2)______. (1) (2) A. 60 assumptions used in the MSLB return to power event remain valid B. 60 boron will not precipitate out of solution C. 55 boron will not precipitate out of solution D. 55 assumptions used in the MSLB return to power event remain valid

2015 NRC Exam SRO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 22 of 50 Proposed Answer: C Explanation: (Optional)

A. Incorrect: Per TS 3.1.2.8, BAM Tank solution temperature is verified greater than or equal to 60°F when RAB temperature is less than 55°F (not 60°F). It is plausible to assume the temperatures may be the same. The minimum temperature for the CSP is verified in the same way the BAMT minimum temperature is verified. The temperature limits are also the same. This distractor is plausible because the CSP basis is to ensure the assumptions used in the MSLB return to power event remain valid (TS 3.7.1.3). B. Incorrect. Per TS 3.1.2.8, BAM Tank solution temperature is verified greater than or equal to 60°F when RAB temperature reaches 55°F (not 60°F). Part 2 is correct. C. CORRECT: Per TS 3.1.2.8, BAM Tank solution temperature is verified greater than or equal to 60°F when RAB temperature is less than 55°F. TS 3/4.1.2 basis states that the minimum BAM Tank temperature is to ensure that boron will not precipitate.

D. Incorrect. Part 1 is correct. The minimum temperature for the CSP is verified in the same way the BAMT minimum temperature is verified. The temperature limits are also the same. This distractor is plausible because the CSP basis is to ensure the assumptions used in the MSLB return to power event remain valid (TS 3.7.1.3).

Technical Reference(s): TS 3/4.1.2 basis (Attach if not previously provided) TS 3.7.1.3 basis (including version/revision number) TS 4.1.2.8 Proposed references to be provided to applicants during examination:

None Learning Objective: WLP-OPS-CVC00 obj. 7 (As available)

Question Source: Bank #

Modified Bank #

(Note changes or attach parent)

New X Question History: Last NRC Exam None Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 b2 Comments:

2015 NRC Exam SRO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 23 of 50 Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 005 A2.02 Importance Rating 3.7 K/A Statement A2.02: Ability to (a) predict the impacts of the following malfunctions or operations on the RHRS, and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Pressure transient protection during cold shutdown.

Proposed Question: SRO 12 Rev: 0 Given: RCS temperature is 180°F and steady RCS pressure is 320 PSIA and steady SDC Train B is in service SDC Train A is secured The following annunciators are received:

SIAS Train A Logic Initiated (Cabinet K, G-19) SIAS Train B Logic Initiated (Cabinet K, G-20) LOOP 1 SDC RELIEF VLV ACTIVE (Cabinet M, A-7)

The addition of borated water ____(1)____ the capacity of SI-406B, RC Loop 1 SDC Suction LTOP relief to CNTMT Sump. The CRS will secure High Pressure Safety Injection Pumps in accordance with the guidance in ____(2)____. (1) (2) A. is within OP-901-131, Shutdown Cooling Malfunction B. exceeds OP-901-131, Shutdown Cooling Malfunction C. exceeds OP-901-504, Inadvertent ESFAS Actuation D. is within OP-901-504, Inadvertent ESFAS Actuation

2015 NRC Exam SRO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 24 of 50 Proposed Answer: D Explanation: (Optional)

A. Incorrect. Part 1 is correct. OP-901-131, SDC malfunction may be entered but the guidance to secure HPSI pumps is located in OP-901-504. RCS pressure and temperature are steady and there are no containment water leakage high alarms, the applicant will assume there is no RCS leak. B. Incorrect. The basis (TS 3.4.8.3) for the LTOP relief valves states that for conditions where the RCS cold legs are less than 200°F, the capacity of one LTOP exceeds the transient of the simultaneous start of all three HPSI pumps, three charging pumps and all pressurizer heaters. In this case, SDC temperature is less than 200°F. OP-901-131, SDC malfunction may be entered but the guidance to secure HPSI pumps is located in OP-901-504. C. Incorrect. The basis (TS 3.4.8.3) for the LTOP relief valves states that for conditions where the RCS cold legs are less than 200°F, the capacity of one LTOP exceeds the transient of the simultaneous start of all three HPSI pumps, three charging pumps and all pressurizer heaters. In this case, SDC temperature is less than 200°F. Part 2 is correct.

D. CORRECT: The basis (TS 3.4.8.3) for the LTOP relief valves states that for conditions where the RCS cold legs are less than 200°F, the capacity of one LTOP exceeds the transient of the simultaneous start of all three HPSI pumps, three charging pumps and all pressurizer heaters. The guidance to secure HPSI pumps is located in OP-901-504, Inadvertent ESFAS Acuation. RCS pressure and temperature are steady and there are no containment water leakage high alarms, the applicant will assume there is no RCS leak.

Technical Reference(s): TS 3.4.8.3 basis (Attach if not previously provided) OP-901-504 section E1 revision 9 (including version/revision number)

Proposed references to be provided to applicants during examination:

None Learning Objective: WLP-OPS-SDC00 obj. 1 (As available)

WLP-OPS-PPO50 obj.1 Question Source: Bank #

Modified Bank #

(Note changes or attach parent)

New X Question History: Last NRC Exam None Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 5 Comments:

2015 NRC Exam SRO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 25 of 50 Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 010 2.2.42 Importance Rating 4.6 K/A Statement 2.2.42: Ability to recognize system parameters that are entry-level conditions for Technical Specifications.

Proposed Question: SRO 13 Rev: 0 A turbine control system malfunction has caused the turbine to runback from 100% power. The following indications are present two minutes after the turbine stops running back: REACTOR COOLANT Tave-Tref HI alarm in SELECTED COLD LEG 1 TEMPERATURE HI alarm in PRESSURIZER PRESSURE HI/LO alarm in Tcold is 547°F PZR Pressure is 2277 psia All PZR Heaters have de-energized automatically Pressurizer pressure controller output is 0% Pressurizer spray valve controller output is 0%

The CRS will address pressurizer pressure control by entering OP-901-120, Pressurizer Pressure Control Malfunction, subsection ____(1)____ and enter the Tech Spec(s) for _____(2)_____. (1) (2) A. E2, Pressurizer Pressure Controller Malfunction pressurizer pressure (TS 3.2.8) only B. E3, Pressurizer Spray Valve Malfunction pressurizer pressure (TS 3.2.8) only C. E2, Pressurizer Pressure Controller Malfunction pressurizer pressure (TS 3.2.8) and reactor coolant cold leg temperature (TS 3.2.6)

D. E3, Pressurizer Spray Valve Malfunction pressurizer pressure (TS 3.2.8) and reactor coolant cold leg temperature (TS 3.2.6)

2015 NRC Exam SRO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 26 of 50 Proposed Answer: A Explanation: (Optional)

A. CORRECT: PZR pressure controller output should be high (i.e. >75%) for the given conditions. The applicant must also recognize that a 0% output from the pressure controller will input into the spray valve controller and prevent the spray controller from responding in automatic. Tcold temperature is within the TS LCO band of 536-549°F. Only pressurizer pressure is exceeding the TS LCO band. B. Incorrect: The spray valve controller will not respond in automatic because it is receiving a 0% signal from the master controller. The CRS will enter E2 which direct actions to place both the PZR pressure and spray valve controllers in manual and control pressure manually. Part 2 is correct. C. Incorrect: Part 1 is correct. Tcold temperature is within the TS LCO band of 536-549°F. Only pressurizer pressure is exceeding the TS LCO band. D. Incorrect: The spray valve controller will not respond in automatic because it is receiving a 0% signal from the master controller. The CRS will enter E2 which direct actions to place both the PZR pressure and spray valve controllers in manual and control pressure manually. Tcold temperature is within the TS LCO band of 536-549°F. Only pressurizer pressure is exceeding the TS LCO band.

Technical Reference(s): OP-901-120 (rev302) (Attach if not previously provided) TS 3.2.6 & 3.2.8 (including version/revision number) SD-PLC pp. 30,33 (rev9)

Proposed references to be provided to applicants during examination:

None Learning Objective: WLP-OPS-PLC00 obj.5 (As available)

Question Source: Bank #

Modified Bank #

(Note changes or attach parent)

New X Question History: Last NRC Exam none Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 5 Comments:

2015 NRC Exam SRO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 27 of 50 Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 022 2.4.21 Importance Rating 4.6 K/A Statement 2.4.21: Knowledge of the parameters and logic used to assess the status of safety functions, such as reactivity control, core cooling and heat removal, reactor coolant system integrity, containment conditions, radioactivity release control, Proposed Question: SRO 14 Rev: 0 Given: An Excess Steam Demand occurred SIAS, CIAS, MSIS and CSAS have been initiated After entering OP-902-004, Excess Steam Demand Recovery, the BOP reports that CFC A has tripped on overcurrent The CRS will determine that the ____(1)____ safety function is not met. The crew will _____(2)_____. (1) (2) A. containment temperature and pressure control go to OP-902-008, Functional Recovery procedure B. containment isolation go to OP-902-008, Functional Recovery procedure C. containment isolation remain in OP-902-004, Excess Steam Demand Recovery procedure D. containment temperature and pressure control remain in OP-902-004, Excess Steam Demand Recovery procedure

2015 NRC Exam SRO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 28 of 50 Proposed Answer: C Explanation: (Optional)

A. Incorrect. The containment temperature and pressure control safety function is met due to > 1 CFC in operation and Containment Spray running. The guidance for closing the CFC CCW Isolation Valves is located in OP-902-004 as a contingency step. Therefore, the crew does not have to exit to OP-902-008 even though the Containment Isolation Safety Function is not being met. B. Incorrect. Part 1 is correct. The guidance for closing the CFC CCW Isolation Valves is located in OP-902-004 as a contingency step. Therefore, the crew does not have to exit to OP-902-008 even though the Containment Isolation Safety Function is not being met. C. CORRECT: If a SIAS/CIAS is initiated, the crew is to check all CFCs running. If any CFC is not operating, the crew is no longer meeting the Containment Isolation Safety Function until the CFC CCW isolation valves are overridden closed. The guidance for closing the CFC CCW Isolation Valves is located in OP-902-004 as a contingency step. Therefore, the crew does not have to exit to OP-902-008 even though the Containment Isolation Safety Function is not being met.

D. Incorrect. The containment temperature and pressure control safety function is met due to > 1 CFC in operation and Containment Spray running. Part 2 is correct.

Technical Reference(s): OP-902-004 step 25 and SFSC (Attach if not previously provided) for CI and CTPC (including version/revision number) TGOP-902-004 step 25 Proposed references to be provided to applicants during examination:

None Learning Objective: WLP-OPS-PPE04 obj. 7 (As available)

Question Source: Bank #

Modified Bank #

(Note changes or attach parent)

New X Question History: Last NRC Exam None Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 5 Comments:

2015 NRC Exam SRO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 29 of 50 Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 076 A2.02 Importance Rating 3.1 K/A Statement A2.02: Ability to (a) predict the impacts of the following malfunctions or operations on the SWS and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Service water header pressure.

Proposed Question: SRO 15 Rev: 0 Given: Plant is at 100% power Turbine Cooling Water (TCW) Pump B is in service Turbine Cooling Water (TCW) Pump A is tagged for repair Turbine Cooling Water Pump B trips and the crew enters OP-901-512, Loss of Turbine Cooling Water Pumps.

The most limiting component for this event is the ____(1)____. The crew will perform OP-902-000, Standard Post Trip Actions, and then diagnose to _____(2)_____ Recovery Procedure.

(1) (2) A. main turbine bearings OP-902-001, Reactor Trip B. main generator OP-902-006, Loss of Main Feedwater C. main generator OP-902-001, Reactor Trip D. main turbine bearings OP-902-006, Loss of Main Feedwater

2015 NRC Exam SRO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 30 of 50 Proposed Answer: B Explanation: (Optional)

A. Incorrect. Main Turbine bearings are affected but are not the limiting component. OP-902-001, Reactor Trip recovery procedure would be diagnosed to in this event if Main Feed Pumps were not tripped due to closing MSIVs because condenser vacuum is broken. B. CORRECT: The caution before the steps for tripping the reactor in section E1 of OP-901-512 states that main turbine damage will occur in 2-3 minutes with the generator as the most limiting component. The crew will diagnose to OP-902-006, Loss of Main Feedwater, because both Main Feed Pumps will trip when the MSIVs are closed. MSIVs are closed and condenser vacuum broken to stop the turbine quicker. These steps are IAW OP-901-512. C. Incorrect. Part 1 is correct. OP-902-001, Reactor Trip recovery procedure would be diagnosed to in this event if Main Feed Pumps were not tripped due to closing MSIVs because condenser vacuum is broken. D. Incorrect. Main Turbine bearings are affected but are not the limiting component. Part 2 is correct.

Technical Reference(s): OP-901-512 section E1 revision 3 (Attach if not previously provided) OP-902-006 revision 16 (including version/revision number)

Proposed references to be provided to applicants during examination:

None Learning Objective: WLP-OPS-TC00 obj. 8 (As available)

Question Source: Bank #

Modified Bank #

(Note changes or attach parent)

New X Question History: Last NRC Exam None Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 5 Comments:

2015 NRC Exam SRO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 31 of 50 Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 2 K/A # 011 A2.01 Importance Rating 3.1 K/A Statement A2.01: Ability to (a) predict the impacts of the following malfunctions or operations on the PZR LCS; and (b) based on those predictions, use Procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Excessive letdown.

Proposed Question: SRO 16 Rev: 0 Given: The plant is at 100% power. Letdown flow is 50 gpm and rising. PZR level indicator RC-ILI-110X indicates 55% and lowering. PZR level indicator RC-ILI-110Y indicates 54% and lowering. The ATC reports that the pressurizer level setpoint on the Pressurizer Level Controller (RC-ILIC-0110) is 55.6%. The output of the Pressurizer Level Controller (RC-ILIC-0110) is 90%.

The CRS will enter ____(1)____ and direct the ATC to take manual control of the ____(2)____ to restore Pressurizer level.

(1) (2) A. OP-901-110, Pressurizer Level Control Malfunction Pressurizer Level Controller (RC-ILIC-0110)

B. OP-901-112, Charging and Letdown Malfunction Pressurizer Level Controller (RC-ILIC-0110)

C. OP-901-110, Pressurizer Level Control Malfunction Letdown Flow Control Valves controller (RC-IHIC-0110)

D. OP-901-112, Charging and Letdown Malfunction Letdown Flow Control Valves controller (RC-IHIC-0110)

2015 NRC Exam SRO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 32 of 50 Proposed Answer: A Explanation: (Optional)

A. CORRECT: The applicant will determine that a problem exists with the Pressurizer Level Controller (RC-ILIC-0110) because the control inputs to the PLCS are reading normal for the condition (PZR level setpoint and PZR level control channels). The PZR level controller should be reading minimum (less letdown). Section E3 of OP-901-110 provides the guidance for a pressurizer level controller malfunction. B. Incorrect. Section E2 of OP-901-112 provides guidance for a letdown malfunction for everything downstream of the Pressurizer Level Controller (RC-ILIC-0110). The general actions of OP-901-112 will direct the crew to OP-901-110 if the malfunction is due to the Pressurizer level control system. Section E2 of OP-901-112 directs the crew to take manual control of Letdown Flow Control Valves controller (RC-IHIC-0110). C. Incorrect. Section E2 of OP-901-112 provides guidance for a letdown malfunction for everything downstream of the Pressurizer Level Controller (RC-ILIC-0110). The general actions of OP-901-112 will direct the crew to OP-901-110 if the malfunction is due to the Pressurizer level control system. Section E2 of OP-901-112 directs the crew to take manual control of Letdown Flow Control Valves controller (RC-IHIC-0110).

D. Incorrect. Section E2 of OP-901-112 provides guidance for a letdown malfunction for everything downstream of the Pressurizer Level Controller (RC-ILIC-0110). The general actions of OP-901-112 will direct the crew to OP-901-110 if the malfunction is due to the Pressurizer level control system. Section E2 of OP-901-112 directs the crew to take manual control of Letdown Flow Control Valves controller (RC-IHIC-0110).

Technical Reference(s): OP-901-110 section E0 and E3 revision 8 (Attach if not previously provided) OP-901-112 section E0 revision 6 (including version/revision number)

Proposed references to be provided to applicants during examination:

None Learning Objective: WLP-OPS-PPO10 obj. 3 (As available)

Question Source: Bank #

Modified Bank #

(Note changes or attach parent)

New X Question History: Last NRC Exam None Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 5 Comments:

2015 NRC Exam SRO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 33 of 50 Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 2 K/A # 045 2.4.31 Importance Rating 4.1 K/A Statement 2.4.31: Knowledge of annunciator alarms, indications, or response procedures.

Proposed Question: SRO 17 Rev: 0 Given: The Turbine Building watch reports that "Generator Hydrogen Press. Low" annunciator is locked in at the H2 Control Panel Main Generator Hydrogen pressure is 29 psig and the crew has tripped the Main Turbine The crew will enter OP-901-101, Reactor Power Cutback and ____(1)____ concurrently. To ensure no hot work or open flames are occurring in the Turbine Building due to the possible explosive mixture from Hydrogen Leakage the CRS will first _____(2)_____. (1) (2) A. OP-901-210, Turbine Trip contact the Work Management Center B. OP-901-210, Turbine Trip make a plant page C. OP-901-211, Generator Malfunction make a plant page D. OP-901-211, Generator Malfunction contact the Work Management Center

2015 NRC Exam SRO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 34 of 50 Proposed Answer: C Explanation: (Optional)

A. Incorrect. OP-901-210 provides no guidance for responding to a hydrogen gas leak in the main generator. The first step in the Excessive Generator Hydrogen Gas Leakage section of the offnormal will direct the crew to make a plant page to secure hot work or open flames in the TGB. The reason for this action is indicated in the Warning prior to step 1. B. Incorrect. OP-901-210 provides no guidance for responding to a hydrogen gas leak in the main generator. The offnormal is plausible since the main turbine was manually tripped. Part 2 is correct.

C. CORRECT: (W3 OE, CR-WF3-2008-1165) OP-901-211 provides the guidance for responding to lowering hydrogen pressure in the Main Generator. The crew is required to trip the turbine at 34 psig in the Main Generator. The first step in the Excessive Generator Hydrogen Gas Leakage section of the offnormal will direct the crew to make a plant page to secure hot work or open flames in the TGB. The reason for this action is indicated in the Warning prior to step 1. D. Incorrect. Part 1 is correct. Contacting the WMC is plausible since this is the group that would identify what work that is presently in progress in the Turbine Building. Although, the procedure directs a plant page which would be more immediate.

Technical Reference(s): OP-901-211 section E6 revision 8 (Attach if not previously provided) (including version/revision number)

Proposed references to be provided to applicants during examination:

None Learning Objective: WLP-OPS-PPO50 obj. 3 (As available)

Question Source: Bank #

Modified Bank #

(Note changes or attach parent)

New X Question History: Last NRC Exam None Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 5 Comments:

2015 NRC Exam SRO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 35 of 50 Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 2 K/A # 033 A2.02 Importance Rating 3.0 K/A Statement A2.02: 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: Loss of SFPCS.

Proposed Question: SRO 18 Rev: 0 Given: Spent Fuel Pooling Cooling Pump A is tagged out. A loss of the 3B safety bus has occurred.

The CRS will direct the crew to enter ___(1)___To restore Spent Fuel Pool Cooling, the crew will start Spent Fuel Pool (SFP) Cooling Pump B after the sequencer has timed out by ___(2)___. (1) (2) A. OP-901-311, Loss of Train B Safety Bus only manually energizing the SFP Cooling Pump at the 314B bus then taking the C/S to start B. OP-901-311, Loss of Train B Safety Bus only taking the C/S to start only C. OP-901-311, Loss of Train B Safety Bus and OP-901-513, Spent Fuel Pool Cooling Malfunction concurrently taking the C/S to start only D. OP-901-311, Loss of Train B Safety Bus and OP-901-513, Spent Fuel Pool Cooling Malfunction concurrently manually energizing the SFP Cooling Pump at the 314B bus then taking the C/S to start

2015 NRC Exam SRO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 36 of 50 Proposed Answer: D Explanation: (Optional)

A. Incorrect. The SFP cooling pump is located on the non-safety side of the 314B bus thus power is not restored to the breaker when the EDG energizes the bus. The applicant may assume the SFP cooling pump is powered from the safety side of MCC-314B or the breaker manipulations required are located in OP-901-311, which they are not. Part 2 is correct. B. Incorrect. The crew is required to energize the SFP Cooling Pump at the 314B bus. The SFP cooling pump is located on the non-safety side of the 314B bus thus power is not restored to the breaker when the EDG energizes the bus. OP-901-311 step 8 directs the crew to perform OP-901-513 concurrently to energize SFP cooling pump B. The applicant could assume that the SFP cooling pump is powered from the safety side of MCC-314B and will be energized on EDG B sequencer which would require only taking Fuel Pool Pump B C/S or the breaker manipulations required are located in OP-901-311, which they are not. C. Incorrect. Part 1 is correct. The SFP cooling pump is located on the non-safety side of the 314B bus, therefore power is not restored to the breaker when the EDG energizes the bus. The applicant could assume that the SFP cooling pump is powered from the safety side of MCC-314B and will be energized on EDG B sequencer which would require only taking Fuel Pool Pump B C/S to start.

D. CORRECT: The crew is required to energize the SFP Cooling Pump locally at the 314B bus. The SFP cooling pump is located on the non-safety side of the 314B bus, therefore, power is not restored to the breaker when the EDG energizes the bus. OP-901-311 step 8 directs the crew perform OP-901-513, Spent Fuel Pool Cooling Malfunction concurrently to energize SFP cooling pump B.

Technical Reference(s): OP-901-513 revision 15 (Attach if not previously provided) OP-901-311 step 8 revision 309 (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 #

Modified Bank #

(Note changes or attach parent)

New X Question History: Last NRC Exam None Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 5 Comments:

2015 NRC Exam SRO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 37 of 50 Examination Outline Cross-

Reference:

Level RO SRO Tier # 3 Group # 1 K/A # 2.1.5 Importance Rating 3.9 K/A Statement 2.1.5 Ability to use procedures related to shift staffing, such as minimum crew complement, overtime limitations, etc Proposed Question: SRO 19 Rev: 0 Per EN-OP-115, Conduct of Operations, a minimum of ___(1)___ members of the site fire brigade and a minimum of ___(2)___ emergency communicator(s) is required to fulfill minimum shift staffing requirements.

(1) (2) A. four one B. four two C. five one D. five two 2015 NRC Exam SRO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 38 of 50 Proposed Answer: C Explanation: (Optional)

A. Incorrect. Four fire brigade members is plausible because W3 once used the CRS as the fire brigade leader and four other fire brigade members was required, now only the auxiliary operators stand fire brigade. Part 2 is correct. B. Incorrect. Four fire brigade members is plausible because W3 once used the CRS as the fire brigade leader and four other fire brigade members was required, now only the auxiliary operators stand fire brigade. Two emergency communicators is plausible because two switchgear operators is required per the same attachment in EN-OP-115.

C. CORRECT: EN-OP-115, Conduct of Operations Attachment 9.10, a minimum of five members of the site fire brigade and a minimum of one emergency communicator is required to fulfill minimum shift staffing requirements.

D. Incorrect. Part 1 is correct. Two emergency communicators is plausible because two switchgear operators is required per the same attachment in EN-OP-115.

Technical Reference(s): EN-OP-115 attachment 9.10 revision 15 (Attach if not previously provided) (including version/revision number)

Proposed references to be provided to applicants during examination:

None Learning Objective: WLP-OPS-TS03 obj. 4 (As available)

Question Source: Bank #

Modified Bank #

(Note changes or attach parent)

New X Question History: Last NRC Exam None Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 1 Comments:

2015 NRC Exam SRO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 39 of 50 Examination Outline Cross-

Reference:

Level RO SRO Tier # 3 Group # 1 K/A # 2.1.34 Importance Rating 3.5 K/A Statement 2.1.34 Knowledge of primary and secondary plant chemistry limits.

Proposed Question: SRO 20 Rev: 0 The surveillance requirements of TS 3.4.7, Specific Activity, states that Chemistry will be notified to perform a sample of the RCS for an isotopic iodine analysis following a reactor power change of greater than ___(1)___ percent within a one hour period. This surveillance requirement states that the sample will be taken two to ___(2)___

hours following the downpower.

(1) (2) A. 10 six B. 10 four C. 15 four D. 15 six 2015 NRC Exam SRO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 40 of 50 Proposed Answer: D Explanation: (Optional)

A. Incorrect. The surveillance requirements of TS 3.4.7, Specific Activity, states that Chemistry will be notified to perform a sample of the RCS for an isotopic iodine analysis following a reactor power change of greater than 15% power. Part 2 is correct. B. Incorrect. The surveillance requirements of TS 3.4.7, Specific Activity, states that Chemistry will be notified to perform a sample of the RCS for an isotopic iodine analysis following a reactor power change of greater than 15% power. The sample will be taken 2 to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> following the downpower. 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is plausible since it is the sample frequency if specific activity of iodine exceeds 1.0 per TS table 4.4-4. C. Incorrect. Part 1 is correct. The sample will be taken 2 to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> following the downpower. 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is plausible since it is the sample frequency if specific activity of iodine exceeds 1.0 per TS table 4.4-4.

D. CORRECT: The surveillance requirements of TS 3.4.7, Specific Activity, states that Chemistry will be notified to perform a sample of the RCS for an isotopic iodine analysis following a reactor power change of greater than 15% power. The sample will be taken 2 to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> following the downpower. (TS Table 4.4-4)

Technical Reference(s): OP-901-212 step 10 revision 7 (Attach if not previously provided) TS table 4.4-4 (including version/revision number)

Proposed references to be provided to applicants during examination:

None Learning Objective: WLP-OPS-CHM03 obj. 14 (As available)

Question Source: Bank #

Modified Bank #

(Note changes or attach parent)

New X Question History: Last NRC Exam None Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 2 Comments:

2015 NRC Exam SRO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 41 of 50 Examination Outline Cross-

Reference:

Level RO SRO Tier # 3 Group # 2 K/A # 2.2.18 Importance Rating 3.9 K/A Statement 2.2.18 Knowledge of the process for managing maintenance activities during shutdown operations, such as risk assessments, work prioritization, etc.

Proposed Question: SRO 21 Rev: 0 For planned outages, a risk assessment of the outage schedule is performed per the requirements located in ___(1)___. Based on the risk assessment process, the responsibility for ensuring the hanging, tracking, and prompt removal of protected equipment postings is the responsibility of the ___(2)___. (1) (2) A. PLG-009-014, Conduct of Planned outages Shift Manager B. PLG-009-014, Conduct of Planned outages Outage Manager C. OI-037-000, Operations' Risk Assessment Guideline Shift Manager D. OI-037-000, Operations' Risk Assessment Guideline Outage Manager

2015 NRC Exam SRO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 42 of 50 Proposed Answer: A Explanation: (Optional)

A. CORRECT: For modes 4, 5 and 6, risk is assessed via the guidance in PLG-009-014, Conduct of Planned Outages. This is defined in both OI-037-000 and PLG-009-014. Ensuring the hanging, tracking, and prompt removal of protected equipment postings is the responsibility of the Shift Manager as defined in EN-OP-119, Protected Equipment Postings. B. Incorrect. Part 1 is correct. Ensuring the hanging, tracking, and prompt removal of protected equipment postings is the responsibility of the Shift Manager as defined in EN-OP-119, Protected Equipment Postings. The Outage Manager is the title of a position at W3, but the position is responsible for coordinating refuel outages, not planned outages. C. Incorrect. For modes 4, 5 and 6, risk is assessed via the guidance in PLG-009-014, Conduct of Planned Outages. For modes 1-3, OI-037-000, would be used to perform risk assessments.This is defined in both OI-037-000 and PLG-009-014. Part 2 is correct. D. Incorrect. For modes 4, 5 and 6, risk is assessed via the guidance in PLG-009-014, Conduct of Planned Outages. This is defined in both OI-037-000 and PLG-009-014. For modes 1-3, OI-037-000, would be used to perform risk assessments. The Outage Manager is the title of a position at W3, but the position is responsible for coordinating refuel outages, not planned outages.

Technical Reference(s): OI-037-000 step 5.2.5.2 revision 306 (Attach if not previously provided) PLG-009-014 revision 313 (including version/revision number) EN-OP-119 page 13 revision 7 Proposed references to be provided to applicants during examination:

None Learning Objective: WLP-OPS-ORA obj.2 (As available)

WLP-OPS-ORA obj.3 Question Source: Bank #

Question 22 Modified Bank #

(Note changes or attach parent)

New X Question History: Last NRC Exam 2012 NRC SRO Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 5 Comments:

2015 NRC Exam SRO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 43 of 50 Examination Outline Cross-

Reference:

Level RO SRO Tier # 3 Group # 3 K/A # 2.3.5 Importance Rating 2.9 K/A Statement 2.3.5 Ability to use radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc.

Proposed Question: SRO 22 Rev: 0 Given: RAD MONITOR SYS ACTIVITY HI-HI annunciator is in alarm Plant Stack PIG A and B show rising activity HVAC Duct PIG D shows rising activity HVAC Duct PIG A is in alarm The CRS should carry out the actions in:

A. OP-901-401, High Airborne Activity in Control Room B. OP-901-402, High Airborne Activity in Reactor Auxiliary Building C. OP-901-403, High Airborne Activity in Containment D. OP-901-404, High Airborne Activity in Fuel Handling Building

2015 NRC Exam SRO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 44 of 50 Proposed Answer: B Explanation: (Optional)

A. Incorrect. CROAI (North & South) must be in alarm with rising activity B. CORRECT. RAB HVAC D (RAB +46) and RAB HVAC A (RAB -4) are in alarm with Plant Stack activity rising. HVAC Duct PIG A, B, C and D detect airborne activity in the RAB. HVAC D samples at the inlet to the RAB Normal Exhaust Fans. HVAC D should rise in activity if HVAC A, B, or C is in alarm. The applicant will use this knowledge of Rad Monitors to determine what offnormal to enter. C. Incorrect. RCB +46 and +21 monitors must be in alarm with rising activity D. Incorrect. FHB EXH monitors must be in alarm with rising activity Technical Reference(s): OP-901-402 revision 3 (Attach if not previously provided) (including version/revision number)

Proposed references to be provided to applicants during examination:

None Learning Objective: WLP-OPS-RMS00 Obj.3 (rev27) (As available)

WLP-OPS-PPO40 Obj.1 (rev4)

Question Source: Bank #

X Question # 22 Modified Bank #

(Note changes or attach parent)

New Question History: Last NRC Exam 2011 NRC SRO Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 4 Comments:

2015 NRC Exam SRO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 45 of 50 Examination Outline Cross-

Reference:

Level RO SRO Tier # 3 Group # 3 K/A # 2.3.13 Importance Rating 3.8 K/A Statement 2.3.13 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: SRO 23 Rev: 0 The following plant conditions exist: A radiological release is in progress following a LOCA outside of Containment The Emergency Director is evaluating E Plan Data is now available for a PARs determination The EC directs you to perform this determination per EP-002-052, Protective Action Guidelines. The following data is available: Duration of release is unknown Wind Directions is from 345 º EAB TEDE Mr/hr EAB CDE Thyroid Mr/hr 2 mile TEDE Mr/hr 2 mile CDE Thyroid Mr/hr 5 mile TEDE Mr/hr 5 mile CDE Thyroid Mr/hr 1250 3200 750 1900 50 730 Which ONE of the following PAR actions meets the required recommendations of EP-002-052, Protective Action Guidelines?

EVACUATE A. A1, B1, C1, D1 B. A1, B1, C1, D1, D2 C. A1, B1, C1, D1, B2, D2 D. A1, B1, C1, D1, D2, D3, D4

2015 NRC Exam SRO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 46 of 50 Proposed Answer: B Explanation: (Optional)

A. INCORRECT: Student would arrive at this answer if he doesn't multiply the dose by 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> as required, since duration length is NOT known. B. CORRECT: Double rate and use response areas from sector R C. INCORRECT: Correct dose but uses sector Q vice R. D. INCORRECT: Uses correct sector, but evacuates all 3 response areas. Not required to evacuate 5 miles downwind.

Technical Reference(s): EP-002-050, Page 7, revision 306 (Attach if not previously provided) EP-002-052 attachments 7.2 and 7.3 revision 24 (including version/revision number)

Proposed references to be provided to applicants during examination: EP-002-052 attachments 7.2 and 7.3 revision 24 Learning Objective: WLP-OPS-EP02 obj. 23 (As available)

Question Source: Bank #

X Question #23 Modified Bank #

(Note changes or attach parent)

New Question History: Last NRC Exam 2009 NRC SRO Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 5 Comments:

2015 NRC Exam SRO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 47 of 50 Examination Outline Cross-

Reference:

Level RO SRO Tier # 3 Group # 4 K/A # 2.4.38 Importance Rating 4.4 K/A Statement 2.4.38 Ability to take actions called for in the facility emergency plan, including supporting or acting as emergency coordinator if required.

Proposed Question: SRO 24 Rev: 0 As the Emergency Director, what is the time limit for notification of Operations Hotline Members when plant conditions warrant upgrading from an Unusual Event to an Alert?

A. 10 minutes B. 15 minutes C. 30 minutes D. 60 minutes

2015 NRC Exam SRO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 48 of 50 Proposed Answer: B Explanation: (Optional)

A. Incorrect. 10 minutes is the target time limit for notifying emergency response members following an Emergency Declaration B. CORRECT: The time limit for notifying off site agencies following an upgrade is 15 minutes. C. Incorrect. This is the time limit for sending the Notification Message Form after using a Short Message Form. D. Incorrect. This is the time limit for notifying the NRC following an emergency classification upgrade.

Technical Reference(s): EP-002-010 note before step 5.1.6.1 revision 311 (Attach if not previously provided) (including version/revision number)

Proposed references to be provided to applicants during examination:

None Learning Objective: WLP-OPS-EP02 obj. 9 (As available)

Question Source: Bank #

X Question #25 Modified Bank #

(Note changes or attach parent)

New Question History: Last NRC Exam 2009 SRO Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 5 Comments:

2015 NRC Exam SRO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 49 of 50 Examination Outline Cross-

Reference:

Level RO SRO Tier # 3 Group # 4 K/A # 2.4.35 Importance Rating 4.0 K/A Statement 2.4.35 Knowledge of local auxiliary operator tasks during an emergency and the resultant operational effects.

Proposed Question: SRO 25 Rev: 0 Given: A reactor trip has occurred The crew is performing Standard Post Trip actions per OP-902-000 and power to the Moisture Separator Re-heater (MSR) Control panel is not available.

The BOP will direct the auxiliary operator to isolate the Moisture Separator Re-heaters by ____(1)____ valves. This action is required to _____(2)_____ . (1) (2) A. manually closing temperature control valve isolation prevent overcooling the RCS B. failing closed air operated temperature control prevent overcooling the RCS C. failing closed air operated temperature control prevent over-pressurization of the MSR D. manually closing temperature control valve isolation prevent over-pressurization of the MSR 2015 NRC Exam SRO Written Exam Worksheet Revision 0 Facility: Waterford 3 Page 50 of 50 Proposed Answer: B Explanation: (Optional)

A. Incorrect. OI-038-000, EOP Operations/Expectations procedure directs the crew to isolate the MSR by isolating the air and opening the petcock to all 5 inch and 10 inch valves. MOV isolation valves are located on the same line as the AOVs but are automatically closed by the AOV limit switches. Part 2 is correct. B. CORRECT: OI-038-000, EOP Operations/Expectations procedure states "If power to the MSR control panel is not available, then the Control Room should direct the NAO to isolate the MSR. The NAO should isolate the air and open the petcock to all 5 inch and 10 inch valves. This guidance is not located in EOPs and is applicable during all events. Therefore this question is Tier 3. C. Incorrect. Part 1 is correct. These actions will eliminate steam to the MSR thereby it is plausible to consider that closing the valves will prevent over-pressurization of the MSR. D. Incorrect. OI-038-000, EOP Operations/Expectations procedure directs the crew to isolate the MSR by isolating the air and opening the petcock to all 5 inch and 10 inch valves. MOV isolation valves are located on the same line as the AOVs but are automatically closed by the AOV limit switches. This action will eliminate steam to the MSR thereby it is plausible to consider that closing the valves will prevent over-pressurization of the MSR.

Technical Reference(s): OI-038-000, step 5.2.3 revision 10 (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. 4 (As available)

Question Source: Bank #

Modified Bank #

(Note changes or attach parent)

New X Question History: Last NRC Exam None Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 5 Comments:

.ES-401 PWR Examination Outline Form ES-401-2 Facility: Waterford 3 (RO Exam) Date of Exam:

September 14, 2015 Tier Group RO K/A Category Points SRO-Only Points K 1 K 2 K 3 K 4 K 5 K 6 A 1 A 2 A 3 A 4 G* Total A2 G* Total 1. Emergency & Abnormal Plant Evolutions 1 3 3 3 N/A 3 3 N/A 3 18 6 2 2 1 2 1 2 1 9 4 Tier Totals 5 4 5 4 5 4 27 10 2. Plant Systems 1 3 2 2 3 2 3 3 2 2 3 3 28 5 2 1 1 1 1 1 1 1 1 1 1 10 3 Tier Totals 4 3 3 4 3 4 3 3 3 4 4 38 8 3. Generic Knowledge and Abilities Categories 1 3 2 2 3 2 4 3 10 1 2 3 4 7 Note: 1. Ensure that at least two topics from every applicable K/A category are sampled within each tier of the RO and SRO-only outlines (i.e., except for one category in Tier 3 of the SRO

-only outline, the "Tier Totals" in each K/A category shall not be less than two). (One Tier 3 Radiation Control K/A is allowed if the K/A is replaced by a K/A from another Tier 3 Category).

2. The point total for each group and tier in the proposed outline must match that specified in the table. The final point total for each group and tier may deviate by +/-1 from that specified in the table based on NRC revisions. The final RO exam must total 75 points and the SRO

-only exam must total 25 points.

3. Systems/evolutions within each group are identified on the associated outline; systems or evolutions that do not apply at the facility should be deleted with justification; operationally important, site

-specific systems/evolutions that are not included on the outline should be added. Refer to Section D.1.b of ES

-401 for guidance regarding the elimination of inappropriate K/A statements.

4. Select topics from as many systems and evolutions as possible; sample every system or evolution in the group before selecting a second topic for any system or evolution.

5. Absent a plant

-specific priority, only those K/As having an importance rating (IR) of 2.5 or higher shall be selected. Use the RO and SRO ratings for the RO and SRO

-only portions, respectively.

6. Select SRO topics for Tiers 1 and 2 from the shaded systems and K/A categories.

7. The generic (G) K/As in Tiers 1 and 2 shall be selected from Section 2 of the K/A Catalog, but the topics must be relevant to the applicable evolution or system. Refer to Section D.1.b of ES

-401 for the applicable K/As. 8. On the following pages, enter the K/A numbers, a brief description of each topic, the topics' importance ratings (IRs) for the applicable license level, and the point totals (#) for each system and category. Enter the group and tier totals for each category in the table above; if fuel handling equipment is sampled in a category other than Category A2 or G* on the SRO

-only exam, enter it on the left side of Column A2 for Tier 2, Group 2 (Note #1 does not apply). Use duplicate pages for RO and SRO

-only exams.

9. For Tier 3, select topics from Section 2 of the K/A catalog, and enter the K/A numbers, descriptions, IRs, and point totals (#) on Form ES

-401-3. Limit SRO selections to K/As that are linked to 10 CFR 55.43.

G* Generic K/As ES-401 2 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant Evolutions

- Tier 1/Group 1 (RO / SRO) E/APE # / Name / Safety Function K 1 K 2 K 3 A 1 A 2 G* K/A Topic(s)

IR # 000007 (BW/E02&E10; CE/E02) Reactor Trip - Stabilization

- Recovery / 1 X CE/E02, EK2.2:

Knowledge of the interrelations between the (Reactor Trip Recovery) and the following:

Facility's heat removal systems, including primary coolant, emergency

coolant, the decay heat removal system s, and relations between the

proper operation of these systems to the operation of the facility.

(CFR: 41.7 / 45.7) 3.5 1 000008 Pressurizer Vapor Space Accident / 3 X AA2.15: Ability to determine and interpret the following as they apply

to the Pres surizer Vapor Space Accident:

ESF control board, valve controls, and indicators (CFR: 43.5 /

45.13). 3.9 1 000009 Small Break LOCA / 3 X EA2.11: Ability to determine or interpret the following as they apply

to a small break LOCA

Containment temperat ure, pressure, and humidity (CFR 43.5 / 45.13).

3.8 1 000011 Large Break LOCA / 3 X 2.4.50: Ability to verify system alarm setpoints and operate controls identified in the alarm response manual (CFR: 41.10 / 43.5 / 45.3).

4.2 1 000015/17 RCP Malfunctions / 4 X AA1.03: Ability to operate and / or monitor the following as they apply to the Reactor Coolant Pump Malfunctions (Loss of RC Flow):

Reactor trip alarms, switches, and indicators (CFR 41.7 / 45.5 / 45.6).

3.7 1 000022 Loss of Rx Coolant Makeup / 2 X AA1.01: Ability to operate and / or monitor the following as they apply to the Loss of Reactor Coolant Makeup:

CVCS Letdown and Charging (CFR 41.7 /

45.5 / 45.6)

. 3.4 1 000025 Loss of RHR System / 4 X 2.1.30: Ability to locate and op erate components, including local controls (CFR: 41.7 / 45.7).

4.4 1 000026 Loss of Component Cooling Water / 8 X AK3.04: Knowledge of the reasons for the following responses as they apply to the Loss of Component Cooling Water: Effect on the CCW fl ow header of a loss of CCW (CFR 41.5, 41.10 /

45.6 / 45.13).

3.5 1 000027 Pressurizer Pressure Control System Malfunction / 3 X AA2.11: Ability to determine and interpret the following as they apply to the Pressurizer Pressure Control Malfunctions:

R CS pressure (CFR: 43.5

/ 45.13). 4.0 1 000029 ATWS / 1 X EK1.03: Knowledge of the operational implications of the following concepts as they apply to the ATWS:

Effects of boron on reactivity (CFR 41.8 / 41.10

/ 45.3). 3.6 1 000038 Steam Gen. Tube Rupture / 3 X EA1.39: Ability to operate and monitor the following as they apply to a SGTR:

Drawing S/G into the RCS, using the "feed and bleed" method (CFR 41.7 /

45.5 / 45.6)

. 3.6 1 000040 (BW/E05; CE/E05; W/E12) Steam Line Rupture

- Excessive Heat Transfer / 4 X CE/E05, EK2.1: Knowledge of the interrelations between the (Excess

Steam Demand) and the following:

Components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic an d manual features (CFR: 41.7 / 45.7).

3.3 1 000054 (CE/E06) Loss of Main Feedwater / 4 X CE/E06, EK3.2: Knowledge of the reasons for the following responses as

they apply to the (Loss of Feedwater):

Normal, abnormal and emergency operating procedure s associated with (Loss of Feedwater) (CFR: 41.5 /

41.10, 45.6 / 45.13).

3.2 1 000055 Station Blackout / 6 X 2.2.37: Ability to determine operability and/or availability of

safety related equipment (CFR: 41.7 /

41.7 / 43.5 / 45.1 2). 3.6 1 000056 Loss of Off-site Power / 6 X AK1.03: Knowledge of the operational implications of the following concepts as they apply to Loss of Offsite Power: Definition of subcooling: use of steam tables to determine it (CFR 41.8 / 41.10 / 45.3)

. 3.1 1 000057 Loss of Vital AC Inst. Bus / 6 000058 Loss of DC Power / 6 X AK1.01: Knowledge of the operational implications of the following concepts as they apply to Loss of DC Power:

Battery charger equipment and instrumentation (CFR 41.8 / 41.10 /

45.3). 2.8 1 000062 Loss of Nuclear Svc Water / 4 X AK3.04: Knowledge of the reasons for the following responses as they apply to the Loss of Nuclear Service Water:

Effect on the nuclear service water discharge flow header of a loss of CCW (CFR 41.4, 41.8 / 4 5.7 ). 3.5 1 000065 Loss of Instrument Air / 8 W/E04 LOCA Outside Containment / 3 W/E11 Loss of Emergency Coolant Recirc. / 4 BW/E04; W/E05 Inadequate Heat Transfer - Loss of Secondary Heat Sink / 4 000077 Generator Voltage and Electric Grid Disturbances / 6 X AK2.01: Knowledge of the interrelations between Generator

Voltage and Electric Grid Disturbances and the following:

Motors (CFR: 41.4, 41.5, 41.7, 41.10 / 45.8).

3.1 1 K/A Category Totals: 3 3 3 3 3 3 Group Point Total:

18/6 ES-401 3 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant Evolutions

- Tier 1/Group 2 (RO / SRO) E/APE # / Name / Safety Function K 1 K 2 K 3 A 1 A 2 G* K/A Topic(s)

I R # 000001 Continuous Rod Withdrawal / 1 X AA2.03: Ability to determine and interpret the following as they apply to the Continuous Rod Withdrawal

Proper actions to be taken if automatic safety functions have not taken place (CFR: 43.5 / 45.13).

4.5 1 000003 Dropped Control Rod / 1 000005 Inoperable/Stuck Control Rod / 1 X AK3.05: Knowledge of the reasons for the following responses as

they apply to the Inoperable /

Stuck Control Rod:

Power limits on rod misalignment (CFR 41.5, 41.10 / 45.6 / 45.13).

3.4 1 000024 Emergency Boration / 1 000028 Pressurizer Level Malfunction / 2 000032 Loss of Source Range NI / 7 X AA1.01: Ability to operate and /

or monitor the following as they

apply to the Loss of Source Range Nuclear Instrumentation:

Manual restoration of power (CFR 41.7 /

45.5 / 45.6).

3.1 1 000033 Loss of Intermediate Range NI / 7 000036 (BW/A08) Fuel Handling Accident / 8 X 2.4.11 Knowledge of abnormal condition procedures (CFR: 41.10

/ 43.5 / 45.13). 4.0 1 000037 Steam Generator Tube Leak / 3 X AK1.01 Knowledge of the operational implications of the following concepts as they apply to Steam Generator Tube Leak:

Use of Steam Tables (CFR 41.8 /

41.10 / 45.3).

2.9 1 000051 Loss of Condenser Vacuum / 4 000059 Accidental Liquid Radwaste Rel. / 9 000060 Accidental Gaseous Radwaste Rel. / 9 000061 ARM System Alarms / 7 X AK1.01: Knowledge of the operational implications of the following concepts as they appl y to Area Radiation Monitoring (ARM) System Alarms:

Detector limitations (CFR 41.8 / 41.10 /

45.3). 2.5 1 000067 Plant Fire On

-site / 8 000068 (BW/A06) Control Room Evac. / 8 X AA2.08: Ability to determine and interpret the following as the y

apply to the Control Room Evacuation:

S/G pressure (CFR: 43.5 / 45.13).

3.9 1 000069 (W/E14) Loss of CTMT Integrity / 5 X AAK2.03: Knowledge of the interrelations between the Loss of Containment Integrity and the

following:

Personnel access hatch a nd emergency access hatch (CFR 41.7 / 45.7).

2.8 1 000074 (W/E06&E07)

Inad. Core Cooling

/ 4 000076 High Reactor Coolant Activity / 9 W/EO1 & E02 Rediagnosis

& SI Termination

/ 3 W/E13 Steam Generator Over

-pressure / 4 W/E15 Containment Flooding / 5 W/E16 High Containment Radiation / 9 BW/A01 Plant Runback / 1 BW/A02&A03 Loss of NNI

-X/Y / 7 BW/A04 Turbine Trip / 4 BW/A05 Emergency Diesel Actuation / 6 BW/A07 Flooding / 8 BW/E03 Inadequate Subcooling Margin / 4 BW/E08; W/E03 LOCA Cooldown - Depress. / 4 BW/E09; CE/A13; W/E09&E10 Natural Circ. / 4 BW/E13&E14 EOP Rules and Enclosures CE/A11; W/E08 RCS Overcooling

- PTS / 4 CE/A16 Excess RCS Leakage / 2 X AK3.3: Knowledge of the reasons for the following responses as they apply to the (Excess RCS Leakage): Manipulation of controls required to obtain desired operating results during abnormal, and emergency situations

. (CFR: 41.5 / 41.10, 45.6, 45.13).

3.3 1 CE/E09 Functional Recovery K/A Category Point Totals:

2 1 2 1 2 1 Group Point Total:

9/4 ES-401 4 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Plant Systems

- Tier 2/Group 1 (RO / SRO) System # / Name K 1 K 2 K 3 K 4 K 5 K 6 A 1 A 2 A 3 A 4 G* K/A Topic(s)

IR # 003 Reactor Coolant Pump X X A1.09: Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the RCPS controls including

Seal flow and D/P

. (CFR: 41.5 /

45.5). A3.05: Ability to monitor automatic operation of the RCPS, including:

RCP lube oil and bearing lift pumps (CFR:

41.7 / 45.5).

2.8

2.7 1

1 004 Chemical and Volume Control X K1.02: Knowledge of the physical connections and/or

cause-effect relationships between the CVCS and the following systems:

PZR and RCS temperature and pressure relationships (CFR: 41.2 to 41.9 / 45.7 to 45.8).

3.5 1 005 Residual Heat Removal X A1.03: Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the RHRS controls including:

Closed cooling water flow rate and temperature (CFR:

41.5 / 45.5). 2.5 1 006 Emergency Core Cooling X X K2.04: Knowledge of bus power supplies to the following:

ESFAS-operated valves (CFR:

41.7).

K6.03: Knowledge of the effect of a loss or malfunction on the following will have on the ECCS:

Safety Injectio n Pumps (CFR: 41.7 /

45.7) 3.6 3.6 1

1 007 Pressurizer Relief/Quench Tank X K3.01: Knowledge of the effect that a loss or

malfunction of the PRTS will have on the following:

Containment (CFR: 41.7 /

45.6). 3.3 1 008 Component Cooling Water X A1.01: Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the CCWS controls including

CCW flow rate (CFR: 41.5 / 45.5).

2.8 1 010 Pressurizer Pressure Control X K6.03: Knowledge of the effect of a loss or malfunction of the following will have on the PZR PCS:

PZR sprays and heaters (CFR: 41.7

/ 45.7). 3.2 1 012 Reactor Protection X X K2.01: Knowledge of bus power supplies to the following:

RPS channels, com ponents, and interconnections (CFR: 41.7).

2.4.45: Ability to prioritize and interpret the significance of each annunciator or alarm (CFR: 41.10 / 43.5 / 45.3 / 45.12).

3.3 4.5 1 1 013 Engineered Safety Features Actuation X X K5.01: Knowledge of the operational implications of

the following concepts as they apply to the ESFAS:

Definitions of safety train and ESF channel (CFR: 41.5 /

45.7). A3.01: Ability to monitor automatic operation of the ESFAS including:

Input channels and logic (CFR: 4 1.7 / 45.5). 2.8

3.7 1

1 022 Containment Cooling X K 4.02: Knowledge of CCS design feature(s) and/or interlock(s) which provide for the following:

Correlation of fan speed and flowpath changes with containment pressure (CFR:

41.7). 3.1 1 025 Ice Condenser System not part of plant design. 026 Containment Spray X A4.05: Ability to manually operate and/or monitor in the control room:

Containment spray reset switches (CFR:

41.7 / 45.5 to 45.8).

3.5 1 039 Main and Reheat Steam X K5.05: Knowledge of the operational implications of

the following concepts as the apply to the MRSS:

Bases for RCS cooldown limits (CFR:

441.5 / 45.7).

2.7 1 059 Main Feedwater X K4.16: Knowledge of MFW design feature(s) and/or interlock(s) which provide

for the following:

Automatic trips for MFW pumps (CFR:

41.7). 3.1 1 061 Auxiliary/Emergency Feedwater X K6.01: Knowledge of the effect of a loss or malfunction of the following will have on the AFW components:

Controllers and positioners (CFR: 41.7 /

45.7). 2.5 1 062 AC Electrical Distribution X X A2.09: 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:

Consequences of exceeding current limitations (CFR: 41.5 / 43.5 / 45.3 /

45.13). 2.2.36: Ability to analyze the effect of maintenance activities, such as de graded power sources, on the status of limiting conditions for operations.(CFR: 41.10 / 43.2

/ 45.13). 2.7

3.1 1

1 063 DC Electrical Distribution X K3.01: Knowledge of the effect that a loss or malfunction of the DC electrical syst em will have on the following:

ED/G (CFR:

41.7 / 45.6).

3.7 1 064 Emergency Diesel Generator X X A2.09: 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:

Synchronization of the ED/G with other electric power supplies. (CFR: 41.5 / 43.5 /

45.3 / 45.13).

2.1.27: Knowledge of system purpose and/or function (CFR: 41.7). 3.1

3.9 1

1 073 Process Radiation Monitoring X X 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 (CFR: 41.2 to 41.9 /

45.7 to 45.8)

. A4.03: Ability to manually operate and/or monitor in the control room:

Check source for operability demonstration (CFR: 41.7 / 45.5 to 45.8).

3.6

3.1 1

1 076 Service Water X K4.02: Knowledge of SWS design feature(s) and/o r interlock(s) which provide for the following:

Automatic start features associated with SWS pump controls.

(CFR: 41/7). 2.9 1 078 Instrument Air X A4.01: Ability to manually operate and/or monitor in the control room:

Pressure gauges (CFR: 41.7

/ 45.5 to 45.8).

3.1 1 103 Containment X K1.03: Knowledge of the physical connections and/or cause/effect relationships between the containment system and the following

systems: Shield building vent system (CFR: 41.2 to 41.9 /

45.7 to 45.8).

3.1 1 K/A Category Point Totals:

3 2 2 3 2 3 3 2 2 3 3 Group Point Total:

28/5 ES-401 5 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Plant Systems

- Tier 2/Group 2 (RO / SRO) System # / Name K 1 K 2 K 3 K 4 K 5 K 6 A 1 A 2 A 3 A 4 G* K/A Topic(s)

IR # 001 Control Rod Drive X K5.08 Knowledge of the operational implication of the following concepts as they apply to the CRDS:

Reason for rod insertion limits and their effect on shutdown margin (CFR: 41.5 / 45.7).

3.9 1 002 Reactor Coolant X K6.12: Knowledge of the effect or a loss or malfunction on the following RCS components:

Code Safety valves (CFR: 41.7 /

45.7). 3.0 1 011 Pressurizer Level Control X A1.04: Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the PZR LCS controls including:

T-ave (CFR: 41.5 /

45.5). 3.1 1 014 Rod Position Indication 015 Nuclear Instrumentation X A3.02: Ability to monitor automatic operation of the NIS, including:

Annunciator and alarm signals (CFR: 41.7 / 45.5).

3.7 1 016 Non-Nuclear Instrumentation X K4.03: Knowledge of NNIS design feature(s) and/or interlock(s) which provide for the following:

Input to control systems

. (CFR: 41.7). 2.8 1 017 In-Core Temperature Monitor 027 Containment Iodine Removal 028 Hydrogen Recombiner and Purge Control X 2.1.20: Ability to interpret and execute proc edure steps (CFR: 41.10 / 43.5 / 45.12).

4.6 1 029 Containment Purge 033 Spent Fuel Pool Cooling 034 Fuel Handling Equipment 035 Steam Generator X A4.01: Ability to manually operate and/or monitor in the control room:

Shift of S/G controls between manual and automatic control, by bumpless transfer (CFR: 41.7 / 45.5 to 45.8). 3.7 1 041 Steam Dump/Turbine Bypass Control 045 Main Turbine Generator 055 Condenser Air Removal X K1.06: Knowledge of the physical connections and/or cause/effect relationships between the CARS and the following systems:

PRM system (CFR: 41.2 to 41.9 / 45.7 to 45.8). 2.6 1 056 Condensate

068 Liquid Radwaste 071 Waste Gas Disposal X K3.05: Knowledge of the effect that a loss or malfunction of the Waste Gas Disposal System will have on the following:

ARM and PRM systems

. (CFR: 41.7 /

45.6). 3.2 1 072 Area Radiation Monitoring 075 Circulating Water X K2.03: Knowledge of bus power supplies to the following:

Emergency/essential SWS pumps (CFR: 41.7).

2.6 1 079 Station Air 086 Fire Protection K/A Category Point Totals:

1 1 1 1 1 1 1 0 1 1 1 Group Point Total:

10/3 ES-401 Generic Knowledge and Abilities Outline (Tier 3)

Form ES-401-3 Facility: Waterford 3 (RO) Date of Exam:

September 14, 2015 Category K/A # Topic RO SRO-Only IR # IR # 1. Conduct of Operations 2.1.1 Knowledge of conduct of operations requirements.

(CFR: 41.10 / 45.13) 3.8 1 2.1.42 Knowledge of new and spent fuel movement procedures. (CFR: 41.10 / 43.7 / 45.13) 2.5 1 2.1.45 Ability to identify and interpret diverse indications to validate the response of another indication. (CFR: 41.7 / 43.5 / 45.4) 4.3 1 2.1. Subtotal 2. Equipment Control 2.2.21 Knowledge of pre

- and post-maintenance operability requirements.

(CFR: 41.10 / 43.2) 2.9 1 2.2.40 Ability to apply Technical Specifications for a system. (CFR: 41.10 / 43.2 / 43.5 / 45.3) 3.4 1 2.2. Subtotal 3. Radiation Control 2.3.11 Ability to control radiation releases. (CFR: 41.11 / 43.4 / 45.10) 3.8 1 2.3.14 Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or emergency conditions or activities.

(CFR: 41.12 / 43.4 / 45.10) 3.4 1 2.3. Subtotal 4. Emergency Procedures /

Pl an 2.4.1 2 2.4.12 Knowledge of general operating crew responsibilities during emergency operations (CFR: 41.10 / 45.12) 4.0 1 2.4.42 Knowledge of emergency response facilities.

(CFR: 41.10 / 45.11) 2.6 1 2.4.49 Ability to perform without reference to procedures those actions that require immediate operation of system components and controls. (CFR: 41.10 / 43.2 / 45.6) 4.6 1 2.4. Subtotal Tier 3 Point Total 10 7 ES-401 Record of Rejected K/As Form ES-401-4 Tier / Group Randomly Selected K/A Reason for Rejection 1/1 Question #6 000022/AA1.07 W3 does not have an excess letdown system. Randomly selected another K/A from the AA1 section of 000022.

Randomly selected 000022 AA1.01, CVCS letdown and charging. 1/1 Question

1. 14 00055/ 2.2.39 W3 does not have one hour TS actions for a Station Blackout. Tech Specs are not addressed when in EOP space. Randomly selected a different K/A under Equipment Control (2.2.3 7) while remaining in the SBO area. 1/2 Question

1. 22 00036/ 2.4.1 W3 has no EOP entry or immediate action steps for a fuel handling accident.

W3 has an offnormal procedure for guidance upon a fuel handling accident. Randomly selected a different K/A under the Emergency Procedures section (2.4.11) while remaining in the fuel accident area.

1/2 Question

1. 23 00059/ AK 1.05 Calculation of offsite doses due to a release from the LWM system is not performed by a LO at W3, therefore, no guidance exist in any operations procedures.

Randomly selected a n Emergency or Plant Evolution from Tier 1/Group 2 limited to those not used in the original outline, then randomly selected a K/A from the

AK1 area. Randomly selected 037 Steam Generator tube leakage AK1.01 1/2 Question

1. 27 CE/A16 AK3.1 A question could not be developed for this K/A that was supported by written guidance.

The W3 RCS leak offnormal does not go into the level of detail to support this K/A. Randomly selected a different K/A for RCS leakage while staying in the AK3 area. Randomly selected CE/A16 AK3.3

. 2/1 Question

1. 2 8 003 A1.01 A question could not be developed for this K/A. The RCP off-normal does not have specific guidance that would support a test question. The W3 off

-normal has guidance to get Engineering input on high vibration and actions to be taken if the SM/CRS determines that the vibration is valid. There is no trigger points nor expected mil values for the RO to take actions based on RCP vibration. Randomly selected a different K/A for the RCP system while staying in the A1 area. Randomly selected 003 A1.09 2/1 Question #46 062 A2.08 A question could not be developed or modified for this K/A that would deviate enough from question s developed from the previous two exams. Randomly selected a different K/A for the AC electrical distribution system while staying in the A2 area.

The allotment of 4 questions from the previous two exams has already been used.

Randomly selected 062 A2.09 2/1 Question

1. 49 064 A2.16 A question could not be developed for this K/A that was supported by written guidance.

W3 has a procedure for full load testing of the EDG but no guidance in the surveillance procedure to take if a loss of offsite power occurred during this testing. Randomly selected a different K/A for the EDG system while staying in the A2 area. Randomly selected 064 A2.0 9. 2/1 Question #53 076 K4.06 The W3 Auxiliary Component Cooling Water (ACCW) system does not have an interlock or design feature that will separate trains.

The W3 component cooling water system will separate trains and isolate safety and non safety systems. We refer to our ACCW system as the Service Water system as represented in the K/A catalog.

Randomly selected a different K/A for the Service Water system while staying in the K4 area. Randomly selected 076 K4.02 2/2 Question #61 041 K4.11 The W3 SDS system does not input to nor receive an input from the Tave/Tref program. Randomly select a system from Tier 2/ Group 2 not previously used.

Randomly selected a new K4 K/A from this system. Randomly selected 016 K4.03. 2/2 Question

1. 63 068/ K 5.04 Could not develop a question for Biological hazards of radiation and the resulting goal of ALARA for LWM release without selecting a theory question.

Operations procedures also have no written guidance for this subject. Randomly selected a K/A from Tier 2/Group 2 system not previously used while remaining in the K5 area. Randomly selected 001 K5.08 2/2 Question #64 071 K3.04 W3 does not have written guidance on a loss or malfunction of the waste gas disposal system having an affect on ventilation systems.

The Gaseous Waste Management (GWM) system will isolate on high radiation but will not affect a ventilation system. Due to this, we were unable to develop a question that met the K/A. Randomly selected a different K/A for the Waste Gas Disposal system while staying in the K3 area. Randomly selected 071 K3.05 3/4 Question #73 2.4.22 W3 does not have written guidance on the bases for prioritizing safety functions during emergency operations. Could not develop a question with an adequate LOD without having the RO prioritize safety functions in our functional recovery procedure (SRO knowledge). Randomly selected a K/A from section 2.4 ensuring not to use a 2.4 K/A already selected on this outline. Randomly selected 2.4.12 Knowledge of general operating crew responsibilities during emergency operations

ES-401 PWR Examination Outline Form ES-401-2 Facility: Waterford 3 (SRO) Date of Exam:

September 14, 2015 Tier Group RO K/A Category Points SRO-Only Points K 1 K 2 K 3 K 4 K 5 K 6 A 1 A 2 A 3 A 4 G* Total A2 G* Total 1. Emergency & Abnormal Plant Evolutions 1 N/A N/A 18 3 3 6 2 9 2 2 4 Tier Totals 27 5 5 10 2. Plant Systems 1 28 2 3 5 2 10 2 1 3 Tier Totals 38 4 4 8 3. Generic Knowledge and Abilities Categories 1 2 3 4 10 1 2 2 1 3 2 4 2 7 Note: 1. Ensure that at least two topics from every applicable K/A category are sampled within each tier of the RO and SRO-only outlines (i.e., except for one category in Tier 3 of the SRO

-only outline, the "Tier Totals" in each K/A category shall not be less than two). (One Tier 3 Radiation Control K/A is allowed if the K/A is replaced by a K/A from another Tier 3 Category).

2. The point total for each group and tier in the proposed outline must match that specified in the table. The final point total for each group and tier may deviate by +/-1 from that specified in the table based on NRC revisions. The final RO exam must total 75 points and the SRO

-only exam must total 25 points.

3. Systems/evolutions within each group are identified on the associated outline; systems or evolutions that do not apply at the facility should be deleted with justification; operationally important, site

-specific systems/evolutions that are not included on the outline should be added. Refer to Section D.1.b of ES

-401 for guidance regarding the elimination of inappropriate K/A statements.

4. Select topics from as many systems and evolutions as possible; sample every system or evolution in the group before selecting a second topic for any system or evolution.

5. Absent a plant-specific priority, only those K/As having an importance rating (IR) of 2.5 or higher shall be selected. Use the RO and SRO ratings for the RO and SRO

-only portions, respectively.

6. Select SRO topics for Tiers 1 and 2 from the shaded systems and K/A categories.

7. The generic (G) K/As in Tiers 1 and 2 shall be selected from Section 2 of the K/A Catalog, but the topics must be relevant to the applicable evolution or system. Refer to Section D.1.b of ES

-401 for the applicable K/As. 8. On the following pages, enter the K/A numbers, a brief description of each topic, the topics' importance ratings (IRs) for the applicable license level, and the point totals (#) for each system and category. Enter the group and tier totals for each category in the table above; if fuel handling equipment is sampled in a category other than Category A2 or G* on the SRO

-only exam, enter it on the left side of Column A2 for Tier 2, Group 2 (Note #1 does not apply). Use duplicate pages for RO and SRO

-only exams.

9. For Tier 3, select topics from Section 2 of the K/A catalog, and enter the K/A numbers, descriptions, IRs, and point totals (#) on Form ES

-401-3. Limit SRO selections to K/As that are linked to 10 CFR 55.43.

G* Generic K/As

ES-401 2 Form ES-401-2 ES-40 1 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant Evolutions

- Tier 1/Group 1 (RO / SRO) E/APE # / Name / Safety Function K 1 K 2 K 3 A 1 A 2 G* K/A Topic(s)

IR # 000007 (BW/E02&E10; CE/E02) Reactor Trip - Stabilization

- Recovery / 1 000008 Pressurizer Vapor Space Accident / 3 000009 Small Break LOCA / 3 X 2.2.38: Knowledge of conditions and limitations in the facility license (CFR: 41.7 / 41.10 / 43.1 / 45.13).

4.5 1 000011 Large Break LOCA / 3 X 2.4.9: Knowledge of low power/shutdown implications in accident (e.g., loss of coolant accident or loss of residual heat removal) mitigation strategies (CFR: 41.10 / 43.5 /

45.13). 4.2 1 000015/17 RCP Malfunctions / 4 000022 Loss of Rx Coolant Makeup / 2 000025 Loss of RHR System / 4 000026 Loss of Component Cooling Water / 8 X 2.2.12: Knowledge of surveillance procedures.

(CFR: 41.10 / 45.13) 4.1 1 000027 Pressurizer Pressure Control System Malfunction / 3 000029 ATWS /

1 000038 Steam Gen. Tube Rupture / 3 X EA2.15: Ability to determine or interpret the following as they apply to a SGTR:

Pressure at which to maintain RCS during S/G cooldown (CFR: 43.5 / 45.13).

4.4 1 000040 (BW/E05; CE/E05; W/E12) Steam Line Rupture

- Excessive Heat Transfer / 4 000054 (CE/E06) Loss of Main Feedwater / 4 000055 Station Blackout / 6 000056 Loss of Off

-site Power / 6 000057 Loss of Vital AC Inst. Bus / 6 X AA2.19: Ability to determine and interpret the following as they apply to the Loss of Vital AC Instrument

Bus: The plant automatic actions that will occur on the loss of a vital ac electrical instrument bus.

(CFR: 43.5 / 45.13). 3.1 1 000058 Loss of DC Power / 6 000062 Loss of Nuclear Svc Water / 4 000065 Loss of Instrument Air / 8 X AA2.07: Ability to determine and interpret the following as they apply to the Loss of Instrument Air: Whether backup nitrogen supply is controlling valve position (CFR: 43.5

/ 45.13). 3.2 1 W/E04 LOCA Outside Containment / 3 W/E11 Loss of Emergency Coolant Recirc. / 4 BW/E04; W/E05 Inadequate Heat Transfer - Loss of Secondary Heat Sink / 4 000077 Generator Voltage and Electric Grid Disturbances

/ 6 K/A Category Totals:

Group Point Total:

18/6 ES-401 3 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant Evolutions

- Tier 1/Group 2 (RO / SRO) E/APE # / Name / Safety Function K 1 K 2 K 3 A 1 A 2 G* K/A Topic(s)

IR # 000001 Continuous Rod Withdrawal / 1 000003 Dropped Control Rod / 1 X AA2.01: Ability to determine and interpret the following as they apply to the Dropped Control

Rod: Rod position indication to actual rod position (CFR: 43.5 /

45.13). 3.9 1 000005 Inoperable/Stuck Control Rod / 1 000024 Emergency Boration / 1 X 2.2.22: Knowledge of limiting conditions for operations and safety limits (CFR: 41.5 / 43.2 / 45.2). 4.7 1 000028 Pressurizer Level Malfunction / 2 000032 Loss of Source Range NI / 7 000033 Loss of Intermediate Range NI / 7 000036 (BW/A08) Fuel Handling Accident / 8 000037 Steam Generator Tube Leak / 3 000051 Loss of Condenser Vacuum / 4 X AA2.02: Ability to determine and interpret the following as they apply to the Loss of Condenser Vacuum: Conditions requiring reactor and/or turbine trip (CFR: 43.5 / 45.13).

4.1 1 000059 Accidental Liquid Radwaste Rel. / 9 000060 Accidental Gaseous Radwaste Rel. / 9 000061 ARM System Alarms / 7 000067 Plant Fire On

-site / 8 000068 (BW/A06) Control Room Evac. / 8 000069 (W/E14) Loss of CTMT Integrity / 5 000074 (W/E06&E07)

Inad. Core Cooling

/ 4 000076 High Reactor Coolant Activity / 9 W/EO1 & E02 Rediagnosis

& SI Termination

/ 3 W/E13 Steam Generator Over

-pressure / 4 W/E15 Containment Flooding / 5 W/E16 High Containment Radiation / 9 BW/A01 Plant Runback / 1 BW/A02&A03 Loss of NNI

-X/Y / 7 BW/A04 Turbine Trip / 4 BW/A05 Emergency Diesel Actuation / 6 BW/A07 Flooding / 8 BW/E03 Inadequate Subcooling Margin / 4 BW/E08; W/E03 LOCA Cooldown - Depress. / 4 BW/E09; CE/A13; W/E09&E10 Natural Circ. / 4 BW/E13&E14 EOP Rules and Enclosures

CE/A11; W/E08 RCS Overcooling

- PTS / 4 X 2.1.7: Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation (CFR: 41.5 / 43.5 / 45.12 / 45.13).

4.7 1 CE/A16 Excess RCS Leakage / 2 CE/E09 Functional Recovery K/A Category Point Totals:

Group Point Total:

9/4 ES-401 4 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Plant Systems

- Tier 2/Group 1 (RO / SRO) System # / Name K 1 K 2 K 3 K 4 K 5 K 6 A 1 A 2 A 3 A 4 G* K/A Topic(s)

IR # 003 Reactor Coolant Pump 004 Chemical and Volume Control X 2.2.25: Knowledge of the bases in Technical Specifications for limiting conditions for operations and safety limits (CFR: 41.5 / 41.7 / 43.2).

4.2 1 005 Residual Heat Removal X A2.02: Ability to (a) predict the impacts of the following malfunctions or operations on the RHRS, and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Pressure transient protection during cold shutdown (CFR: 41.5 / 43.5 / 45.3 / 45.13).

3.7 1 006 Emergency Core Cooling 007 Pressurizer Relief/Quench Tank 008 Component Cooling Water 010 Pressurizer Pressure Control X 2.2.42: Ability to recognize system parameters that are

entry-level conditions for Technical Specifications (CFR: 41.7 / 41.10 / 43.2 / 43.3 / 45.3).

4.6 1 012 Reactor Protection 013 Engineered Safety Features Actuation 022 Containment Cooling X 2.4.21: Knowledge of the parameters and logic used to assess the status of safety functions, such as reactivity control, core cooling and heat removal, reactor coolant system integrity, containment conditions, radioactivity release control, etc. (CFR: 41.7 / 43.5 / 45.12) 4.6 1 025 Ice Condenser 026 Containment Spray 039 Main and Reheat Steam 059 Main Feedwater 061 Auxiliary/Emergency Feedwater 062 AC Electrical Distribution 063 DC Electrical Distribution 064 Emergency Diesel Generator

073 Process Radiation Monitoring 076 Service Water X A2.02: Ability to (a) predict the impacts of the following malfunctions or operations on the SWS and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Service water header pressure

(CFR: 41.5 / 43.5 / 45.3 / 45.13). 3.1 1 078 Instrument Air 103 Containment K/A Category Point Totals:

Group Point Total:

28/5 ES-401 5 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Plant Systems

- Tier 2/Group 2 (RO / SRO) System # / Name K 1 K 2 K 3 K 4 K 5 K 6 A 1 A 2 A 3 A 4 G* K/A Topic(s)

IR # 001 Control Rod Drive 002 Reactor Coolant 011 Pressurizer Level Control X A2.01: Ability to (a) predict the impacts of the following malfunctions or operations on the PZR LCS; and (b) based on those predictions, use Procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Excessive letdown

. (CFR: 41.5 / 43.5 / 45.3 /

45.13). 3.1 1 014 Rod Position Indication 015 Nuclear Instrumentation 016 Non-Nuclear Instrumentation 017 In-Core Temperature Monitor 027 Containment Iodine Removal 028 Hydrogen Recombiner and Purge Control 029 Containment Purge 033 Spent Fuel Pool Cooling X A2.02: Ability to (a) predict the impacts of the following malfunctions or operations on the Spent Fuel Pool Coolin g

System; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Loss of SFPCS

. (CFR: 41.5 / 43.5 / 45.3 / 45.13) 3.0 034 Fuel Handling Equipment 035 Steam Generator 041 Steam Dump/Turbine Bypass Control 045 Main Turbine Generator X 2.4.31: Knowledge of annunciator alarms, indications, or response procedures (CFR: 41.10 / 45.3).

4.1 1 055 Condenser Air Removal 056 Condensate 068 Liquid Radwaste 071 Waste Gas Disposal 072 Area Radiation Monitoring 075 Circulating Water

079 Station Air 086 Fire Protection K/A Category Point Totals:

Group Point Total:

10/3 ES-401 Generic Knowledge and Abilities Outline (Tier 3)

Form ES-401-3 Facility: Waterford 3 (SRO) Date of Exam:

September 14, 2015 Category K/A # Topic RO SRO-Only IR # IR # 1. Conduct of Operations 2.1.5 Ability to use procedures related to shift staffing, such as minimum crew complement, overtime limitations, etc (CFR: 41.10 / 43.5 / 45.12

). 3.9 1 2.1.34 Knowledge of primary and secondary plant chemistry limits (CFR: 41.10 / 43.5 / 45.12).

3.5 1 2.1. Subtotal 2. Equipment Control 2.2.18 Knowledge of the process for managing maintenance activities during shutdown operations, such as risk assessments, work prioritization, etc. (CFR: 41.10 / 43.5 / 45.13) 3.9 1 2.2. 2.2. Subtotal 3. Radiation Control 2.3.5 Ability to use radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc. (CFR: 41.11 / 41.12 / 43.4 / 45.9) 2.9 1 2.3.13 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. (CFR: 41.12 / 43.4 / 45.9 / 45.10) 3.8 1 2.3. Subtotal 4. Emergency Procedures / Plan 2.4.3 8 Ability to take actions called for in the facility emergency plan, including supporting or acting as emergency coordinator if required.

(CFR: 41.10 / 43.5 / 45.1 1). 4.4 1 2.4.35 Knowledge of local auxiliary operator tasks during an emergency and the resultant operational effects (CFR: 41.10 / 43.5 / 45.13).

4.0 1 2.4. Subtotal Tier 3 Point Total 10 7 ES-401 Record of Rejected K/As Form ES-401-4 Tier / Group Randomly Selected K/A Reason for Rejection 1/1 Question #1 0008 2.2.38 Not enough information in the facility license to support development of a question for a vapor space accident. Kept 2.2.38 and randomly selected a Tier1/Group 1 plant evolution not previously used. Randomly selected 00009 Small Break LOCA.

1/1 Question

1. 3 000022 2.2.25 The W3 basis did not have enough written information to develop a SRO type question related to the loss of Rx coolant makeup.

Randomly selected a Tier 1/ Group 1 plant evolution and an Equipment Control generic K/A not previously used. Randomly selected 026 (Loss of Component Cooling Water) and 2.2.12 (Knowledge of Surveillance Procedures). 1/1 Question

1. 4 000054 AA2.04 Could not develop or modify a question for a Loss of Main Feedwater Event that would not conflict with the 2015 RO , 2012 , or 2014 written exams. We have already used the allotted number of questions that were previously used in the last two NRC exams. Randomly selected a Tier 1/ Group 1 plant evolution not presently used and a K/A from the A2 field. Randomly selected 038 EA 2.15 1/1 Question #5 000057 AA2.10 W3 has no written guidance for effects on turbine load limiter control upon a loss of vital AC instrument bus.

W3 does not use an automatic system for turbine load control, therefore the turbine load limiter is not mentioned in the offnormal for a loss of Vital AC power.

Randomly selected a different K/A under the same evolution in the A2 field. Randomly selected AA2.19. 1/2 Question

1. 7 000003 AA2.05 W3 has no system or process for interpreting computer in-core TC map for dropped rod location. W3 has a CET map available on our QSPDS system, but this map is not used in any procedure to determine a location of a dropped CEA. Randomly selected another K/A for the same evolution and stayed with the A2 field. Randomly selected AA2.01.

2/1 Question #15 103 A2.05 Could not develop an SRO question for this K/A that would predict the impact of a containment entry. Randomly selected a different Tier 2/Group 1 system not being used and a different K/A while staying in the A2 field. Randomly selected 076 A2.02 Service water header pressure

. 2/2 Question

1. 16 027 A2.01 The CIRS system at W3 is the ARRS system. W3 has no guidance on the ARRS system other than starting and stopping the unit.

To actuate the deluge system, the operator must only operate a pull station (RO knowledge). Randomly selected a different Tier 2/Group 2 system not being used and a different K/A while staying in the A2 field. Randomly selected 0 11 A2.01 Excessive letdown. 2/2 Question

1. 18 086 A2.02 W3 does not have procedura l guidance for responding to low pressure on the fire protection header.

Randomly selected a different Tier 2/Group 2 system not being used and stayed in the A2 field. Randomly selected 033 A2.02 Loss of SFPCS

. 3/1 Question

1. 19 2.1.21 W3 guidance for verifying field controlled procedures is limited, mostly guidance for updating the database.

RO and AO knowledge.

Randomly selected another TIER 3 K/A in the Conduct of Operations section.

Randomly selected 2.1.5 Ability to use procedures related to shift staffing, such as minimum crew complement, overtime limitations, etc.

3/4 Question

1. 24 2.4.8 Could not develop or modify an exam question for this K/A that would not conflict with question s on the previous two written exams and maintain this question TIER 3. Randomly selected another TIER 3 K/A in the Emergency Procedures/Plan section. Randomly selected 2.4.38 Ability to take actions called for in the facility emergency plan, including supporting or acting as emergency coordinator if required

.