Text

2012-010 - Final Written Exam
	ML13003A231
Person / Time
Site:	Waterford
Issue date:	10/01/2012
From:	Vincent Gaddy; Operations Branch IV
To:	; Entergy Operations
	LAURA HURLEY
References
	50-382/OL-12
	Download: ML13003A231 (202)
	v • d • e

2012 NRC Written Examination Waterford 3 Reactor Operator and Senior Reactor Operator (Rev 2)

1. B 26. B

2. C 27. D

3. C 28. D

4. A 29. D

5. B 30. A

6. B 31. A

7. C 32. D

8. C 33. C

9. Qu e s tio n De le te d 34. D

10. D 35. D

11. D 36. A

12. C 37. B

13. A 38. D

14. C 39. B

15. C 40. A

16. A 41. C

17. A 42. A

18. D 43. B

19. A 44. C

20. B 45. D

21. A 46. C

22. C 47. C

23. B 48. A

24. B 49. D

25. A 50. A Page 1 of 2

2012 NRC Written Examination Waterford 3 Reactor Operator and Senior Reactor Operator (Rev 2)

51. B S1. B

52. A S2. B

53. A S3. C

54. A S4. C

55. C S5. D

56. B S6. B

57. B S7. C

58. A S8. C

59. D S9. D

60. D S10. A

61. B S11. A

62. A S12. D

63. B S13. C

64. D S14. D

65. D S15. D

66. C S16. A 2 correct

67. B S17. C D answers

68. A S18. A

69. A S19. B

70. A S20. D

71. D S21. B

72. C S22. A

73. C S23. B

74. D S24. A

75. D S25. A Page 2 of 2

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

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

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

Feedwater Pump A is running with FWCS 1 and FWCS 2 in RTO with both SG levels > 27.4 % NR and slowly rising

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

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

(1) (2)

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

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

Loss of Forced Flow Facility: Waterford 3 Page 1 of 150

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

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

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

B. CORRECT: The Safety Function is correct. The only running RCP (RCP 2B) loses power in this scenario due to the 86STB lockout actuation. The lockout causes a loss of the 1B bus by opening the feeder breaker to the bus from Startup Transformer B. Previously the remaining 3 RCPs were secured for the pressure control malfunction. Natural circulation does not support a loop delta t of < 13 degrees as required by OP-902-001, Reactor Trip Recovery Procedure, Safety Function Status Checklist. The procedure is correct. With no RCPs running, the procedure that covers this is OP-902-003, Loss of Offsite Power/Loss of Forced Flow Recovery Procedure.

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

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

OP-902-001, Reactor Trip Recovery, Rev.013, Technical Reference(s): SFSC OP-902-009, Standard Appendices, Attachment 1, (Attach if not previously provided) Rev. 307 (including version/revision number)

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

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

Modified Bank # (Note changes or attach parent)

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

Modified explanation B. Updated Technical References.

Facility: Waterford 3 Page 2 of 150

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

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

Proposed Question: RO 2 Rev: 1 Given:

Plant experienced an inadvertent opening of a Pressurizer relief valve

Crew has entered OP-902-002, Loss of Coolant Accident Recovery One hour into the event, the following indications exist:

Pressurizer level is 100% and stable

Pressurizer pressure is 1450 PSIA and stable

Representative CET temperature is 595°F and slowly rising

HJTC/UHJTC T indicates ~ 350°F at all levels on QSPDS 1 and 2

High Pressure Safety Injection (HPSI) Pump B tripped on startup

HPSI Cold Leg Injection flows are 0 gpm Core cooling is ____(1)____ because ________(2)________.

(1) (2)

A. adequate Pressurizer level is above HPSI throttle criteria B. adequate HPSI flow curve is met C. inadequate CET temperatures indicate superheat conditions D. inadequate only one HPSI pump is running Facility: Waterford 3 Page 3 of 150

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

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

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

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

D. Incorrect. One HPSI Pump being available is not the reason for the low flow condition, high pressure in the RCS is.

Technical Reference(s): Steam tables OP-902-009, Standard Appendices, Attachment (Attach if not previously provided) 2E & 2F, revision 307 OP-902-002, Loss of Coolant Accident Recovery, (including version/revision number) revision 14 Proposed references to be provided to applicants during examination: Steam Tables Learning Objective: WLP-OPS-PPE02 Obj. 8 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

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

Minor editorial changes, modified 4th bullet in stem to make applicant determine if vessel levels are voided. Added attachment numbers to OP-902-009 reference.

Facility: Waterford 3 Page 4 of 150

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

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

Proposed Question: RO 3 Rev: 1 Given:

Small Break LOCA occurred one hour ago

Pressurizer level is 0%

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

(1) (2)

A. 1 only 20 B. 1 only 80 C. 1 or 2 20 D. 1 or 2 80 Facility: Waterford 3 Page 5 of 150

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

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

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

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

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

PZR level was given in the stem.

D. Incorrect. Wrong plenum level requirement.

OP-902-002, Loss of Coolant Accident Recovery, Technical Reference(s): Rev. 15 (Attach if not previously provided) OP-903-013, Monthly Channel Checks, Rev. 16 (including version/revision number)

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

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

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

Minor editorial changes Facility: Waterford 3 Page 6 of 150

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

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

Proposed Question: RO 4 Rev: 1 Given:

Plant is at 100% power

Loss of offsite power has occurred

Crew has entered OP-902-003, Loss of Offsite Power/Loss of Forced Circulation Recovery Verification of single phase natural circulation cannot be accomplished until approximately ____(1)____ minutes following the loss of offsite power. RCS subcooling will be verified using __(2)__ subcooling monitors.

(1) (2)

A. 5 to 15 CET B. 20 to 30 CET C. 20 to 30 Thot D. 5 to 15 Thot Facility: Waterford 3 Page 7 of 150

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

A. CORRECT: Per OI-038-000, Verification of natural circulation cannot be accomplished until approximately 5-15 minutes following the action. Guidance is also provided in OI-038-000 that while on natural circulation, CETs will be used to verify subcooling.

B. Incorrect. Per OI-038-000, Verification of natural circulation cannot be accomplished until approximately 5-15 minutes following the action.

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

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

OI-038-000, Emergency Operating Procedures Technical Reference(s): Operations Expectation/Guidance, Rev. 5 (Attach if not previously provided)

(including version/revision number)

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

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

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

Changed from steam flow and feed flow changes to just verification of natural circulation flow setup after LOOP. Minor editorial changes.

Facility: Waterford 3 Page 8 of 150

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

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

Proposed Question: RO 5 Rev: 0 Given:

Plant is at 100% power

A loss of all Charging Pumps has occurred due to gas intrusion The crew will implement ____(1)____. If normal Charging flow can not be restored, a

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

(1) (2)

A. OP-901-110, Pressurizer Level Low Pressure Safety Injection Pump Control Malfunction B. OP-901-112, Charging or High Pressure Safety Injection Pump Letdown Malfunction C. OP-901-110, Pressurizer Level High Pressure Safety Injection Pump Control Malfunction D. OP-901-112, Charging or Low Pressure Safety Injection Pump Letdown Malfunction Facility: Waterford 3 Page 9 of 150

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

A. Incorrect. OP-901-112 Section E3 contains the guidance for gas bound charging pumps. OP-901-112 directs the crew to depressurize the RCS and start a HPSI pump (not a LPSI pump) to restore pressurizer level.

B. CORRECT: OP-901-112 Section E3 contains the guidance for gas bound charging pumps. OP-901-112 directs the crew to depressurize the RCS and start a HPSI pump to restore pressurizer level.

C. Incorrect. OP-901-112 Section E3 contains the guidance for gas bound charging pumps. Part 2 is correct.

D. Incorrect. Part 1 is correct . OP-901-112 directs the crew to depressurize the RCS and start a HPSI pump (not a LPSI pump) to restore pressurizer level OP-901-112, Charging or Letdown Malfunction, Technical Reference(s): Rev. 4 (Attach if not previously provided)

(including version/revision number)

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

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

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

Rewrote the question to require the applicant to determine OP-901-112 entry conditions.

Facility: Waterford 3 Page 10 of 150

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

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

The plant is in MODE 5

The RCS was drained to 14.5 feet MSL

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

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

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

(1) (2)

A. A 1 B. A 2 C. B 1 D. B 2 Facility: Waterford 3 Page 11 of 150

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

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

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

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

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

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

OP-901-131, Shutdown Cooling Malfunction, Rev.

Technical Reference(s): 302 (Attach if not previously provided)

(including version/revision number)

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

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

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

Minor Reformatting. Added procedure title to Tech Ref.

Facility: Waterford 3 Page 12 of 150

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

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

Plant has tripped from 100% power

RCS pressure is 1680 PSIA and slowly lowering

Containment pressure is 16.8 PSIA and slowly rising

CCW Trains A and B were in a normal MODE 1 alignment prior to the trip Currently, CC-200 A and CC-727, CCW A to AB Header Isolation valves,

_____(1)_____. This allows __________________(2)_________________ with no further operator action.

(1) (2)

A. are closed CCW Train B to supply cooling to the Reactor Coolant Pumps B. are closed CCW Train B to supply cooling to the Fuel Pool Heat Exchangers C. remain open CCW Train A to supply cooling to the Reactor Coolant Pumps D. remain open CCW Train A to supply cooling to the Fuel Pool Heat Exchangers Facility: Waterford 3 Page 13 of 150

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

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

In this condition only a SIAS is present. The Train B to AB isolations close on a SIAS. Therefore the Reactor Coolant Pumps are being supplied from Train A of CCW.

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

CC-620, FPHX TCV closed due to a SIAS and closure of the Train B to AB isolation valves. Operator action to override CC-620 would have to be taken to supply cooling to the Fuel Pool Heat Exchangers.

C. CORRECT: CC-200A and CC-727 remain open, these two valves close on a CSAS.

In this condition only a SIAS is present. Therefore the Reactor Coolant Pumps are being being supplied from Train A of CCW.

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

CC-620, FPHX TCV closed due to a SIAS and closure of the Train B to AB isolation valves. Operator action to override CC-620 would have to be taken to supply cooling to the Fuel Pool Heat Exchangers.

OP-902-009, Standard Appendices, Appendix 4, Technical Reference(s): Rev. 307 (Attach if not previously provided) SD-CC, Rev. 18 (including version/revision number)

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

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

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

Rewrote in 2/2 form with modification to second part to prevent incredible distractors.

Modified explanations accordingly.

Facility: Waterford 3 Page 14 of 150

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 000027 AK2.03 Importance Rating 2.6 K/A Statement Knowledge of the interrelations between the Pressurizer Pressure Control Malfunctions and the following: Controllers and positioners Proposed Question: RO 8 Rev: 2 Given:

Plant is at 100% power

Pressurizer Spray Controller fails to 100% output

Crew has entered OP-901-120, Pressurizer Pressure Control Malfunction Based on this failure, the power supplied to the Proportional Heaters will go to (1) and the selected Pressurizer Spray valves will be (2) .

(1) (2)

A. zero open B. zero closed C. maximum open D. maximum closed Facility: Waterford 3 Page 15 of 150

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

A. Incorrect. Actual PZR pressure would be dropping with the Spray vales open. The Proportional Heaters receive their signal from the Pressurizer Pressure controller, not the Pressurizer Spray controller. Part 2 is correct.

B. Incorrect. With the Pressurizer Spray controller failed at 100% output, the Spray valves would be open.

C. CORRECT: Actual PZR pressure would be dropping with the Spray vales open.

The Proportional Heaters receive their signal from the Pressurizer Pressure controller, not the Pressurizer Spray controller and would therefore be a full power.

With the Pressurizer Spray controller failed at 100% output, the Spray valves would be open.

D. Incorrect. Part 1 is correct. With the Pressurizer Spray controller failed at 100%

output, the Spray valves would be open.

Technical Reference(s): OP-901-120, Pressurizer Pressure Control (Attach if not previously provided) Malfunction, Rev. 301 SD-PLC, Pressurizer Level and Pressure Control, (including version/revision number) Rev 8 Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PP010 obj. 3 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

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

Rev 1, Changed first part of C and D to remain closed from require manual opening.

Capitalization and grammatical changes per generic comments.

Rev 2, changed focus of question from Pressure controller to Spray controller to improve distractors A & B.

Facility: Waterford 3 Page 16 of 150

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 000029 EA1.12 Importance Rating 4.1 K/A Statement Ability to operate and monitor the following as they apply to a ATWS: M/G set power supply and reactor trip breakers Proposed Question: RO 9 Rev: 2 Given:

Plant is at 100% power

An event occurs and automatic reactor trips fail

Manual Reactor trip at CP-2 fails

Manual Diverse Reactor Trip is successful Which of the following identifies the current status of the Reactor Trip breakers and the CEDM MG Set Supply breakers on the 32A(B) buses?

CEDM MG Set Reactor Trip breakers Supply breakers A. Open Open B. Open Closed C. Closed Open D. Closed Closed Facility: Waterford 3 Page 17 of 150

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

A. Incorrect - both parts. Stem states auto and manual Reactor trip failed, therefore Reactor Trip breakers failed to open. Stem also states DRT is successful, therefore CEDM MG set output breakers would be open but power supply breakers remained closed.

B. Incorrect - first part. Stem states auto and manual Reactor trip failed, therefore Reactor Trip breakers failed to open. CEDM MG set power supply breakers remained closed.

C. Incorrect - second part. Stem states auto and manual Reactor trip failed, therefore Reactor Trip breakers failed to open. Stem also states Diverse Reactor Trip is successful, therefore MG set output breakers are open but power supply breakers remained closed.

D. CORRECT: Stem states auto and manual Reactor trip failed, therefore Reactor Trip breakers failed to open. Stem also states DRT is successful, therefore CEDM MG set output breakers would be open but power supply breakers remained closed.

Technical Reference(s): SD-CED, Rev. 9 (Attach if not previously provided) SD-PPS, Rev. 13 (including version/revision number)

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

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

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

Rev 1, Question replaced with NRC supplied question Rev 2, typographical and editorial for clarity.

Facility: Waterford 3 Page 18 of 150

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

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

Proposed Question: RO 10 Rev: 2 Given:

Steam Generator tube rupture event is in progress involving both Steam Generators

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

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

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

Both Startup Feedwater Reg Valves are ~ 17% open

Chemistry has reported the following activity samples:

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

(1) (2)

A. 2 only 1 B. 1 and 2 1 C. 2 only 2 D. 1 and 2 2 Facility: Waterford 3 Page 19 of 150

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

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

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

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

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

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

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

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

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

Rev 1, changes per NRC comments. Added manual reactor trip to second bullet, minor reformatting.

Rev 2, typographical and editorial for clarity.

Facility: Waterford 3 Page 20 of 150

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

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

Proposed Question: RO 11 Rev: 2 Given:

Excess Steam Demand event has occurred

Crew has entered OP-902-004, Excess Steam Demand Recovery, and is performing step to Stabilize RCS Temperature

BOP has stabilized RCS Tcold using the unaffected Steam Generator ADV and EFW The ATC will then stabilize RCS pressure ____(1)____ HPSI shutoff head and adjust RCS pressure such that the differential pressure across the Steam Generator tubes does not exceed the design differential pressure of _____(2)________.

(1) (2)

A. above 1400 psid B. below 1400 psid C. below 1600 psid D. above 1600 psid Facility: Waterford 3 Page 21 of 150

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

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

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

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

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

OP-902-004, Excess Steam Demand Recovery, Technical Reference(s): Rev. 13 TGOP-902-004, Technical Guide for Excess (Attach if not previously provided) Steam Demand Recovery, Rev. 303 (including version/revision number)

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

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

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

Rev 1, revised per examiner comments, including generic comments.

Rev 2, typographical.

Facility: Waterford 3 Page 22 of 150

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

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

Proposed Question: RO 12 Rev: 2 Given:

Reactor has tripped due to a loss of Main Feedwater

Crew has entered OP-902-006, Loss of Main Feedwater Recovery

Condensate Storage Pool (CSP) level is 30% and lowering with CMU-141, CSP LCV Bypass, open When CSP level reaches ____(1)_____ %, the crew should commence alignment of Emergency Feedwater Pump suctions to _______(2)______ train of Auxiliary Component Cooling Water (ACCW).

(1) (2)

A. 11 one operating B. 11 their respective C. 25 one operating D. 25 their respective Facility: Waterford 3 Page 23 of 150

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

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

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

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

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

OP-902-006, Loss of Main Feedwater Recovery, Technical Reference(s): Rev. 13 OP-902-009, Standard Appendices, Appendix 10, (Attach if not previously provided) Rev. 307 (including version/revision number)

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

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

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

Rev 1, removed train of ACCW from each second blank answer. Minor reformatting.

Rev 2, typographical and editorial for clarity.

Facility: Waterford 3 Page 24 of 150

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 000055 EK3.02 Importance Rating 4.3 K/A Statement Knowledge of the reasons for the following responses as the apply to the Station Blackout: Actions contained in EOP for loss of offsite and onsite power Proposed Question: RO 13 Rev: 2 Given:

Station Blackout has occurred

Crew is performing actions contained in OP-902-005, Station Blackout Recovery

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

_____(2)________.

(1) (2)

A. 400°F Shutdown Margin requirements are met B. 520°F Shutdown Margin requirements are met C. 400°F RCS subcooled margin requirements are not exceeded D. 520°F RCS subcooled margin requirements are not exceeded Facility: Waterford 3 Page 25 of 150

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

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

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

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

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

Technical Reference(s): OP-902-005, Station Blackout Recovery, Rev. 15 TG-OP-902-005, Technical Guide for Station (Attach if not previously provided) Blackout Recovery, Rev. 304 (including version/revision number)

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

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

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

Rev 1, revised per NRC comments, including generic formatting comments.

Rev 2, typographical and editorial for clarity.

Facility: Waterford 3 Page 26 of 150

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 000056 AA2.46 Importance Rating 4.2 K/A Statement Ability to determine and interpret the following as they apply to the Loss of Offsite Power: That the ED/Gs have started automatically and that the bus tie breakers are closed Proposed Question: RO 14 Rev: 2 Given:

Plant was at 100% power when a loss of offsite power occurred The BOP is performing verification of the Maintenance of Vital Auxiliaries and observes the following for EDG B:

EDG B frequency is 59.8 Hertz

EDG B output breaker is open

EDG B voltage is 3910 AC volts In accordance with OP-902-000, Standard Post Trip Actions, which of the following should the BOP perform?

A. Manually close EDG B output breaker.

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

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

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

Facility: Waterford 3 Page 27 of 150

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

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

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

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

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

Technical Reference(s): OP-902-000, Standard Post Trip Actions, Rev. 13 (Attach if not previously provided)

(including version/revision number)

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

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

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

Rev 1, added In accordance with OP-902-000, Standard Post Trip Actions to stem and combined first and second bullet into one bullet.

Rev 2, typographical and editorial for clarity.

Facility: Waterford 3 Page 28 of 150

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 000057 AA2.17 Importance Rating 3.1 K/A Statement Ability to determine and interpret the following as they apply to the Loss of Vital AC Instrument Bus: System and component status, using local or remote controls Proposed Question: RO 15 Rev: 2 Given:

Plant is at 100% power

Loss of Instrument SUPS MA has occurred The ATC will report to the CRS that Reactor Trip Breakers 1, 5, ________(1)_______

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

(1) (2)

A. 2 and 6 trip B. 4 and 8 trip C. 2 and 6 remain at 100% power D. 4 and 8 remain at 100% power Facility: Waterford 3 Page 29 of 150

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

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

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

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

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

OP-901-312, Loss of Vital Instrument Bus, Rev.

Technical Reference(s): 306 (Attach if not previously provided) SD-PPS, Fig. 05, Rev. 1 (including version/revision number)

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

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

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

Rev 1, minor reformatting. Added SP-PPS technical reference.

Rev 2, typographical and editorial for clarity.

Facility: Waterford 3 Page 30 of 150

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

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

CVC-101, Letdown Stop Valve

CVC-109, Letdown Outside Containment Isolation

CVC-510, Volume Control Tank Makeup All 3 components indicate closed on the PMC.

Which SUPS AC output breaker has tripped open?

A. SA B. SB C. SAB D. SMD Facility: Waterford 3 Page 31 of 150

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

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

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

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

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

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

(including version/revision number)

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

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

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

Corrected CVC-109 title, eliminated unnecessary verbiage from stem.

Facility: Waterford 3 Page 32 of 150

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

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

Proposed Question: RO 17 Rev: 1 Given:

Plant has been manually tripped due to a loss of instrument air

Crew has entered OP-901-511, Instrument Air Malfunction and OP-902-001, Reactor Trip Recovery The Essential Air System will be aligned to service ____(1)____ during a loss of instrument air. Essential Air system is required to ensure _____(2)_____ is maintained.

(1) (2)

A. after 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months containment integrity B. after 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months ECCS functionality C. within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months containment integrity D. within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months ECCS functionality Facility: Waterford 3 Page 33 of 150

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

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

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

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

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

Technical Reference(s): OP-901-511, Instrument Air Malfunction, 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-AIR00 Obj. 4 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

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

Changed first part of distractors C and D from immediately to within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . Minor reformatting.

Facility: Waterford 3 Page 34 of 150

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

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

Reactor power is 100%

BOP operator has been directed to raise Main Generator voltage to address grid instability issues Reactive load (MVAR) is administratively limited to __(1)__ MVAR out. Excessive over-excitation is a major concern due to increased chances of _______(2)______.

(1) (2)

A. 75 pole slippage B. 75 high rotor field winding temperature C. 400 pole slippage D. 400 high rotor field winding temperature Facility: Waterford 3 Page 35 of 150

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

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

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

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

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

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

Technical Reference(s): OP-010-004, Power Operation, Rev. 315 (Attach if not previously provided) WLP-OPS-GEN00, Rev. 8 (including version/revision number)

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

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 5,10 55.43 Comments: Minor reformatting. Added WLP-OPS-Gen00 as a technical reference.

Facility: Waterford 3 Page 36 of 150

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

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

Plant is at 5% power

Group P CEAs are 85 inches withdrawn

All other CEAs are at the Upper Electrical Limit

ATC withdraws Group P CEAs 3 steps and releases the CEA Manual Shim switch

CEA Group P CEAs continued stepping out Which of the following is performed FIRST per OP-901-102, CEA or CEDMCS Malfunction?

A. Place the CEDMCS Mode Select switch to OFF.

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

C. Trip the Reactor and GO TO OP-902-000, Standard Post Trip Actions.

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

Facility: Waterford 3 Page 37 of 150

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

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

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

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

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

OP-901-102, CEA or CEDMCS Malfunction, Technical Reference(s): Rev. 5 (Attach if not previously provided)

(including version/revision number)

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

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

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

Replaced rods with CEAs in fourth bullet, made CEA plural in fifth bullet, minor reformatting. Changed original D distractor per NRC comment, then rearranged selections shortest to longest.

Facility: Waterford 3 Page 38 of 150

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

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

Valves Proposed Question: RO 20 Rev: 1 Given:

Plant is in MODE 3

Emergency Diesel Generator A is Danger Tagged out

Plant has experienced a steam line leak resulting in an uncontrolled cooldown

Loss of power to the 3A safety bus has occurred

CRS directs ATC operator to commence emergency boration Which of the following will be the emergency boration flow path aligned by the ATC?

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

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

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

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

Facility: Waterford 3 Page 39 of 150

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

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

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

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

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

< 15%.

Technical Reference(s): OP-901-103, Emergency Boration, Rev. 2 OP-002-005, Chemical And Volume Control (Attach if not previously provided) System, Rev. 37 (including version/revision number)

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

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 6,10 55.43 Comments: Added first bullet, removed mode from second bullet, removed due to uncontrolled cooldown from final bullet, rearranged selections from shortest to longest and rearranged explanations. Changed correct answer to B. Added OP-002-005 as a technical reference.

Facility: Waterford 3 Page 40 of 150

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

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

Plant is operating at 100% power

Charging Pumps A and B are running for additional RCS clean-up

Pressurizer Level Control CHANNEL SELECT switch is selected to Channel X

RC-ILT-0110X has failed low Assuming no operator actions are taken, Pressurizer pressure will INITIALLY:

A. rise due to backup charging pump starting.

B. lower due to pressurizer spray valve opening.

C. lower due to backup charging pump securing.

D. rise due to pressurizer backup heaters energizing.

Facility: Waterford 3 Page 41 of 150

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

A. CORRECT: Pressurizer pressure will initially rise due to the Pressurizer level instrument failing low. The PLCS will see a low level in the pressurizer and start backup Charging Pump AB. This will cause Pzr level to rise with a corresponding rise in pressure.

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

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

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

OP-901-110, Pressurizer Level Control Technical Reference(s): Malfunction, Rev. 6 (Attach if not previously provided)

(including version/revision number)

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

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

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

Rev 1, revised for editorial comments and re-ordered answers from shortest to longest.

Rev 2, typographical and editorial for clarity.

Facility: Waterford 3 Page 42 of 150

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

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

Core reload is in progress ATC observes Startup channel counts have risen from 10 CPS to 50 CPS and rising Startup Rate is 0.2 DPM and stable Indications available to the ATC operator to monitor for criticality are ____(1)____. The conditions given indicate the reactor is _____(2)_____.

(1) (2)

A. Startup channels and Startup Rate NOT critical B. Startup and Control channels NOT critical C. Startup channels and Startup Rate critical D. Startup and Control channels critical Facility: Waterford 3 Page 43 of 150

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

A. Incorrect. Correct indications. Evaluation of criticality is wrong.

B. Incorrect. The control channels are not available to the operators in refueling conditions. Evaluation of criticality is wrong.

C. CORRECT: Startup channels and startup rate are available to the operators at this power level. Startup channel counts steadily rising and SUR positive and sustained are indications the reactor is critical.

D. Incorrect. The control channels are not available to the operators in refueling conditions. Evaluation of criticality is correct.

Technical Reference(s): OP-010-003, Plant Startup, Rev. 324 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-ENI00 obj. 4 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 1, 2 55.43 Comments:

Rev 1, revised NRC comments to give indications of criticality and changed boron dilution monitor alarms to startup rate.

Rev 2, typographical and editorial for clarity.

Facility: Waterford 3 Page 44 of 150

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

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

Proposed Question: RO 23 Rev: 1 Given:

Plant was initially at 100% power

Condenser vacuum is 20.5 INHG and lowering slowly

Crew has entered OP-901-220, Loss of Condenser Vacuum, and OP-901-212, Rapid Plant Power Reduction Which of the following is the appropriate action for this event?

A. Trip the Main Turbine and then trip the Reactor.

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

C. Continue the rapid Plant downpower until vacuum recovers.

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

Facility: Waterford 3 Page 45 of 150

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

A. Tripping the Main Turbine first could challenge the integrity of the condenser by supplying high energy steam directly to the condenser via Steam Bypass valves.

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

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

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

OP-901-220, Loss of Condenser Vacuum, Rev.

Technical Reference(s): 302 (Attach if not previously provided)

(including version/revision number)

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

Question Source: Bank # X 03965a Modified Bank # (Note changes or attach parent)

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

Corrected procedure title in third bullet, reordered selections short to long, changed answer to B, reordered explanations, minor reformatting.

Facility: Waterford 3 Page 46 of 150

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

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

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

(1) (2)

A. manual Reactor trip Safety Injection Actuation Signal B. manual Reactor trip Main Steam Isolation Signal C. rapid Plant shutdown Safety Injection Actuation Signal D. rapid Plant shutdown Main Steam Isolation Signal Facility: Waterford 3 Page 47 of 150

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

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

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

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

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

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

Technical Reference(s): OP-901-503, Isolation Panel Fire, 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-PP050 obj. 3 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Comments: Minor reformatting Facility: Waterford 3 Page 48 of 150

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 2 K/A # 000068 AK3.12 Importance Rating 4.1 K/A Statement Knowledge of the reasons for the following responses as they apply to the Control Room Evacuation: Required sequence of actions for emergency evacuation of control room Proposed Question: RO 25 Rev: 2 Given:

The control room has been evacuated in accordance with OP-901-502, Evacuation of Control Room and Subsequent Plant Shutdown, due to a fire at CP-8

Controls have been established at LCP-43

Pressurizer pressure is 690 PSIA.

To close SI-331B, Safety Injection Tank 1B Outlet Isol, the crew must first bypass the SIT 1B Isol Press Bypass Close Permissive switch.

This SIT 1B Isol Press Bypass Close Permissive switch is located at

________(1)_______. and is required to be manipulated before closing SI-331B from LCP-43 because the _____(2)_____ .

(1) (2)

A. SI-331B breaker SITs cannot be depressurized from LCP-43 B. LCP-43 SITs cannot be depressurized from LCP-43 C. SI-331B breaker SIT Outlet Valve close permissive is lower when transferred to LCP-43 D. LCP-43 SIT Outlet Valve close permissive is lower when transferred to LCP-43 Facility: Waterford 3 Page 49 of 150

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

A. CORRECT: The Bypass switch is located on the breaker for the valve .SI-331A will not close until RCS pressure is < 415 PSIA due to an interlock. Since there are no vent valves for the SITs at LCP-43, the interlock must be defeated to close SI-331A.

B. Incorrect. Part 1 in wrong. Part 2 is correct.

C. Incorrect. Part 1 is correct. Part 2 is wrong. Part 2 is credible because Safety Injection Interlocks are altered when transferred to LCP-43 (Auto open interlock)

D. Incorrect. Parts 1 and 2 are wrong.

OP-901-502, Evacuation of Control Room and Technical Reference(s): Subsequent Plant Shutdown, Revision 21 (Attach if not previously provided) WLP-OPS-SI00 Rev. 21 slide 121 and 122 (including version/revision number)

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

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

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

Rev 1, changed the question entirely due to the original question having a K/A mismatch.

Rev 2, revised column 2 to remove setpoints and give reason for sequence.

Facility: Waterford 3 Page 50 of 150

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

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

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

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

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

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

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

Facility: Waterford 3 Page 51 of 150

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

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

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

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

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

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

(including version/revision number)

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

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

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

Facility: Waterford 3 Page 52 of 150

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

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

Proposed Question: RO 27 Rev: 1 Given:

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

Pressurizer pressure is 2250 PSIA and slowly rising

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

Thot Loop1 is 568°F and stable

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

Thot Loop 2 is 565°F and stable

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

(1) (2)

A. NOT met 1 B. met 1 C. NOT met 2 D. met 2 Facility: Waterford 3 Page 53 of 150

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

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

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

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

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

OP-902-003, Loss of Offsite Power/Loss of Forced Technical Reference(s): Flow Recovery, Rev. 7 TGOP-902-003, Technical Guide for Loss of Offsite Power/Loss of Forced Flow Recovery, Rev.

(Attach if not previously provided) 303 (including version/revision number)

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

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

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

Minor reformatting. Added TGOP-902-003 as a Technical Reference.

Facility: Waterford 3 Page 54 of 150

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

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

Plant is at 100% power

RCP 1A shaft shears RCP amps would indicate ____(1)____ than normal and a reactor trip would occur due to _____(2)_____.

(1) (2)

A. higher Reactor Coolant Flow Low B. higher DNBR Low C. lower DNBR Low D. lower Reactor Coolant Flow Low Facility: Waterford 3 Page 55 of 150

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

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

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

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

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

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

(including version/revision number)

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

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

Modified Bank # (Note changes or attach parent)

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

Rev 1, removed third bullet, changed steam generator flow to reactor coolant flow in second set of selections.

Rev 2, typographical and editorial for clarity.

Facility: Waterford 3 Page 56 of 150

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

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

Solid Plant operations are in progress

RCS temperature is 120°F and stable

Letdown Back Pressure Controller is operating in AUTO

CCW flow is reduced to the in-service SDC Heat Exchanger RCS pressure will ____(1)____ until the Letdown Back Pressure Control Valve throttles

_____(2)____.

(1) (2)

A. lower closed B. rise closed C. lower open D. rise open Facility: Waterford 3 Page 57 of 150

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

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

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

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

D. CORRECT: Per OP-010-005, when setting up solid operation the Letdown Backpressure valves are set to control pressure at ~ 100 psig in AUTO. Lowering CCW flow to the in-service SDC Heat Exchanger when RCS temperature is initially stable will cause temperature and RCS pressure to rise. To restore RCS pressure to setpoint the letdown backpressure valve must open to raise letdown flow to lower RCS pressure to setpoint.

Technical Reference(s): OP-010-005, Plant Shutdown, Rev. 317 (Attach if not previously provided)

(including version/revision number)

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

Question Source: Bank # X 08405a Modified Bank # (Note changes or attach parent)

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

Revised per comments. Revised explanation for correct response to better discuss what is happening within the system.

Facility: Waterford 3 Page 58 of 150

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

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

Plant has experienced a LOCA event

Crew has entered OP-902-002, Loss of Coolant Accident Recovery

Recirculation Actuation Signal (RAS) has just occurred

Low Pressure Safety Injection (LPSI) Pump A failed to automatically or manually trip from the control room LPSI flow from the SI sump must be secured within _____(1)_____ minutes of the SI Pumps Recirc Isolation valves closing. An operator should be dispatched to switchgear ____(2)____ to locally trip LPSI Pump A.

(1) (2)

A. 5 3A B. 5 31A C. 2 3A D. 2 31A Facility: Waterford 3 Page 59 of 150

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

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

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

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

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

OP-902-002 Loss of Coolant Accident Recovery, Technical Reference(s): Rev. 15 TGOP-902-002 , Technical Guide for Loss of (Attach if not previously provided) Coolant Accident Recovery, Rev.14 (including version/revision number) SD-SI, Rev. 14 Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PPE02 obj. 19 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

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

Rev 1, updated Technical References, minor reformatting of question stem and answers.

Rev 2, typographical and editorial for clarity.

Facility: Waterford 3 Page 60 of 150

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

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

Plant has experienced an Excess Steam Demand event

Crew has entered OP-902-004, Excess Steam Demand Recovery

CRS has directed the ATC to reset Safety Injection and Containment Isolation in accordance with OP-902-009,Standard Appendices, Attachment 5-A: SIAS and CIAS Pressurizer Pressure Reset Procedure

Channel A and Channel B LO PZR PRESS trips have been reset To complete the reset of the Safety Injection and Containment Isolation signals, the BOP will:

A. Bypass Channel C LO PZR PRESS trip, reset the Initiation relays and then reset the Actuation relays for SIAS and CIAS.

B. Bypass Channel C LO PZR PRESS trip, reset the Actuation relays and then reset the Initiation relays for SIAS and CIAS.

C. Bypass Channel C and Channel D LO PZR PRESS trips, reset the Initiation relays and then reset the Actuation relays for SIAS and CIAS.

D. Bypass Channel C and Channel D LO PZR PRESS trips, reset the Actuation relays and then reset the Initiation relays for SIAS and CIAS.

Facility: Waterford 3 Page 61 of 150

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

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

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

C. Incorrect. Only Channel C is bypassed before the initiation and actuation relays are reset.

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

OP-902-009, Standard Appendices, Attachment Technical Reference(s): 5-A, Rev. 307 (Attach if not previously provided)

(including version/revision number)

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

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

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

Eliminated unnecessary info from stem, added bistable title to stem and answers, reordered answers short to long, changed proposed answer and reordered explanations.

Facility: Waterford 3 Page 62 of 150

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

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

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

RCS pressure is 600 PSIA

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

Cold leg injection header flow rates are as follows:

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

Fifteen minutes into the event, RVLMS indicates core uncovery and CETs are showing indications of superheat Based on this information, the crew should ____(1)____ because

________(2)________.

(1) (2)

A. Remain in OP-902-002, Core uncovery and superheated conditions LOCA Recovery are expected B. Remain in OP-902-002, the SI flow curves are being met LOCA Recovery C. Exit to OP-902-008, Core uncovery and superheated conditions Functional Recovery are NOT expected D. Exit to OP-902-008, the SI flow curves are NOT being met Functional Recovery Facility: Waterford 3 Page 63 of 150

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

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

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

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

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

OP-902-002, Loss of Coolant Accident Recovery, Technical Reference(s): Rev 15, OP-902-009, Standard Appendices, Attachment (Attach if not previously provided) 2-E, Rev. 307 (including version/revision number)

Proposed references to be provided to applicants during examination: OP-902-009 Appendix 2, Att. 2-E Learning Objective: WLP-OPS-PPE02 obj. 7 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

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

Rev 1, revised per examiner's comments.

Rev 2, typographical and editorial for clarity.

Facility: Waterford 3 Page 64 of 150

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

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

Maintaining quench tank pressure Proposed Question: RO 33 Rev: 1 During a LOCA, venting the vessel head using the RCS vent valves will ____(1)____ in the vessel head, and should be stopped before reaching ____(2)____ in the quench tank to prevent rupturing the quench tank rupture disc.

(1) (2)

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

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

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

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

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

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

OP-902-002, Loss of Coolant Accident Recovery, Technical Reference(s): Rev. 15 OP-902-009, standard Appendices, Appendix 11, (Attach if not previously provided) Rev. 307 (including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-RCS00 obj. 2 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 3, 8 55.43 Comments:

Revised per NRC comments.

Facility: Waterford 3 Page 66 of 150

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

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

(1) (2)

A. Condensate Storage Pool Demineralized Water Storage Tank B. Condensate Storage Pool Condensate Storage Pool C. Demineralized Water Storage Tank Condensate Storage Pool D. Demineralized Water Storage Tank Demineralized Water Storage Tank Facility: Waterford 3 Page 67 of 150

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

A. Incorrect: Wrong makeup source for WCTs. Correct Makeup source for CCW Surge Tank.

B. Incorrect: Wrong Makeup source to the WCTs and wrong source to the CCW Surge Tank. The CSP is the backup makeup source to the CCW Surge Tank.

C. Incorrect. The normal source of makeup to the WCTs is the DWST; however, the CSP is not the normal makeup source to the CCW Surge Tank.

D. CORRECT: The normal source of makeup to the WCTs and the CCW Surge Tank is the Demineralized Water Storage Tank.

Technical Reference(s): SD-CMU, Fig. 2, Rev. 7 (Attach if not previously provided) SD-CC, Fig. 27, Rev. 18 (including version/revision number)

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

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

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

Changed per NRC comments and rewrote explanations. Added SD-CC as a Technical Reference.

Facility: Waterford 3 Page 68 of 150

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

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

Plant is operating at 100% power

Crew is performing OP-903-118, Primary Quarterly IST Valve Tests, for Component Cooling Water valves The BOP closes CC-200A, CCW Header A to AB Supply Isolation, for a stroke time.

CC-963A, Shutdown Heat Exchanger A CCW Flow Control, will open and ____(1)____.

CC-963A, Shutdown Heat Exchanger A CCW Flow Control, opens to _____(2)_____.

(1) (2)

A. remains open maintain minimum flow requirements through the SDC Heat Exchanger B. remains open reduce pressure transients on the CCW system C. will auto close maintain minimum flow requirements through the SDC Heat Exchanger D. will auto close reduce pressure transients on the CCW system Facility: Waterford 3 Page 69 of 150

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

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

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

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

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

OP-903-118, Primary Quarterly IST Valve Tests, Technical Reference(s): Rev. 27 (Attach if not previously provided) SD-CC, Rev. 17 (including version/revision number)

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

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

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

Rev. 1 Changed first part of A and B must be manually closed to remains open. Minor editorial and format changes.

Rev. 2, editorial adjustment of stem.

Facility: Waterford 3 Page 70 of 150

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

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

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

A. CPC Pressurizer Pressure - Low, Aux Trip B. CPC Pressurizer Pressure - High, Aux Trip C. RPS Pressurizer Pressure - Low D. RPS Pressurizer Pressure - High Facility: Waterford 3 Page 71 of 150

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

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

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

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

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

Technical Reference(s): TS 2.0 Bases, DNBR-Low (Attach if not previously provided) SD-PPS Rev. 12 (including version/revision number)

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

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

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

Changed names of trips to reflect TS Section 2.0 and Bases. Minor format changes.

Facility: Waterford 3 Page 72 of 150

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 012 A4.03 Importance Rating 3.6 K/A Statement A4.03 Ability to manually operate and/or monitor in the control room: Channel blocks and bypasses Proposed Question: RO 37 Rev: 3 Following a Reactor trip from 100% power, which of the following Operating Bypasses will automatically change state with no operator action?

A. CPC Trip Bypass B. High Log Power Trip Bypass C. Reactor Coolant Flow Low Trip Bypass D. High Steam Generator Level Trip Bypass Facility: Waterford 3 Page 73 of 150

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

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

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

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

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

Technical Reference(s): OP-009-007, Plant Protection System, Rev. 15 (Attach if not previously provided) SD-PPS, Rev. 14 (including version/revision number)

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

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

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

Rev. 1, revised per NRC comments, corrected operating bypass names.

Rev 2, typographical and editorial for clarity.

Rev 3, enhanced for clarity.

Facility: Waterford 3 Page 74 of 150

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

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

HPSI Pump A is tagged out and HPSI pump AB is NOT aligned to replace HPSI Pump A

A LOCA has occurred

RCS Pressure stabilizes at 750 PSIA

Charging Pumps A and B are running

HPSI Pump B failed to start With NO operator action, which of the following describes the effect on the fuel assemblies?

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

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

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

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

Facility: Waterford 3 Page 75 of 150

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

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

B. Incorrect. Fuel failure will occur. Safety Injection Tanks will not dump until 600-650 PSIA.

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

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

OP-902-002, Loss of Coolant Accident Recovery, Technical Reference(s): Rev. 15 TG-OP-902-002, Technical Guide for Loss of (Attach if not previously provided) Coolant Accident Recovery, Rev. 14 (including version/revision number)

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

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

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

Rev 1, corrected error in selection B (changed large to small), removed unnecessary info from bullets 1 and 5.

Rev 2, typographical and editorial for clarity.

Facility: Waterford 3 Page 76 of 150

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

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

Proposed Question: RO 39 Rev: 3 Given:

Pressurizer pressure is 1650 PSIA and slowly lowering

Containment pressure is 16.7 and slowly rising

Steam Generator 1 pressure is 700 PSIA and lowering

Steam Generator 2 pressure is 680 PSIA and lowering Which of the following ESFAS signal(s) will have been generated?

A. SIAS only B. SIAS and CIAS only C. SIAS, CIAS, and MSIS only D. SIAS, CIAS, MSIS, and CSAS Facility: Waterford 3 Page 77 of 150

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

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

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

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

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

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

(including version/revision number)

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

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

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

Rev. 1 minor format and editorial changes.

Rev 2, typographical and editorial for clarity.

Rev. 3, added question mark to stem.

Facility: Waterford 3 Page 78 of 150

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

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

Plant was at 100% power

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

Containment Fan Cooler C is in standby

CEDM Fans A and D are in operation

Reactor was manually tripped and SIAS was manually actuated Which of the following describes the response during the event?

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

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

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

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

Facility: Waterford 3 Page 79 of 150

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

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

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

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

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

OP-008-004, Control Element Drive Mechanism Technical Reference(s): Cooling System, Rev. 7 OP-008-003, Containment Cooling System, Rev.

(Attach if not previously provided) 301 (including version/revision number)

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

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

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

Rev 1, minor editorial changes.

Rev 2, typographical and editorial for clarity.

Facility: Waterford 3 Page 80 of 150

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

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

A LOCA has occurred

RCS pressure is 700 PSIA

Containment pressure is 18 PSIA

RWSP level is 9% and lowering

NO operator actions have been taken Which of the following describes the alignment of ECCS?

SI-602 A and B SI-120 A and B A. open closed B. closed open C. open open D. closed closed Facility: Waterford 3 Page 81 of 150

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

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

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

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

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

OP-902-002, Loss of Coolant Accident Recovery, Technical Reference(s): Rev. 15 (Attach if not previously provided)

(including version/revision number)

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

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

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

Rev 1, remove second bullet, replaced Containment Spray System with ECCS at end of stem, and corrected valve names.

Rev 2, typographical and editorial for clarity.

Facility: Waterford 3 Page 82 of 150

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 039 A4.04 Importance Rating 3.8 K/A Statement A4.04 Ability to manually operate and/or monitor in the control room: Emergency feedwater pump turbines Proposed Question: RO 42 Rev: 3 Given:

Plant is in MODE 3 A Loss of Offsite Power occurred and both emergency diesels are running loaded 5 minutes into the event, EFAS 1 and EFAS 2 initiate and EFW Pump AB trips on ELECTRICAL overspeed BOP has closed both MS 401A and MS 401B, EFW Pump AB Turb Stm Supply Valves EFW Pump AB steam line has been depressurized In accordance with OP-009-003, Emergency Feedwater System, how will MS-416, EFW Pump AB Turbine Stop valve, be re-opened?

A. MS-416 will drive shut automatically. When closed, the valve operator re-latches. MS-416 will be re-opened from CP-8.

B. MS-416 will drive shut automatically. When closed, the NAO must reset the local linkage. MS-416 will be reopened from CP-8.

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

D. An NAO must manually close MS-416. When closed, the NAO must reset the local linkage. MS-416 must be opened locally by the NAO.

Facility: Waterford 3 Page 83 of 150

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

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

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

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

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

(including version/revision number)

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

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

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

Rev 1, revised per NRC comments.

Rev 2, typographical and editorial for clarity.

Rev 3, revised order of choices.

Facility: Waterford 3 Page 84 of 150

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

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

Plant has experienced an Excess Steam Demand event

Steam Generator 1 pressure is 400 PSIA

Steam Generator 2 pressure is 550 PSIA

Steam Generator 1 level is 5% NR and lowering

Steam Generator 2 level is 7% NR and lowering 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 Facility: Waterford 3 Page 85 of 150

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

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

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

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

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

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

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

OP-009-003, Emergency Feedwater System, Rev.

Technical Reference(s): 304 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-EFW00 obj. 2 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

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

Reworded question stem to focus answer on what is currently actuated, per original intent of the question. Removed unnecessary information from bullets. Rearranged answers short to long and numerically (EFAS 1 first).

Facility: Waterford 3 Page 86 of 150

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

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

Prior to taking action, Reactor trip occurs In accordance with OP-902-000, Standard Post Trip Actions, the BOP will:

A. Fully Close Main Feedwater Reg Valve 1, FW-173A Close Startup Feedwater Reg Valve 1, FW-166, to 13-21% open Reduce Main Feedwater Pump A speed to minimum B. Close Main Feedwater Reg Valve 1, FW-173A, to 13-21% open Fully Close Startup Feedwater Reg Valve 1, FW-166 Reduce Main Feedwater Pump A speed to minimum C. Fully Close Main Feedwater Reg Valve 1, FW-173A Close Startup Feedwater Reg Valve 1, FW-166, to 13-21% open D. Close Main Feedwater Reg Valve 1, FW-173A, to 13-21% open Fully Close Startup Feedwater Reg Valve, 1 FW-166 Facility: Waterford 3 Page 87 of 150

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

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

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

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

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

Technical Reference(s): OP-902-000, Standard Post Trip Actions, Rev. 13 OP-901-201, Steam Generator Level Control (Attach if not previously provided) Malfunction, Rev. 5 (including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PPO20 obj. 2 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

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

Rev 1, added fourth bullet, wrote out equipment names in selections, referenced OP-902-000 in stem.

Rev 2, typographical and editorial for clarity.

Facility: Waterford 3 Page 88 of 150

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

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

Plant is operating at 100% power

Emergency Feedwater Pump A is danger tagged out

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

Crew has entered OP-902-006, Loss of Main Feedwater Recovery

EFW Pump AB tripped 15 minutes after entry into OP-902-006 All running Reactor Coolant Pumps shall be tripped within 30 minutes of the loss of

____(1)____. This action is required to ____(2)____.

(1) (2)

A. both Main Feedwater ensure Core Heat Removal safety Pumps function is satisfied B. both Main Feedwater lower heat generation rate to less than Pumps EFW system heat removal capacity C. EFW Pump AB ensure Core Heat Removal safety function is satisfied D. EFW Pump AB lower heat generation rate to less than EFW system heat removal capacity Facility: Waterford 3 Page 89 of 150

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

A. Incorrect. EC28672 directs the crew to trip all 4 RCPs within 30 minutes of the time that the capacity of only one motor driven EFW Pump is available after a loss of Main Feedwater. The operations of the RCPs does impact the Core Heat Removal safety function, but this is not the reason the pumps must be secured.

B. Incorrect. EC28672 directs the crew to trip all 4 RCPs within 30 minutes of the time that the capacity of only one motor driven EFW Pump is available after a loss of Main Feedwater C. Incorrect. EC28672 directs the crew to trip all 4 RCPs within 30 minutes of the time that the capacity of only one motor driven EFW Pump is available after a loss of Main Feedwater. Two motor driven EFW pumps would be enough capacity to keep 2 RCPs running, therefore all 4 RCPs are not secured until after EFW Pump AB is unavailable.

D. CORRECT: EC28672 directs the crew to trip all 4 RCPs within 30 minutes of the time that the capacity of only one motor driven EFW Pump is available after a loss of Main Feedwater. The capacity of one motor driven EFW pump is not enough to make up for decay heat sources and Reactor Coolant Pump heat input.

OP-902-006, Loss of Main Feedwater Recovery Technical Reference(s): Revision 13 (Attach if not previously provided) EC28672 (including version/revision number)

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

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

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

Rev 1, revised per NRC comment to test on resolution of CR written in 2011 NRC exam.

The resolution was taught to the license class under WLP-OPS-PPE06.

Rev 2, typographical and editorial for clarity.

Facility: Waterford 3 Page 90 of 150

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 061 A3.03 Importance Rating 3.5 K/A Statement A3.03 Ability to monitor automatic operation of the AFW, including: AFW S/G level control on automatic start Proposed Question: RO 46 Rev: 2 A Loss of MFW event has occurred. After a manual Reactor trip, the following conditions are observed:

SG 1 pressure is 980 PSIA

SG 2 pressure is 960 PSIA

SG 1 level is 51% WR level and lowering

SG 2 level is 38% WR level and lowering

SG 2 EFW flow transmitter failed high

Pressurizer pressure 1900 PSIA

Containment pressure 15.3 PSIA Which of the following describes the EFW system response for SG 2?

A. Primary FCV and Backup FCV remain closed.

B. Primary FCV remains closed and backup FCV opens to 175 GPM.

C. Primary FCV opens to preset valve position and Backup FCV remains closed.

D. Primary FCV opens to preset valve position and Backup FCV opens to 400 GPM flow value.

Facility: Waterford 3 Page 91 of 150

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

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

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

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

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

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

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

Technical Reference(s): SD-EFW, Rev. 11 (Attach if not previously provided)

(including version/revision number)

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

Question Source: Bank # X 1786-A Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 4, 8 55.43 Comments:

Rev 1, revised per NRC comments.

Rev 2, typographical and editorial for clarity.

Facility: Waterford 3 Page 92 of 150

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

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

A. 1A B. 2A C. 1B D. 2B Facility: Waterford 3 Page 93 of 150

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

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

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

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

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

OP-003-035, Auxiliary Feedwater System, Rev.

Technical Reference(s): 301 (Attach if not previously provided)

(including version/revision number)

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

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

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

Rev 1, added bus to end of stem and eliminated bus from all answers.

Rev 2, typographical and editorial for clarity.

Facility: Waterford 3 Page 94 of 150

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 062 A2.12 Importance Rating 3.2 K/A Statement A2.12 Ability to (a) predict the impacts of the following malfunctions or operations on the ac distribution system; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Restoration of power to a system with a fault on it Proposed Question: RO 48 Rev: 2 Given:

Loss of offsite power occurred Indications for this condition were normal until:

Auxiliary Component Cooling Water (ACCW) Pump B tripped at the 17 second load block

EDG B Sequencer load block lights have extinguished

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

(1) (2)

A. all Train B sequencer RESET only loads B. all Train B sequencer RESET and then TEST loads C. Train B loads after 17 RESET only second load block D. Train B loads after 17 RESET and then TEST second load block Facility: Waterford 3 Page 95 of 150

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

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

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

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

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

Technical Reference(s): OP-901-311, Loss of Train B Safety Bus, Rev 308 OP-902-003, Loss of Offsite Power/Loss of Forced (Attach if not previously provided) Circulation Recovery, Rev. 7 (including version/revision number) SD-EDG, Rev. 16 Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PPO30 obj. 3 (As available)

Question Source: Bank # X 08930 Modified Bank # (Note changes or attach parent)

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

Rev 1, minor reformatting.

Rev 2, typographical and editorial for clarity.

Facility: Waterford 3 Page 96 of 150

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

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

or interlock(s) which provide for the following: Breaker interlocks, permissives, bypasses and cross-ties Proposed Question: RO 49 Rev: 2 Given:

Plant is at 100% power Loss of the 125 Volt TGB-DC bus has occurred Crew has entered OP-901-313, Loss of a 125 Volt DC Bus The backup over current protection for Reactor Coolant Pumps 1A and 2A is

____(1)____. All other remote manual control and automatic protection for Switchgear 1A including its associated connected components are ________(2)________.

(1) (2)

A. disabled disabled B. disabled available C. available available D. available disabled Facility: Waterford 3 Page 97 of 150

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

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

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

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

D. CORRECT: Backup overcurrent protection for RCPs is the feeder breaker to the 1A bus opening if the RCP load breaker does not open within 4 seconds. The feeder breaker control power is from the A DC bus, which is still available with the conditions given in the stem. The TGB-DC bus supplies the control power to the 1A(B) bus load breakers, which is not available. This renders the remote control and protective functions for the load breakers themselves unavailable.

Technical Reference(s): OP-901-313, Loss of a 125 Volt DC Bus, Rev. 303 (Attach if not previously provided) SD-7KV, Rev. 8 (including version/revision number)

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

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

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

Rev 1, minor reformatting. Added SD-7KV as a Technical Reference.

Rev 2, typographical and editorial for clarity.

Facility: Waterford 3 Page 98 of 150

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 064 K6.07 Importance Rating 2.7 K/A Statement K6.07 Knowledge of the effect of a loss or malfunction of the following will have on the ED/G system: Air receivers Proposed Question: RO 50 Rev: 3 Given:

EDG A Air Receiver A2 relief valve lifts and pressure drops to 0 psig If a loss of offsite power were to occur, EDG A would receive starting air to ____(1)____

cylinders. If the EDG fails to start immediately, it will continue to crank

________(2)________.

(1) (2)

A. all until Air Receiver A1 is depleted B. half until Air Receiver A1 is depleted C. all for a maximum of 25 seconds D. half for a maximum of 25 seconds Facility: Waterford 3 Page 99 of 150

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

A. CORRECT: EDG Starting Air has a cross-connect line downstream of the Air Start valves. This allows Starting Air to be supplied to all cylinders from either air receiver.

Per SD-EDG, the diesel will continue to crank on an emergency start until the air receiver is depleted, if the EDG does not reach normal speed.

B. Incorrect. Wrong number of cylinders. Correct cranking behavior.

C. Incorrect. Correct number of cylinders. Wrong cranking behavior.

D. Incorrect. Wrong number of cylinders. Wrong cranking behavior.

Technical Reference(s): SD-EDG, Rev. 16 (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 #

Modified Bank # (Note changes or attach parent)

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

Rev 1, replaced first part of question to eliminate determination of operability. Modified explanations. Provided additional Technical Reference.

Rev 2, typographical and editorial for clarity.

Rev 3, typographical.

Facility: Waterford 3 Page 100 of 150

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 064 A4.01 Importance Rating 4.0 K/A Statement A4.01 Ability to manually operate and/or monitor in the control room: Local and remote operation of the ED/G Proposed Question: RO 51 Rev: 2 With EDG A operating in parallel with the grid and in Local Control mode, the Speed Adjust control switch is taken to raise at the local control panel. EDG A speed will

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

(1) (2)

A. increase remain the same B. remain the same increase C. increase increase D. remain the same remain the same Facility: Waterford 3 Page 101 of 150

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

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

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

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

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

Technical Reference(s): SD-EDG, Rev. 16 (Attach if not previously provided)

(including version/revision number)

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

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

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

Rev 1, revised per NRC comments.

Rev 2, typographical and editorial for clarity.

Facility: Waterford 3 Page 102 of 150

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

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

Plant is at 100% power DRY COOLING TOWER SUMP 1 ACTIVITY HI annunciator is received High Activity reading on DCT Sump 1 radiation monitor, PRM-IRE-6775, indicated on the RM-11 Which of the following AUTOMATIC actions are expected for DCT 1 Sump?

A. Sump pumps trip.

B. Flow path to Circ Water isolates.

C. Flow path aligns to the Waste Tanks.

D. Flow path to the storm drains isolates.

Facility: Waterford 3 Page 103 of 150

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

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

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

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

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

OP-901-414, Effluent Discharge High Radiation, Technical Reference(s): Rev. 301 (Attach if not previously provided)

(including version/revision number)

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

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

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

Rev 1, reworded stem to Move DCT 1Sump to the Stem and eliminated from each selection. Reordered short to long Rev 2, typographical and editorial for clarity.

Facility: Waterford 3 Page 104 of 150

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

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

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

(1) (2)

A. 102°F after the CCW valves close B. 95°F first, then the CCW valves close C. 95°F after the CCW valves close D. 102°F first, then the CCW valves close Facility: Waterford 3 Page 105 of 150

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

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

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

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

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

Technical Reference(s): SD-CC, Rev. 18 (Attach if not previously provided)

(including version/revision number)

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

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

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

Rev 1, added outlet to title of heat exchanger temperature in stem.

Rev 2, typographical and editorial for clarity.

Facility: Waterford 3 Page 106 of 150

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

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

Proposed Question: RO 54 Rev: 2 Given:

IA Dryer Bypass, IA-123, automatically opened due to an Instrument Air leak

Leak has been isolated and air pressure is being restored IA Dryer Bypass, IA-123, ____(1)____ when the reset value is reached. Position indication for IA-123, can be monitored from ________(2)________.

(1) (2)

A. remains open control room annunciator and locally at the instrument air dryers.

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

C. remains open locally at the instrument air dryers only D. automatically locally at the instrument air dryers only closes Facility: Waterford 3 Page 107 of 150

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

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

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

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

Technical Reference(s): SD-AIR, Rev. 11 (Attach if not previously provided)

(including version/revision number)

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

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

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

Rev 1, removed unnecessary info from stem, changed first part of A and C requires manual closure to remains open.

Rev 2, typographical and editorial for clarity.

Facility: Waterford 3 Page 108 of 150

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

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

______(2)_____ when differential pressure lowers to the reset value.

(1) (2)

A. 8.5 automatically close B. 5.5 automatically close C. 8.5 remain open D. 5.5 remain open Facility: Waterford 3 Page 109 of 150

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

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

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

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

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

OP-008-005, Containment Vacuum Relief, Rev.

Technical Reference(s): 304 (Attach if not previously provided) SD-CB, Rev. 10 (including version/revision number)

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

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

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

Rev 1, replaced second part with valve operation on reset.

Rev 2, typographical and editorial for clarity.

Facility: Waterford 3 Page 110 of 150

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

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

Proposed Question: RO 56 Rev: 1 Given:

Plant is operating at 100% power

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

Using a contact pyrometer, the RCA watch has located increased temperature on the Regenerative Heat Exchanger drain line Which of the following valves is a possible RCS leakage source?

A. RC-204, Reactor Coolant Hot Leg 2 Drain B. RC-213, Reactor Coolant Loop 2A Charging Line Drain Isol C. RC-601, RCP Controlled Bleedoff to Reactor Drain Tank Isol D. RC-510A, RCP 2A Controlled Bleedoff to Reactor Drain Tank Facility: Waterford 3 Page 111 of 150

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

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

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

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

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

E.

OP-901-111, Reactor Coolant System Leak, Rev.

Technical Reference(s): 301.

(Attach if not previously provided)

(including version/revision number)

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

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

Question Source: Bank # X 07605 Modified Bank # (Note changes or attach parent)

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

Revised third bullet to remove line number and replaced with name of line to eliminate direct lookup. Reordered selections short to long.

Facility: Waterford 3 Page 112 of 150

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

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

Plant is at 100% power

AB busses are powered from Train B Charging Pump AB is powered from bus _____(1)______. Currently, Charging Pump AB can replace Charging Pump _____(2)______ without entering Technical Specification 3.1.2.4, Charging Pumps-Operating.

(1) (2)

A. AB3 B only B. AB31 B only C. AB3 A or B D. AB31 A or B Facility: Waterford 3 Page 113 of 150

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

A. Incorrect. Wrong bus. Correct pump.

B. CORRECT: The listed charging pump is powered from 480 VAC bus AB31. The charging pump can be aligned to replace Charging Pump B only for operability purposes. The pump can still be started manually in place of Charging Pump A after verifying Sequencer B is timed out; however to have operability the bus alignment must be the same as the pump replaced. Taking the AB assignment switch to the A side with AB bus aligned to the B bus will generate a Charging Pump AB Not Available/Not Aligned alarm (Cabinet G, A-5). OP-500-007, Attachment 4.5 states under Possible Effects that the pump is not available for an SIAS start.

C. Incorrect. Wrong bus. Wrong pump combination.

D. Incorrect. Correct bus. Wrong pump combination.

Technical Reference(s): SD-CVC, Rev. 15 (Attach if not previously provided) OP-500-007, Control Room Cabinet G, Rev. 20 (including version/revision number)

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

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

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

Rev 1, added bus before blank one, adding numbers to blanks, changed present to given, changed second part of question, revised Technical References.

Rev 2, typographical and editorial for clarity.

Facility: Waterford 3 Page 114 of 150

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

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

Plant is performing a startup and is at 80% power

Group P CEAs are being withdrawn for ASI control

Group Select switch is in the P position

CEA 25 in Group P becomes mechanically bound With the CEA Manual Shim switch in WITHDRAW, PMC pulse counter indication for CEA 25 is ____(1)____ and on the CEACs is ______(2)____.

(1) (2)

A. rising stationary B. stationary stationary C. rising rising D. stationary rising Facility: Waterford 3 Page 115 of 150

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

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

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

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

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

Technical Reference(s): SD-CED, Rev. 10 (Attach if not previously provided) SD-PMC, Rev. 8 (including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-CED00 obj. 17 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

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

Rev 1, removed unnecessary verbiage from stem.

Rev 2, typographical and editorial for clarity.

Facility: Waterford 3 Page 116 of 150

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 2 K/A # 016 K3.02 Importance Rating 3.4 K/A Statement K3.02 Knowledge of the effect that a loss or malfunction of the NNIS will have on the following: PZR LCS Proposed Question: RO 59 Rev: 1 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 Cold Leg (RC-ITI-0111-Y) indicates failed low Letdown flow will initially ____(1)____. In accordance with OP-901-110, Pressurizer Level Control Malfunction, the preferred method of selecting the non-faulted Tcold instrument is to operate the TCOLD LOOP 1 Selector switch located in the

______(2)_____ .

(1) (2)

A. lower RRS local cabinets B. lower rear of CP-2 C. rise RRS local cabinets D. rise rear of CP-2 Facility: Waterford 3 Page 117 of 150

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

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

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

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

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

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

Technical Reference(s): OP-901-110, Pressurizer Level Control (Attach if not previously provided) Malfunction, Rev. 6 (including version/revision number) SD-PLC, Rev. 8 Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-RR00 obj. 2 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

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

Removed unnecessary verbiage from stem and added in accordance with OP-901-110 Facility: Waterford 3 Page 118 of 150

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

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

Proposed Question: RO 60 Rev: 2 Given:

Plant was initially at 100% power Loss of offsite power has occurred Crew has entered OP-902-003, Loss of Offsite Power/Loss of Forced Circulation Recovery CRS has directed the BOP to protect the Main Condenser In accordance with OP-902-003, the BOP operator will close ____(1)____ valves to prevent overpressurizing the Main Condenser due to the absence of the

______(2)_____ .

(1) (2)

A. Steam Bypass Circulating Water System B. Steam Bypass Condenser Air Evacuation System C. Main Steam Condenser Air Evacuation Isolation System D. Main Steam Circulating Water System Isolation Facility: Waterford 3 Page 119 of 150

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

A. ncorrect. OP-902-003 step 8 requires the crew to protect the main condenser if offsite power is lost. The crew is directed to close the MSIVs and Blowdown Isolation Valves. The second portion is correct.

B. Incorrect. OP-902-003 step 8 requires the crew to protect the main condenser if offsite power is lost. The crew is directed to close the MSIVs and Blowdown Isolation Valves. Per the TG-OP-902-003, this step is performed to prevent overpressurization of the condenser due to a loss of the Circulating Water System.

C. Incorrect. The first portion is correct. Per the TG-OP-902-003, this step is performed to prevent overpressurization of the condenser due to a loss of the Circulating Water System.

D. CORRECT: OP-902-003 step 8 requires the crew to protect the main condenser if offsite power is lost. The crew is directed to close the MSIVs and Blowdown Isolation Valves. Per the TG-OP-902-003, this step is performed to prevent overpressurization of the condenser due to a loss of the Circulating Water System.

TG-OP-902-003, Technical Guide for Loss of Technical Reference(s): Offsite Power/Loss of Forced Circulation Recovery OP-902-003, Loss of Offsite Power/Loss of Forced (Attach if not previously provided) Circulation Recovery, Rev. 7 (including version/revision number)

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

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

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

Rev 1, replaced Question Rev 2, typographical and editorial for clarity.

Facility: Waterford 3 Page 120 of 150

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

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

Automatic purge isolation Proposed Question: RO 61 Rev: 3 Which of the following will automatically close Containment Purge Inlet Inside Annulus, CAP-103?

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

B. Hi-Hi alarm on Plant Stack PIG A gas channel (PRM-IRE-0100.1)

C. Containment pressure lowers to -5.5 INWG below atmospheric pressure.

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

Facility: Waterford 3 Page 121 of 150

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

A. Incorrect. RAB Normal exhaust flow decreasing to 84,000 scfm (not 90,000 scfm) will isolate CAP-103.

B. CORRECT: CAP-103 automatically closes on a hi-hi radiation signal from the Plant Stack Rad Monitor (PRM-IRE-0100.1)

C. Incorrect. Containment Pressure lowering to -8.4 INWG (not -5.5 INWG) below atmospheric pressure will close CAP-103.

D. Incorrect. CAP exhaust valves CAP-203, CAP-204 and CAP-205 must be full open for CAP-103 to remain open. The 52° open position is the normal position for these valves at 100% power.

OP-002-010, Reactor Auxiliary Building HVAC and Technical Reference(s): Containment Purge, Rev. 304 (Attach if not previously provided) SD-HVR, Rev. 9 (including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-HVR00 obj. 3 (As available)

Question Source: Bank # X 4118-A Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 9,11 55.43 Comments:

Rev 1, removed bullet from stem, changed B selection to match OP-002-010 automatic actions, the B train rad monitor would also cause the valve to go closed but due to other valves closing on its signal (Containment Purge Interrupt would occur and close any purge valve still open) also added PIG to distinguish the rad monitor for the WRGM.

Rev 2, typographical.

Rev 3, changed choice C from 6.5 to 5.5.

Facility: Waterford 3 Page 122 of 150

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 2 K/A # 034 A4.02 Importance Rating 3.5 K/A Statement A4.02 Ability to manually operate and/or monitor in the control room: Neutron levels Proposed Question: RO 62 Rev: 2 Given:

Plant is in a refueling outage

Refueling Group is preparing to withdraw the first fuel assembly from the Reactor Vessel The Source Range Neutron Flux Monitors shall be operable and operating with (1) channel(s) operable with continuous visible indication in the control room and (2) channel(s) operable with audible indication in the containment and the control room.

(1) (2)

A. two one B. one one C. two two D. one two Facility: Waterford 3 Page 123 of 150

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

A. CORRECT: Source range neutron flux monitors are verified operable by ensuring at least 2 channels operable with continuous visible indication in the control room and 1 channel operable with audible indication in the containment and control room. (TS 3.9.2)

B. Incorrect. Source range neutron flux monitors are verified operable by ensuring at least 2 channels operable with continuous visible indication in the control room.

C. Incorrect. TS 3.9.2 requires only 1 channel operable with audible indication in the containment and control room.

D. Incorrect. Source range neutron flux monitors are verified operable by ensuring at least 2 channels operable with continuous visible indication in the control room and 1 channel operable with audible indication in the containment and control room. (TS 3.9.2)

Technical Reference(s): TS 3.9.2 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-REQ04 obj. 2 (As available)

Question Source: Bank # X 2927-A Modified Bank # (Note changes or attach parent)

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

Rev 1, minor formatting and editorial changes.

Rev 2, typographical and editorial for clarity.

Facility: Waterford 3 Page 124 of 150

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 2 K/A # 041 A3.02 Importance Rating 3.3 K/A Statement A3.02 Ability to monitor automatic operation of the SDS, including: RCS pressure, RCS temperature, and reactor power Proposed Question: RO 63 Rev: 3 Given:

Plant is initially at 100% power

RCS Tavg is 573°F

Reactor Power Cutback system is out of service

Reactor trip on Turbine trip is enabled

Turbine trip occurs Which of the following describes the Steam Bypass Control system response immediately following the Turbine trip?

A. All valves quick open.

B. Only valves 1 through 5 quick open.

C. Only valves 1 through 3 quick open.

D. No valves quick open.

Facility: Waterford 3 Page 125 of 150

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

A. Incorrect. All SBCVs do not quick open on a reactor trip. SBC valve #6 quick open is blocked any time there is a reactor trip.

B. CORRECT: SBC valves 1-5 will quick open on a reactor trip with plant power at 100% power. SBC valve #6 quick open is blocked any time there is a reactor trip.

C. Incorrect. The quick open mode for the SBC system operates in two groups of 3 valves. Only one group of SBCVs quick opening will not handle a trip from 100%

power.

D. Incorrect. RCS Tave less than 561°F would block all SBCVs from quick opening.

The initial conditions in this event has Tave at 573°F.

Technical Reference(s): SD-SBC, Rev. 8 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-SBC00 obj. 5 (As available)

Question Source: Bank # X 07938 Modified Bank # (Note changes or attach parent)

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

Rev 1, minor rewording of stem and first bullet. and each answer to remove unnecessary information.

Rev 2, typographical and editorial for clarity.

Rev 3, typographical.

Facility: Waterford 3 Page 126 of 150

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 2 K/A # 045 K1.06 Importance Rating 2.6 K/A Statement K1.06 Knowledge of the physical connections and/or cause-effect relationships between the MT/G system and the following systems: RCS, during steam valve test Proposed Question: RO 64 Rev: 3 Given:

Plant is reducing turbine load to less than 1109 MWe to perform OP-903-007, Turbine Inlet Valve Cycling Test.

While Main Turbine load is reduced, RCS Tcold will (1)

During this reduction, Tcold is maintained 543°F to 546°F by (2) operation of the Steam Bypass Control System.

(1) (2)

A. lower automatic B. rise automatic C. lower manual D. rise manual Facility: Waterford 3 Page 127 of 150

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

A. Incorrect. Both parts. Tcold will rise as load is taken off the turbine. The SBCV will have to be opened to maintain RCS temperature as load is taken off the turbine. Per OP-903-007, Turbine Inlet Valve Cycling Test, RCS Tcold is maintained 543°F to 546°F by operating a SBCV in manual.

B. Incorrect. Second part. Per OP-903-007, Turbine Inlet Valve Cycling Test, RCS Tcold is maintained 543°F to 546°F by operating a SBCV in manual.

C. Incorrect. First part. Tcold will rise as load is taken off the turbine.

D. CORRECT: Tcold will rise as load is taken off the turbine. The SBCV will have to be opened to maintain RCS temperature as load is taken off the turbine. Per OP-903-007, Turbine Inlet Valve Cycling Test, RCS Tcold is maintained 543°F to 546°F by operating a SBCV in manual.

OP-903-007, Turbine Inlet Valve Cycling Test, Technical Reference(s): Rev. 14 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

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

Rev 1, rewrote question to focus on method for test setup and maintenance of Tcold.

Rev 2, typographical and editorial for clarity.

Rev 3. corrected Steam Bypass to Steam Bypass Control.

Facility: Waterford 3 Page 128 of 150

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 2 K/A # 086 A1.05 Importance Rating 2.9 K/A Statement A1.05 Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the Fire Protection System controls including: FPS lineups Proposed Question: RO 65 Rev: 3 Given:

Plant is in MODE 4 in preparation for a refueling outage

Crew is responding to an inadvertent CIAS

Airborne Radioactivity Removal System (ARRS) B charcoal filter catches fire

Pull stations for E-13B (ARRS B) and FPM-2 (Reactor Bldg Fire Main Header B) have been operated

Fire Brigade Leader reports that there is no deluge flow to ARRS B unit and it continues to burn To comply with OP-009-004, Fire Protection, the BOP should:

A. Operate the pull station for FPM-1, Reactor Bldg Fire Main Header A.

B. Direct the Fire Brigade Leader to spray down the ARRS B unit with the local fire hoses.

C. Direct the Fire Brigade Leader to locally actuate the Deluge System for the ARRS B unit.

D. Place the control switch for FP-601B, Reactor Bldg Fire Main Header B FPM-2 Cntmt Isol, to Close and then to Open.

Facility: Waterford 3 Page 129 of 150

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

A. Incorrect. Actuating FPM-1 will not help because FP-601A is closed and the fire is on the B ARRS unit.

B. Incorrect. Directing the Fire Brigade Leader to spray down the B Airborne Radioactivity Removal System (ARRS) charcoal unit with the local fire hoses will not work because FP-601A and FP-601B are closed due to the inadvertent CIAS.

C. Incorrect. The deluge system for the B Airborne Radioactivity Removal System (ARRS) unit can only be actuated from the control room. Plus, FP-601A and FP-601B are closed due to the inadvertent CIAS.

D. CORRECT: The inadvertent CIAS closed FP-601B, preventing fire protection water from reaching containment. OP-009-004 has the operator verify FP-601B open after the pull stations are actuated. FP-601B control switch must be taken to close then open to reset a CIAS.

Technical Reference(s): OP-009-004, Fire Protection, Rev. 313 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-FP00 obj. 1 (As available)

Question Source: Bank # X 07835 Modified Bank # (Note changes or attach parent)

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

Rev 1, reordered first and second bullet, reworded question to make component names consistent.

Rev 2, typographical and editorial for clarity.

Rev 3, changed order of B and C.

Facility: Waterford 3 Page 130 of 150

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 3 Group # 1 K/A # G2.1.1 Importance Rating 3.8 K/A Statement Conduct of Operations: Knowledge of conduct of operations requirements.

Proposed Question: RO 66 Rev: 1 Per EN-OP-115, Conduct of Operations, if an operating parameter exceeds any of the reactor protection set points and an automatic shutdown does not occur, the licensed operator is required to:

A. take action to restore the parameter within limit; if not successful, manually trip the Reactor.

B. get Control Room Supervisor concurrence and then manually trip the Reactor.

C. report tripping the Reactor while taking action to manually trip the Reactor.

D. get a peer check on the parameter and then manually trip the Reactor.

Facility: Waterford 3 Page 131 of 150

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

A. Incorrect. Per EN-OP-115, this is not the correct action to take when operating parameters exceed any of the reactor protection set points and an automatic shutdown does not occur B. Incorrect. Per EN-OP-115, this is not the correct action to take when operating parameters exceed any of the reactor protection set points and an automatic shutdown does not occur C. CORRECT: Per EN-OP-115, step 5.2, Licensed operators SHALL immediately insert a manual scram whenever Operating parameters exceed any of the reactor protection set points and an automatic shutdown does not occur.

D. Incorrect. Per EN-OP-115, this is not the correct action to take when operating parameters exceed any of the reactor protection set points and an automatic shutdown does not occur Technical Reference(s): EN-OP-115, Revision 12 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PPA00 Obj. 2 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

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

Minor editorial changes.

Facility: Waterford 3 Page 132 of 150

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 3 Group # 1 K/A # G2.1.3 Importance Rating 3.7 K/A Statement Conduct of Operations: Knowledge of shift or short-term relief turnover practices.

Proposed Question: RO 67 Rev: 1 Per EN-OP-115, Conduct of Operations, if a control room operator needs to be relieved for a short term during their shift to obtain a medical physical, permission must be granted from the ____(1)____ and it requires a ______(2)_____ prior to relieving the watch.

(1) (2)

A. FSS review of the EOS index B. SM or CRS verbal turnover C. FSS verbal turnover D. SM or CRS review of the EOS index Facility: Waterford 3 Page 133 of 150

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

A. Incorrect. Per EN-OP-115-3 step 5.7.2, the SM or CRS must give permission to be relieved on shift. A verbal turnover is required to be relieved of the watch B. CORRECT: Per EN-OP-115-3 step 5.7.2, the SM or CRS must give permission to be relieved on shift. A verbal turnover is also required to be relieved of the watch.

C. Incorrect. Per EN-OP-115-3 step 5.7.2, the SM or CRS must give permission to be relieved on shift.

D. Incorrect. Per EN-OP-115, a verbal turnover is required to be relieved of the watch.

Technical Reference(s): EN-OP-115-03, Revision 12 (Attach if not previously provided) OI-042-000, Rev. 27 (including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PPA00 Obj. 2 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

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

Added OI-042-000 as a Technical Reference.

Facility: Waterford 3 Page 134 of 150

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 3 Group # 1 K/A # G2.1.44 Importance Rating 3.9 K/A Statement Conduct of Operations: Knowledge of RO duties in the control room during fuel handling, such as responding to alarms from the fuel handling area, communication with the fuel storage facility, systems operated from the control room in support of fueling operations, and supporting instrumentation.

Proposed Question: RO 68 Rev: 3 Given:

Plant is in a refueling outage The crew will declare the plant to be in the DEFUEL mode of operation when informed by the Refuel SRO that all fuel has been removed from the Reactor Vessel and the last fuel bundle has been _______ .

A. removed from the Fuel Handling Building upender B. grappled, located in the Fuel Handling Building upender C. un-grappled, located in a Spent Fuel Pool storage location D. un-grappled, located in the Fuel Handling Building upender Facility: Waterford 3 Page 135 of 150

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

A. CORRECT: OP-010-006 Limitation 3.2.10 states that the plant is considered to be in the DEFUEL mode of operation when all fuel has been removed from the containment building including the reactor vessel and the last fuel bundle has been removed from the Upender in the FHB.

B. Incorrect. OP-010-006 Limitation 3.2.10 states that the plant is considered to be in the DEFUEL mode of operation when all fuel has been removed from the containment building including the reactor vessel and the last fuel bundle has been removed from the Upender in the FHB.

C. Incorrect. OP-010-006 Limitation 3.2.10 states that the plant is considered to be in the DEFUEL mode of operation when all fuel has been removed from the containment building including the reactor vessel and the last fuel bundle has been removed from the Upender in the FHB.

D. Incorrect. OP-010-006 Limitation 3.2.10 states that the plant is considered to be in the DEFUEL mode of operation when all fuel has been removed from the containment building including the reactor vessel and the last fuel bundle has been removed from the Upender in the FHB.

Technical Reference(s): OP-010-006 Rev. 320 Limitation 3.2.10 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-REQ04 obj. 4 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

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

Rev 1, changed the question since the original question was not considered Tier 3 Rev 2, typographical and editorial for clarity.

Rev 3, correct capital letters.

Facility: Waterford 3 Page 136 of 150

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 3 Group # 2 K/A # G2.2.6 Importance Rating 3.0 K/A Statement Equipment Control: Knowledge of the process for making changes to procedures.

Proposed Question: RO 69 Rev: 1 Prior to forwarding a Procedure Improvement Request (PIR) to the Operations Procedure Administrative Group (OPAG), the PIR must be verified to be complete, accurate, and proper justification provided by obtaining a review from ____(1)____ and

______(2)_____.

(1) (2)

A. the requestors immediate supervisor an independent SRO B. another licensed operator an independent SRO C. another licensed operator a 50.59 qualified reviewer D. the requestors immediate supervisor a 50.59 qualified reviewer Facility: Waterford 3 Page 137 of 150

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

A. CORRECT: Per OI-019-000, Operations Procedure Administration (steps 5.1.2 and 5.1.3) directs the PIR to have a review from the Immediate supervisor and an Independent SRO before forwarding the request to OPAG.

B. Incorrect. Per OI-019-000, Operations Procedure Administration (steps 5.1.2 and 5.1.3) directs the PIR to have a review from the Immediate supervisor and an Independent SRO before forwarding the request to OPAG.

C. Incorrect. Per OI-019-000, Operations Procedure Administration (steps 5.1.2 and 5.1.3) directs the PIR to have a review from the Immediate supervisor and an Independent SRO before forwarding the request to OPAG.

D. Incorrect. Per OI-019-000, Operations Procedure Administration (steps 5.1.2 and 5.1.3) directs the PIR to have a review from the Immediate supervisor and an Independent SRO before forwarding the request to OPAG.

Technical Reference(s): OI-019-000, Revision 303 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PPA00 obj. 2 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

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

Minor editorial changes.

Facility: Waterford 3 Page 138 of 150

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 3 Group # 2 K/A # G2.2.13 Importance Rating 4.1 K/A Statement Equipment Control: Knowledge of tagging and clearance procedures.

Proposed Question: RO 70 Rev: 1 Which of the following components qualifies as a NO TAG component on a Danger Tagout?

A. A drain valve that is going to be removed by maintenance.

B. A handwheel on a Motor Operated Valve being used as an isolation boundary.

C. A breaker that maintenance personnel will need to operate while signed on to the tagout.

D. A pump discharge valve that is used as the inside double valve isolation on a high energy system.

Facility: Waterford 3 Page 139 of 150

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

A. CORRECT: EN-OP-102 identifies a purpose of a no tag is to identify components that cannot be tagged but require positioning when hanging a tagout. Removing a drain valve meets this purpose.

B. Incorrect. Not allowed as a no tag per EN-OP-102.

C. Incorrect. Not allowed as a no tag per EN-OP-102.

D. Incorrect. Not allowed as a no tag per EN-OP-102.

Technical Reference(s): EN-OP-102, Revision 15 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: ELP-OPS-CLR (As available)

Question Source: Bank # X 08462a Modified Bank # (Note changes or attach parent)

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

Reordered selections short to long. Answer changed to A.

Facility: Waterford 3 Page 140 of 150

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 3 Group # 3 K/A # G2.2.22 Importance Rating 4.0 K/A Statement Equipment Control: Knowledge of limiting conditions for operations and safety limits.

Proposed Question: RO 71 Rev: 2 During hydrostatic testing of the RCS in MODE 5, RCS pressure increased to a point exceeding the RCS Pressure Safety Limit.

The RCS Pressure Safety Limit for this event is ____(1)____ PSIA and the maximum time allowed to reduce RCS pressure below the safety limit is ____(2)___ minutes.

(1) (2)

A. 2500 15 B. 2500 5 C. 2750 15 D. 2750 5 Facility: Waterford 3 Page 141 of 150

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

A. Incorrect. 2500 PSIA is the design pressure of the RCS and the pressure at which the Pressurizer relief valves lift. Fifteen minutes is a short time limit for other LCOs but not for RCS pressure.

B. Incorrect. 2500 PSIA is the design pressure of the RCS and the pressure at which the Pressurizer relief valves lift. The second portion of the answer is correct.

C. Incorrect. .The first portion of the answer is correct. Fifteen minutes is a short time limit for other LCOs but not for RCS pressure.

D. CORRECT: TS 2.1.2 Safety limit for RCS pressure states that the safety limit for Modes 3,4 and 5 is 2750 PSIA and whenever RCS pressure has exceeded 2750 PSIA, reduce the pressure within its limits within 5 minutes.

Technical Reference(s): Technical Specification 2.1.2 (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 7828 Modified Bank # (Note changes or attach parent)

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

Rev 1, replaced the K/A and the question.

Rev 2, typographical and editorial for clarity.

Facility: Waterford 3 Page 142 of 150

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 3 Group # 3 K/A # G2.3.4 Importance Rating 3.2 K/A Statement Radiation Control: Knowledge of radiation exposure limits under normal or emergency conditions.

Proposed Question: RO 72 Rev: 0 The routine administrative limit for TEDE with no undocumented quarters is ____(1)____

mrem/year. The Emergency Exposure TEDE guideline for lifesaving activities is

______(2)______ Rem.

(1) (2)

A. 2000 75 B. 4500 75 C. 2000 25 D. 4500 25 Facility: Waterford 3 Page 143 of 150

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

A. Incorrect. Correct routine admin limit. 75 Rem is the lens of the eye limit for lifesaving activities.

B. Incorrect. 4500 mrem/yr is the maximum allow administrative limit. Wrong emergency exposure limit.

C. CORRECT: Per EP-002-030 step 5.2.2, Emergency team members chosen to perform life saving activities shall not exceed 25 Rem. EN-RP-201 states the routine limit with no undocumented quarters is 2000 mrem/yr.

D. Incorrect. Wrong routine admin limit. Correct emergency exposure limit.

Technical Reference(s): EP-002-030, Rev. 10 (Attach if not previously provided) EN-RP-201, Rev. 3 (including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-EP02 obj. 8 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 9,12 55.43 Comments:

Facility: Waterford 3 Page 144 of 150

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 3 Group # 4 K/A # G2.3.13 Importance Rating 3.4 K/A Statement Radiation Control: Knowledge of radiological safety procedures pertaining to licensed operator duties, such as response to radiation monitor alarms, containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc.

Proposed Question: RO 73 Rev: 1 Which of the following is classified as a Very High Radiation Area with no access allowed in MODE 1?

A. Area outside the Regen HX Cubicle with Letdown in service B. -4ft MSL area around the Reactor Drain Tank C. Reactor Vessel Annulus D. Pressurizer cubicle Facility: Waterford 3 Page 145 of 150

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

A. Incorrect. The dose rates in this area would be high with Letdown in service but will not be a VHRA.

B. Incorrect. The background radiation in this area is also high but would not be a VHRA..

C. CORRECT: HP-001-213 step 5.2.2 states that the following areas have been designated as Very High Radiation Areas. Entries into these areas are forbidden when the reactor is in MODE 1: (Hot and Cold Leg D ring wall penetrations, Reactor Vessel Annulus, and the Reactor Cavity.)

D. Incorrect. Dose rates in the Pressurizer cubicle will be high but will not be a VHRA.

HP-001-213, Control of Reactor Containment Technical Reference(s): Building Power Entries, Rev. 303 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PPA00 obj. 3 (As available)

Question Source: Bank # X 3445-B Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 12 55.43 Comments:

Rephrased stem changed selections to one area vice two.

Facility: Waterford 3 Page 146 of 150

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 3 Group # 4 K/A # G2.4.5 Importance Rating 3.7 K/A Statement Emergency Procedures / Plan: Knowledge of the organization of the operating procedures network for normal, abnormal, and emergency evolutions.

Proposed Question: RO 74 Rev: 2 When the Severe Accident Management Guidelines (SAMGs) are implemented, the Emergency Operating Procedure (EOPs) are _____________.

A. exited and become the sole guidance for the event.

B. exited and may be re-entered if the SAMGs direct actions that are covered by the EOPs.

C. continued in parallel with the SAMGs with the Control Room crew implementing the SAMGs.

D. continued in parallel with the SAMGs and become the implementing procedure for the SAMGs.

Facility: Waterford 3 Page 147 of 150

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

A. Incorrect. The EOPs are not exited when SAMGs are implemented, the EOPs are performed in parallel with the SAMGs.

B. Incorrect. The EOPs are not exited when SAMGs are implemented, the EOPs are performed in parallel with the SAMGs.

C. Incorrect. Part 1 is correct. Per SAMG001, the SAMG is the controlling document and the EOPs and other procedures implement the CHLAs.

D. CORRECT: Part 1 is correct. Per SAMG001, the SAMG is the controlling document and the EOPs and other procedures implement the CHLAs.

Technical Reference(s): WLP-OPS-SAM00 slide 31 (Attach if not previously provided) SAMG, Severe Accident Management Guidelines (including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-SAM00 obj. 5 (As available)

Question Source: Bank # X 5229-A Modified Bank # (Note changes or attach parent)

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

Rev 1, replaced this question.

Rev 2, typographical and editorial for clarity.

Facility: Waterford 3 Page 148 of 150

2012 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 3 Group # 4 K/A # G2.4.14 Importance Rating 3.8 K/A Statement Emergency Procedures / Plan: Knowledge of general guidelines for EOP usage.

Proposed Question: RO 75 Rev: 2 Per OI-038-000, Emergency Operating Procedures Operations Expectations/Guidance, all available charging pumps running is defined as ____(1)____ running if no SIAS is initiated and ______(2)_____ running if SIAS is initiated.

(1) (2)

A. two two B. two three C. three three D. three two Facility: Waterford 3 Page 149 of 150

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

A. Incorrect. Per OI-038-000, step 5.4.49, if no SIAS is initiated all available charging pumps running is defined as three charging pumps running and with a SIAS initiated all available charging pumps running is defined as two charging pumps running..

B. Incorrect. Per OI-038-000, step 5.4.49, if no SIAS is initiated all available charging pumps running is defined as three charging pumps running and with a SIAS initiated all available charging pumps running is defined as two charging pumps running.

C. Incorrect. Per OI-038-000, step 5.4.49, if no SIAS is initiated all available charging pumps running is defined as three charging pumps running and with a SIAS initiated all available charging pumps running is defined as two charging pumps running.

D. CORRECT: Per OI-038-000, step 5.4.49, if no SIAS is initiated all available charging pumps running is defined as three charging pumps running and with a SIAS initiated all available charging pumps running is defined as two charging pumps running.

OI-038-000, Emergency Operating Procedures Technical Reference(s): Operations Expectations/Guidance, Rev. 5 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PPE08 obj. 8 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

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

Rev 1, replaced question.

Rev 2, typographical and editorial for clarity.

Facility: Waterford 3 Page 150 of 150

2012 NRC Exam SRO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 000008 AA2.23 Importance Rating 4.3 K/A Statement Ability to determine and interpret the following as they apply to the Pressurizer Vapor Space Accident: Criteria for throttling high-pressure injection after a small LOCA.

Proposed Question: SRO 1 Rev: 3 The following plant conditions exist:

LOCA is in progress due to a Pressurizer Safety Valve failing open

Pressurizer level is 100%

Thot is 532°F

Representative Core Exit Thermocouple temperature is 538°F

Pressurizer pressure is 1170 PSIA

QSPDS Reactor Vessel levels 1-3 indicates void, 4-8 are illuminated

High Pressure Safety Injection Pump (HPSI) B was secured and Train A flow was throttled 10 minutes ago

RCPs are secured

Steam Generator levels are 60% NR and being maintained by Emergency Feedwater In accordance with OP-902-002, Loss of Coolant Accident Recovery, the CRS should (1) because (2) requirement is not met.

(1) (2)

A. restore Pressurizer heaters and raise RCS subcooling RCS pressure B. raise Train A HPSI flow and start RCS subcooling HPSI Pump B C. restore Pressurizer heaters and raise Reactor Vessel level RCS pressure.

D. raise Train A HPSI flow and start Reactor Vessel level HPSI Pump B Facility: Waterford 3 Page 1 of 50

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

A. Incorrect. Restoring Pressurizer heaters and raising RCS pressure is a valid method for restoring subcooling when the RCS is intact. However, for the conditions given it would only serve to increase RCS leakage from the safety valve. Correct parameter evaluation.

B. CORRECT. CET temperature should be used with natural circulation in progress.

Although Thot indicates greater than 28°F subcooling, CETs are < 28°F. This does not meet HPSI Throttle Criteria, requiring the operator to restore HPSI flow as needed to restore RCS subcooling.

C. Incorrect. Wrong action. Reactor Vessel Level for HPSI throttle criteria is already met based on Level 5 NOT voided.

D. Incorrect. Correct action. Wrong parameter evaluation.

OP-902-002, Loss of Coolant Accident Recovery, Technical Reference(s): Rev. 15 TG-OP-902-002, Technical Guide for Loss of (Attach if not previously provided) Coolant Accident Recovery, Rev. 14 (including version/revision number)

Proposed references to be provided to applicants during examination: Steam Tables Learning Objective: WLP-OPS-PPE02, Objective 8 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

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

Rev 1, changed stem to a single sentence with two blanks after the bullets. Added Train A before HPSI flow and deleted commence a cooldown in Part 1 of B and D.

Rev 2. minor editorial changes as requested.

Rev 3, moved information from column 2 into stem.

Facility: Waterford 3 Page 2 of 50

2012 NRC Exam SRO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 000029 G2.4.2 Importance Rating 4.6 K/A Statement Knowledge of system set points, interlocks and automatic actions associated with EOP entry conditions. (ATWS)

Proposed Question: SRO 2 Rev: 2 The following plant conditions exist:

Plant is performing a shutdown due to a 25 gpm primary to secondary leak on Steam Generator 1

ATC operator reports the following indications:

Pressurizer level is 45% and lowering

Pressurizer pressure is 1750 PSIA and lowering

Containment pressure is 16.7 PSIA and rising

RCS temperature is lowering rapidly

Steam Generator 1 and 2 pressure is 750 PSIA and lowering

All Charging Pumps are operating

Reactor power is 78% and lowering

Per CRS direction the ATC performs a manual reactor trip, SIAS, and CIAS An Anticipated Transient Without Scram (ATWS) ______(1)______ occurred. After verification of Standard Post Trip Actions, the CRS will implement _______(2)_______.

(1) (2)

A. has OP-902-004, Excess Steam Demand Recovery B. has OP-902-008, Functional Recovery C. has NOT OP-902-004, Excess Steam Demand Recovery D. has NOT OP-902-008, Functional Recovery Facility: Waterford 3 Page 3 of 50

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

A. Incorrect. Correct ATWS evaluation. Wrong procedure implementation.

B. CORRECT. Prior to the ATC tripping the reactor RCS pressure went below the Aux Trip setpoint in the CPCs for RCS Pressure, therefore an ATWS has occurred. Two events are in progress. The diagnostic flowchart would send the crew to OP-902-008. If for some reason the CRS misuses the Diagnostic Flowchart, OP-902-004 Safety Function Status Checklist will not support implementation of the procedure due to secondary activity, based on given conditions.

C. Incorrect. Wrong ATWS evaluation. Wrong procedure implementation.

D. Incorrect. Wrong ATWS evaluation. Correct procedure implementation.

OP-902-009, Standard Appendices, Attachment 1, Technical Reference(s): Rev. 307 (Attach if not previously provided) TS Bases 2.2.1 Amendment 64 (including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PPE08 Obj. 7 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 5 Comments: Rev 1 - Changed the bullet in the plant conditions from "Reactor power is 78% and steady" to "Reactor power is 78% and lowering".

Rev 2 - Minor editorial changes to stem.

Facility: Waterford 3 Page 4 of 50

2012 NRC Exam SRO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 000055 EA2.06 Importance Rating 4.1 K/A Statement Ability to determine or interpret the following as they apply to a Station Blackout:

Faults and lockouts that must be cleared prior to re-energizing buses Proposed Question: SRO 3 Rev: 2 The following plant conditions exist:

Emergency Diesel Generator B is tagged out

Loss of offsite power has occurred

Emergency Diesel Generator A started; then tripped on Generator Differential

Gretna TOC reports offsite power is available from the grid The SRO should implement ____(1)____. To allow restoration of offsite power to the A3 bus, ____(2)____.

(1) (2)

A. OP-902-003, Loss of Offsite 86A2/HR must be reset and the 2A to 3A Tie Power/ Loss of Forced breaker must be shut Circulation Recovery B. OP-902-003, Loss of Offsite only 86A2/HR must be reset Power/ Loss of Forced Circulation Recovery C. OP-902-005, Station 86A2/HR must be reset and the 2A to 3A Tie Blackout Recovery breaker must be shut D. OP-902-005, Station only 86A2/HR must be reset Blackout Recovery Facility: Waterford 3 Page 5 of 50

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

A. Incorrect: First part. The plant is in a station blackout condition, therefore OP-902-005 is the correct EOP entry.

B. Incorrect: Both parts. The plant is in a station blackout condition, therefore OP-902-005 is the correct EOP entry. 86A2/HR must be reset after power is restored to the 2A bus in order to close the 2A to 3A Tie Breaker. To restore power to the 3A bus, the 2A to 3A Tie breaker is closed to re-energize the 3A to 2A Tie Bus UV relays in order to meet the interlocks for closing the 3A to 2A Tie breaker.

C. CORRECT: The plant is in a station blackout condition, therefore OP-902-005 is the correct EOP entry. 86A2/HR must be reset after power is restored to the 2A bus in order to close the 2A to 3A Tie Breaker. To restore power to the 3A bus, the 2A to 3A Tie breaker is closed to re-energize the 3A to 2A Tie Bus UV relays in order to meet the interlocks for closing the 3A to 2A Tie breaker.

D. Incorrect: Second part. 86A2/HR must be reset after power is restored to the 2A bus in order to close the 2A to 3A Tie Breaker. To restore power to the 3A bus, the 2A to 3A Tie breaker is closed to re-energize the 3A to 2A Tie Bus UV relays in order to meet the interlocks for closing the 3A to 2A Tie breaker.

OP-902-009,Standard Appendices, Attachments Technical Reference(s): 1, 12-A, and 12-B, Rev. 307 (Attach if not previously provided) SD-4KV, Rev. 6, Pgs 20-21 (including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-ED00, Obj. 4,7 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 5 Comments: Rev 1 - Added WLP-OPS-ED00, Rev 17, pages 140 and 141 to the technical references as it is the reference for part 2 of the question. (2) Deleted wording in the conditions due to it being irrelevant to answer the question.

Rev 2 - Revised second part of question to cover interlocks only. This eliminates potential overlap with OP Test. Changed initial conditions to support second part of question.

Facility: Waterford 3 Page 6 of 50

2012 NRC Exam SRO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 000058 AA2.03 Importance Rating 3.9 K/A Statement AA2.03 Ability to determine and interpret the following as they apply to the Loss of DC Power: DC loads lost; impact on ability to operate and monitor plant systems Proposed Question: SRO 4 Rev: 2 The following plant conditions exist:

Plant was at 100% power

Loss of A-DC-S Bus has occurred

Reactor trip has occurred

On the Generator trip, 2A Bus failed to transfer to Startup Transformer A

Crew has entered OP-902-000, Standard Post Trip Actions and OP-901-313, Loss of 125V DC Bus, concurrently Emergency Diesel Generator A is ___(1)___. The CRS should implement __(2)__.

(1) (2)

A. NOT running and its output OP-902-003, Loss of Offsite Power/ Loss breaker Open of Forced Circulation Recovery B. running with its output breaker OP-902-003, Loss of Offsite Power/ Loss Closed of Forced Circulation Recovery C. NOT running with its output OP-902-001, Reactor Trip Recovery, and breaker Open OP-901-310, Loss of Train A Safety Bus D. running with its output breaker OP-902-001, Reactor Trip Recovery, and Closed OP-901-310, Loss of Train A Safety Bus Facility: Waterford 3 Page 7 of 50

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

A. Incorrect: EDG A did not start due to the loss of DC power to the starting air solenoids.

EDG A Output Breaker cannot auto close due to the loss of DC control power. OP-902-003, Loss of Offsite Power/ Loss of Forced Circulation Recovery procedure is not diagnosed into because of the availability of the B train 7 Kv and 4 Kv busses.

B. Incorrect: EDG A did not start due to the loss of DC power to the starting air solenoids.

EDG A Output Breaker cannot auto close due to the loss of DC control power.

Diagnostics would result in implementing OP-902-001 Reactor Trip. Power is lost to the 3A bus so OP-901-310 is also applicable.

C. CORRECT: EDG A cannot start due to the loss of DC power to the starting air solenoids. EDG A Output Breaker cannot auto close due to the loss of DC control power. Because the B train 7 Kv and 4 Kv busses are energized, the Diagnostic flow chart would result in implementing OP-902-001 Reactor Trip. Power is lost to the 3A bus so OP-901-310 is also applicable.

D. Incorrect: EDG A did not start due to the loss of DC power to the starting air solenoids.

EDG A Output Breaker cannot auto close due to the loss of DC control power.

Diagnostics would result in implementing OP-902-001 Reactor Trip.

Technical Reference(s): OP-901-310, Loss of Train A Safety Bus, Rev. 308 OP-902-009, Standard Appendices, Attachment 1, (Attach if not previously provided) Rev. 307 (including version/revision number) SD-EDG Rev 16 pages 24, 43,100 Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PPO30 Obj. 1 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 5 Comments: Rev 1 - Clarified the explanations to identify why OP-902-001 is implemented vice OP-902-003.

Rev 2 - Added bullet for initial power. Changed Control Room Supervisor to CRS.

Minor editorial changes as requested to question and explanations.

Facility: Waterford 3 Page 8 of 50

2012 NRC Exam SRO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 000062 G2.1.32 Importance Rating 4.0 K/A Statement Ability to explain and apply system limits and precautions (Loss of Nuclear Service Water)

Proposed Question: SRO 5 Rev: 2 Given:

Reactor power is 100%

Wet Cooling Tower (WCT) A and B basin levels are verified to be 95%

The CRS should enter Technical Specification ___(1)___. The minimum water level in each WCT basin ensures a ____(2)____ day cooling water supply to essential equipment.

(1) (2)

A. 3.0.3 7 B. 3.0.3 30 C. 3.7.4 7 D. 3.7.4 30 Facility: Waterford 3 Page 9 of 50

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

A. Incorrect: Both trains of UHS are inoperable due to water levels below the minimum 97% required by TS LCO 3.7.4a. TS 3.0.3 is not entered since TS 3.7.4 provides an action for both UHS Trains inoperable. The limitation on minimum water level is based on providing a 30-day cooling water supply vice 7 days, which is a EDG fuel oil volume limitation.

B. Incorrect: TS 3.0.3 is not entered since TS 3.7.4 provides an action for both UHS Trains inoperable. The limitation on minimum water level is based on providing a 30-day cooling water supply.

C. Incorrect: Both trains of UHS are inoperable due to water levels below the minimum 97% required by TS LCO 3.7.4a. TS 3.7.4, Action b is entered for both UHS Trains inoperable. The limitation on minimum water level is based on providing a 30-day cooling water supply vice 7 days, which is a EDG fuel oil volume limitation.

D. CORRECT: Both trains of UHS are inoperable due to water levels below the minimum 97% required by TS LCO 3.7.4a. TS 3.7.4, Action b is entered for both UHS Trains inoperable. The limitation on minimum water level is based on providing a 30-day cooling water supply.

Technical Reference(s): TS 3.7.4 and Bases (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-CC00, Obj. 9 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 2 Comments: Rev 1 - Changed the Cognitive level to Memory or Fundamental Knowledge.

Rev 2 - Minor editorial changes as requested.

Facility: Waterford 3 Page 10 of 50

2012 NRC Exam SRO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 000077 G2.2.36 Importance Rating 4.2 K/A Statement Ability to analyze the effect of maintenance activities, such as degraded power sources, on the status of limiting conditions for operations. (Generator Voltage and Electric Grid Disturbances)

Proposed Question: SRO 6 Rev: 2 The following plant conditions exist:

Plant is in MODE 1

Shift Manager received a call from the grid operator informing him that maintenance activities are in progress have resulted in grid voltage at the Waterford 3 switchyard lowering to 222 kV, with a predicted post-trip voltage of 220 kV

All breakers from offsite to Class 1E distribution are closed

Voltage on the B31 bus is below the acceptance criteria of OP-903-066, Electrical Breaker Alignment Check The CRS should implement, (1) and is required to enter (2) .

(1) (2)

A. OP-901-314, Degraded Grid TRM 3.8.1.1 only Conditions only B. OP-901-314, Degraded Grid TRM and TS 3.8.1.1 Conditions only C. OP-901-314, Degraded Grid TRM 3.8.1.1 only Conditions, and OP-901-310, Loss of Train A Safety Bus D. OP-901-314, Degraded Grid TRM and TS 3.8.1.1 Conditions, and OP-901-310, Loss of Train A Safety Bus Facility: Waterford 3 Page 11 of 50

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

A. Incorrect: Second part. It is required to enter TS 3.8.1.1 if any voltages on the bus do not meet the acceptance criteria for the TS surveillance (OP-903-066).

B. CORRECT: Only OP-901-314 is required to be entered. Step 7 of E1 specifically discusses degraded voltage conditions of 223.1 kV. It is required to enter TRM 3.8.1.1 when your predicted post trip voltage at the W3 switchyard is less than 223 kV. It is required to enter TS 3.8.1.1 if any voltages on the bus do not meet the acceptance criteria for the TS surveillance (OP-903-066).

C. Incorrect: Both parts. It is not required to enter OP-901-310 unless the 480v bus is de-energized. It is required to enter TS 3.8.1.1 if any voltages on the bus do not meet the acceptance criteria for the TS surveillance (OP-903-066).

D. Incorrect: It is not required to enter OP-901-310 unless the 480v bus is de-energized.

OP-901-314, Degraded Grid Conditions, Rev. 2 Technical Reference(s): E1 Step 7 (Attach if not previously provided) TRM 3.8.1.1, Tech Spec 3.8.1.1 (including version/revision number) OP-901-310, Loss of Train A safety Bus, Rev.

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PPO30 Obj. 05 (As available)

Question Source: Bank # WF3-OPS-08876 Modified Bank # X (Note changes or attach parent)

New Question History: Last NRC Exam 2009 NRC Exam SRO #6 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 2 Comments: Rev. 1 - Changed part 2 of distracters B and C from one hour to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Minor editorial changes.

Rev. 2 - Rewrote question to eliminate need for references and changed first part of question to address procedures to enter.

Facility: Waterford 3 Page 12 of 50

2012 NRC Exam SRO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 2 K/A # 000036 G2.4.41 Importance Rating 4.6 K/A Statement Knowledge of the emergency action level thresholds and classifications. (Fuel Handling Accident)

Proposed Question: SRO 7 Rev: 2 The following plant conditions exist:

Plant is in MODE 6

Full core offload is in progress

Containment Purge is in operation with the Maintenance Hatch open

Crew received the following annunciators:

o CLASS 1E RAD MONITORING SYS ACTIVITY HI-HI (CP-18, Panel SA) o CLASS 1E RAD MONITORING SYS ACTIVITY HI-HI (CP-18, Panel SB) o RAD MONITORING SYS ACTIVITY HI-HI (CP-36, Panel L) o EFFLUENT RAD MONITORING SYS ACTIVITY HI-HI (CP-36, Panel L)

Refueling SRO reported that a fuel assembly slipped off the grapple during withdrawal and fell back into the core

ATC operator reports the following radiation monitor indications:

o All four Containment Purge Isolation Radiation Monitors ARM-IRE-5024(5025)(5026)(5027) are in High Alarm o Plant Stack PIG Gas Channels (PRM-IRE-0100.1S(2S)) are reading 1.00 E-1 Ci/cc o Plant Stack WRGM (PRM-IRE-0110, RE0110-4) indicates release rate 2.62E+08 Ci/sec and slowly rising

BOP operator reports that Containment Purge will not secure automatically or manually.

A station blackout occurred All indications above have existed for > 20 minutes. Which of the following is the highest Emergency Plan classification for this event?

A. Unusual Event B. Alert C. Site Area Emergency D. General Emergency Facility: Waterford 3 Page 13 of 50

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

A. Incorrect: Containment Purge Isolation Monitors are in High alarm, which meets part of AA2 criteria, but the Plant Stack WRGM reading supersedes the Containment Purge Monitor data, requiring entry into a higher classification.

B. Incorrect: Containment Purge Isolation Monitors are in High alarm, which exceeds part of AU2 criteria, but the Plant Stack PIGs and Plant Stack WRGM readings supersede the Containment Purge Monitor data, requiring entry into a higher classification.

C. CORRECT: The Plant Stack WRGM release rate reading is >2.55E+08 Ci/sec, requiring the candidate to classify the event as a Site Area Emergency, AS1 based on EAL 1.

D. Incorrect: No information provided reaches the General Emergency threshold. The Plant Stack WRGM release rate would have to be >2.55E+09 Ci/sec to classify a General Emergency.

EP-001-001, Recognition and Classification of Technical Reference(s): Emergency Conditions, Rev 030 - TAB A (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided EP-001-001, Recognition and Classification of to applicants during examination: Emergency Conditions, Attachment 7.1 only.

Learning Objective: WLP-OPS-EP02 Obj. 17 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 4 Comments: Rev 1 - Removed the classification numbers following the event classification since all answers are Abnormal Radiation Levels/ Radiological Effluents.

Rev 2, minor editorial changes.

Rev 3 - Added station blackout and reduced applicant handout to just attachment 7.1.

Facility: Waterford 3 Page 14 of 50

2012 NRC Exam SRO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 2 K/A # 000037 G2.2.22 Importance Rating 4.7 K/A Statement Knowledge of limiting conditions for operations and Safety Limits. (Steam Generator Tube Leak)

Proposed Question: SRO 8 Rev: 2 Given:

Plant at 100% power and steady state

ATC operator reports N-16 Radiation Monitor for Steam Generator 2 reads 65 GPD

PSLR reads 85 GPD

Chemistry reports primary to secondary leakage in Steam Generator 2 is consistent with PSLR Technical Specification 3.4.5.2, Operational Leakage, is _____(1)______.

The Nuclear Energy Institute recommended a limit of 150 GPD for primary to secondary leakage; however, the limit in Technical Specification 3.4.5.2 is more restrictive

____(2)____.

(1) (2)

A. not met to ensure the magnitude of leakage does not interfere with the detection of UNIDENTIFIED LEAKAGE by the leakage detection systems B. met to ensure the magnitude of leakage does not interfere with the detection of UNIDENTIFIED LEAKAGE by the leakage detection systems C. not met because the proximity of the east ADV to the east control room air intake could result in unacceptable radiological consequences D. met because the proximity of the east ADV to the east control room air intake could result in unacceptable radiological consequences Facility: Waterford 3 Page 15 of 50

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

A. Incorrect: Part 1 is correct - T.S. 3.4.5.2 Action a. With any PRESSURE BOUNDARY LEAKAGE, or primary to secondary leakage not within limit, be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . T.S. LCO 3.4.5.2 LCO c requires entry for 75 gallons per day primary to secondary leakage through any one steam generator (SG), which 85 GPD exceeds.

The description in Part 2 is the bases for IDENTIFIED LEAKAGE limits, not primary to secondary leakage.

B. Incorrect: Primary to secondary leakage is specifically excluded from TS 3.4.5.2, Action b. The description in Part 2 is the bases for IDENTIFIED LEAKAGE limits, not primary to secondary leakage.

C. CORRECT: T.S. 3.4.5.2 Action a. is correct based on given leakage greater than 75 gallons per day primary to secondary leakage through SG #2. The bases is correct based on not exceeding 5 REM TEDE per person for the duration of the accident.

D. Incorrect: Primary to secondary leakage is specifically excluded from TS 3.4.5.2, Action b. The bases is correct based on not exceeding 5 REM TEDE per person for the duration of the accident.

Technical Reference(s): Technical Specification 3.4.5.2, Amendment 204 (Attach if not previously provided) and Bases (including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-RCS00 Obj. 9 (As available)

WLP-OPS-PPO20 Obj. 5 Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 2 Comments: Rev 1 - Changed the stem of the question to ask the applicant if TS 3.4.5.2 is met or not met vice what action is required to be entered. Combined the last two sentences of the stem. Changed part 1 of the answers to met or not met for TS 3.4.5.2 vice what action is to be entered.

Rev 2 - Corrected title of Tech Spec 3.4.5.2 in first sentence in stem. Minor editorial changes as requested.

Facility: Waterford 3 Page 16 of 50

2012 NRC Exam SRO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 2 K/A # 000061 AA2.06 Importance Rating 4.1 K/A Statement Ability to determine or interpret the following as they apply to the Area Radiation Monitoring (ARM) System Alarms: Required actions if alarm channel is out of service Proposed Question: SRO 9 Rev: 0 The following plant conditions exist:

Plant is in MODE 3

Fuel movement is occurring in the Spent Fuel Pool Containment Purge Monitor Reading Monitor Alarm/Trip Setpoint ARM-IRE-5024S reads 150 mR/hr 320 mR/hr ARM-IRE-5025S reads 67 mR/hr 120 mR/hr ARM-IRE-5026S reads 15 mR/hr 50 mR/hr ARM-IRE-5027S reads 19 mR/hr 40 mR/hr ARM-IRE-____(1)___ are inoperable. Enter Technical Specification ___(2)____.

(1) (2)

A. 5025S and 5027S 3.3.3.1 & 3.9.4 B. 5024S and 5026S 3.3.3.1 & 3.9.4 C. 5025S and 5027S 3.3.3.1 only D. 5024S and 5026S 3.3.3.1 only Facility: Waterford 3 Page 17 of 50

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

A. Incorrect: ARM-IRE-5025S and ARM-IRE-5027S are operable. 5025S based on setpoint less than 2 times background and 5027S based on setpoint appropriately at 40 mR/hr. No Tech Spec entry is required for these monitors.

B. Incorrect: ARM-IRE-5024S is inoperable because its setpoint is greater than 2 times background. ARM-IRE-5026S is inoperable because its setpoint should be the higher of 2 times background (30 mR/hr) or 40 mR/hr. Based on both monitors inoperable on the same train, Tech Spec 3.3.3.1 entry is required. T.S. 3.9.4 entry not required since is only applicable during CORE ALTERATIONS or load movements with or over irradiated fuel within the containment, which cannot occur in MODE 3.

C. Incorrect: ARM-IRE-5025S and ARM-IRE-5027S are operable. 5025S based on setpoint less than 2 times background and 5027S based on setpoint appropriately at 40 mR/hr. No Tech Spec entry is required for these monitors.

D. CORRECT: ARM-IRE-5024S is inoperable because its setpoint is greater than 2 times background. ARM-IRE-5026S is inoperable because its setpoint should be the higher of 2 times background (30 mR/hr) or 40 mR/hr. Based on both monitors inoperable on the same train, Tech Spec 3.3.3.1 entry is required. T.S. 3.9.4 is only applicable during CORE ALTERATIONS or load movements with or over irradiated fuel within the containment, which cannot occur in MODE 3.

Technical Reference(s): Tech Specs 3.3.3.1 & 3.9.4 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: Tech Specs 3.3.3.1 & 3.9.4 Learning Objective: WLP-OPS-RMS00 Obj. 2,7 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

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

Facility: Waterford 3 Page 18 of 50

2012 NRC Exam SRO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 2 K/A # CE/E09 EA2.1 Importance Rating 4.4 K/A Statement Ability to determine and interpret the following as they apply to the (Functional Recovery): Facility conditions and selection of appropriate procedures during abnormal and emergency operations.

Proposed Question: SRO 10 Rev: 3 The following plant conditions exist:

EDG B is danger tagged out

Loss of offsite power has occurred

Plant has tripped

EDG A tripped

RCS pressure is 1150 PSIA

Containment pressure is 17.8 PSIA

CET temperature is 451 °F

Activity is indicated on Main Steam Line Rad Monitor 1 (ARM-IRE-5500A)

Steam Generator 1 pressure is 300 PSIA and lowering

Steam Generator 2 pressure is 450 PSIA and steady

Steam Generator 1 level is 35 percent WR and lowering rapidly

Steam Generator 2 level is 78 percent WR and lowering slowly

Subcooled margin is 111 °F

Containment temperature is 216 °F

RVLMS level indicates 100 percent for both head and plenum

Crew has entered OP-902-008, Functional Recovery procedure Which section of OP-902-008, Functional Recovery, should be implemented first?

A. MVA-AC-2 B. IC-2 C. HR-2 D. CI-1 Facility: Waterford 3 Page 19 of 50

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

A. CORRECT: AC-2 is the highest priority Safety Function that does not meet any Success Path criteria based on no offsite power and no EDG available.

B. Incorrect: IC-2 Safety Function does not meet any Success Path criteria due to no SI pumps available but is below AC-2 in order of prioritization.

C. Incorrect: HR-2 Safety Function does not meet any Success Path criteria due to no SI pumps available but is below AC-2 in order of prioritization.

D. Incorrect: CI-1 Safety Function is not met. This is due to the fact that all release paths from the most affected SG to the environment are NOT closed but CI-1 is below AC-2 in order of prioritization.

Technical Reference(s): OP-902-008, Functional Recovery, Rev. 018 (Attach if not previously provided) Pages 14-23 (including version/revision number)

Proposed references to be provided OP-902-008, Functional Recovery Procedure, to applicants during examination: Pages 11, 14-23 Learning Objective: WLP-OPS-PPE08, Objective 6 (As available)

Question Source: Bank # WF3-OPS-6925-A Modified Bank # X (Note changes or attach parent)

New Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 5 Comments: Rev 1 - Added to the conditions that EDG B is tagged out. With the other indications present, a Station Blackout would be occurring. This would make the correct answer A since MVA-AC-2 success path is not being met and it ranks the highest in priority. Changed all four explanations to include the station blackout occurring. Deleted the bullet in the conditions for HPSI flow is 100 gpm per leg because in a station blackout there would be no SI pumps available.

Rev 2, Moved fourth bullet to bottom. Deleted section about assumptions. Minor editorial changes as requested.

Rev 3, Added information for S/G 1 level and S/G 2 pressure.

Facility: Waterford 3 Page 20 of 50

2012 NRC Exam SRO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 006 A2.02 Importance Rating 4.3 K/A Statement A2.02 Ability to (a) predict the impacts of the following malfunctions or operations on the ECCS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Loss of flow path Proposed Question: SRO 11 Rev: 3 The following plant conditions exist:

LOCA occurred 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months ago

Crew is performing OP-902-002, Loss of Coolant Accident Recovery

Recirculation Actuation Signal has occurred, and required actions have been taken

RCS is 15°F subcooled; CET temperatures are rising very slowly

QSPDS RVLMS levels 1-6 indicates void

BOP notes that HPSI A and B pumps cold leg flow has dropped off slowly with discharge pressure rising and no oscillations

HPSI Cold Leg flow criteria is met

Containment Spray Pumps A and B are running with flow ~ 2000 GPM each train and stable The CRS should ______(1)______. The conditions given indicate _____(2)_____.

(1) (2)

A. remain in OP-902-002 and align core flow blockage is starting to occur due hot and cold leg injection to boron precipitation B. secure one HPSI pump and go core flow blockage is starting to occur due to OP-902-008, Functional to boron precipitation Recovery C. remain in OP-902-002 and align the SI Sump Strainers are starting to clog hot and cold leg injection from debris collection D. secure one HPSI pump and go the SI Sump Strainers are starting to clog to OP-902-008, Functional from debris collection Recovery Facility: Waterford 3 Page 21 of 50

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

A. CORRECT. The indications given indicate a problem developing on the discharge side of the High Pressure Safety Injection Pumps. At 2-3 hours post LOCA the crew should align hot and cold leg injection to avoid concentrating boric acid in the reactor vessel and possible boron precipitation in the core which could restrict coolant flow through the core.

B. Incorrect. Core flow blockage indications are given making Part 2 correct. Either condition listed in Part 2 is addressed in OP-902-002, and therefore, OP-902-008 entry is not required.

C. Incorrect. The indications given indicate a problem developing on the discharge side of the High Pressure Safety Injection Pumps, not the suction side (SI Sump Strainers). Step 47 of OP-902-002 directs performance of Hot and Cold Leg Injection 2 to3 hours post-LOCA. Either condition listed in Part 2 is addressed in OP-902-002, and therefore, OP-902-008 entry is not required.

D. Incorrect. Both parts are incorrect. The indications given indicate a problem developing on the discharge side of the High Pressure Safety Injection Pumps, not the suction side (SI Sump Strainers). Step 45 of OP-902-002 directs going down to one HPSI pump operating to prevent heat buildup and subsequent pump seal damage in low flow conditions. This condition is addressed in OP-902-002, and therefore, OP-902-008 entry is not required.

OP-902-002, Loss of Coolant Accident Recovery, Technical Reference(s): Rev. 15 TGOP-902-002, Technical Guide for Loss of (Attach if not previously provided) Coolant Accident Recovery, Rev. 14 (including version/revision number)

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

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 5 Comments: Rev 1 - Editorial changes to the conditions were made as suggested.

Rev 2 - Editorial changes were made as suggested.

Rev 3, Editorial changes for void display.

Facility: Waterford 3 Page 22 of 50

2012 NRC Exam SRO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 013 G2.2.22 Importance Rating 4.7 K/A Statement Knowledge of limiting conditions for operations and safety limits. (Engineering Safeguards Features Actuation)

Proposed Question: SRO 12 Rev: 2 Given:

Plant in MODE 3

Preparations are being made for energizing Control Element Drive Mechanisms

While performing OP-903-107, Plant Protection System Function Test, operator notes Low Steam Generator Water Level No. 1 trip bistable is tripping below the surveillance requirements Based on these conditions, the CRS should enter Technical Specification(s) ______(1)_____.

The action is to bypass _________(2)__________.

(1) (2)

A. 2.2.1, 3.3.1, and Steam Generator 1 Level Low, Steam Generator P 1 and 2 3.3.2 B. 3.3.2 only Steam Generator 1 Level Low, Steam Generator P 1 and 2 C. 2.2.1, 3.3.1, and Steam Generator 1 Level Low, Steam Generator P 1 only 3.3.2 D. 3.3.2 only Steam Generator 1 Level Low, Steam Generator P 1 only Facility: Waterford 3 Page 23 of 50

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

A. Incorrect: 2.2.1 is only applicable for the instrumentation that is inoperable for the modes that the associated trip function instrumentation is applicable for. Steam Generator Level Low is only applicable in MODE 1 and 2 for TS 3.3.1. Steam Generator P 2 would only have to be bypassed if a Steam Generator Pressure channel were out of service.

B. Incorrect. The listed action is correct for the current plant mode. Steam Generator P 2 would only have to be bypassed if a Steam Generator Pressure channel were out of service.

C. Incorrect: 2.2.1 is only applicable for the instrumentation that is inoperable for the modes that the associated trip function instrumentation is applicable for. Steam Generator Level Low is only applicable in MODE 1 and 2 for TS 3.3.1.

D. CORRECT: Steam Generator Level Low is applicable in Modes 1, 2, and 3, for TS 3.3.2. The action is to bypass Steam Generator 1 Level Low, Steam Generator P 1 only. Steam Generator P 2 would only have to be bypassed if a Steam Generator Pressure channel were out of service. TS 3.3.1 and 2.2.1 are not applicable for Steam Generator Level low in MODE 3.

Technical Reference(s): TS 2.2.1 (Attach if not previously provided) TS 3.3.1 (including version/revision number) TS 3.3.2 OP-903-107, Plant Protection System, Rev. 304 Proposed references to be provided to applicants during examination: TS 2.2.1, 3.3.1, 3.3.2 Learning Objective: WLP-OPS-PPS00 Obj. 8 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 2 Comments: Rev 1 - Eliminated the portion of the second bullet that indicated the value at which the trip bistable is tripping and indicated that the bistable is tripping below the surveillance requirement. Eliminated the third bullet that indicates what TS have been reviewed. This information is not needed to answer the question. Created new bullet describing why surveillance was being performed and removed that information from the statement regarding plant MODE.

Rev 2 - Revised stem and conditions as requested.

Facility: Waterford 3 Page 24 of 50

2012 NRC Exam SRO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 026 G2.1.20 Importance Rating 4.6 K/A Statement Ability to interpret and execute procedure steps.

Proposed Question: SRO 13 Rev: 2 The following plant conditions exist:

Plant has experienced a LOCA concurrently with a Loss of offsite power

EDG A started and loaded, but EDG B tripped on overspeed

Crew has entered OP-902-002, Loss of Coolant Accident Recovery

Containment pressure is 17.8 PSIA and rising

RCS pressure is 850 PSIA and steady

Containment Spray Pump A tripped on overcurrent In response to this event, the SRO should ____(1)___ and ____(2)____ to address the Containment Temperature and Pressure Control safety function.

(1) (2)

A. continue in OP-902-002, Loss of realign LPSI Pump A to replace Coolant Accident Recovery Containment Spray Pump A B. continue in OP-902-002, Loss of override and close CS-125A Coolant Accident Recovery C. go to OP-902-008, Functional realign LPSI Pump A to replace Recovery Containment Spray Pump A D. go to OP-902-008, Functional override and close CS-125A Recovery Facility: Waterford 3 Page 25 of 50

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

A. Incorrect: The Safety Function Status Checklist for OP-902-002 requires at least one Containment Spray (CS) pump running with greater than or equal to 1750 GPM flow requiring the crew to either enter directly into, or rediagnose into OP-902-008.

OP-902-008 Continuing Actions will lead the crew to Appendix 28 to align LPSI Pump A to replace CS Pump A.

B. Incorrect: The Safety Function Status Checklist for OP-902-002 requires at least one Containment Spray (CS) pump running with greater than or equal to 1750 GPM flow requiring the crew to enter directly into, or rediagnose into OP-902-008. Although the crew could override closed CS-125A to meet the CI-1 safety function, they would have to re-open CS-125A to meet the CTPC-2 safety function while aligning LPSI Pump A to replace CS Pump A.

C. CORRECT: Entry into OP-902-008 is correct. OP-902-008 Continuing Actions will lead the crew to Appendix 28 to align LPSI Pump A to replace CS Pump A with TSC concurrence.

D. Incorrect: Entry into OP-902-008 is correct. The crew could override closed CS-125A to meet the CI-1 safety function, but they would have to re-open CS-125A to meet the CTPC-2 safety function while aligning LPSI Pump A to replace CS Pump A.

OP-902-002, Loss of Coolant Accident Recovery, Technical Reference(s): Rev. 015 (Attach if not previously provided) OP-902-008, Functional Recovery, Rev. 018 OP-902-009, Standard Appendices, Rev. 307, (including version/revision number) Standard Appendix 28, Attachment 28-A TG-OP-902-009, Rev. 306 Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PPE02, Objective 19 (As available)

Question Source: Bank # X WF3-OPS-07137 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 5 Comments: Rev 1 - Made editorial changes to the conditions as requested. Re-wrote the explanations to make them more understandable.

Rev 2 - Made editorial changes to the conditions as requested.

Facility: Waterford 3 Page 26 of 50

2012 NRC Exam SRO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 064 A2.06 Importance Rating 3.3 K/A Statement A2.06 Ability to (a) predict the impacts of the following malfunctions or operations on the ED/G system; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Operating unloaded, lightly loaded, and highly loaded time limit Proposed Question: SRO 14 Rev: 3 Plant conditions are as follows:

EDG B was started for OP-903-068, Emergency Diesel Generator and Subgroup Relay Operability Verification

EDG B load is 2.1 MW

EDG B reactive load is 1.0 MVAR

A malfunction occurred that prevented any additional load changes for 7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months The impact of operating in the described condition is that there is a potential for the (1) . To mitigate the consequences of this condition, the CRS should direct the BOP to carry out the direction in OP-009-002, Emergency Diesel Generator, and (2) .

(1) (2)

A. fuel injection pumps to develop raise DC field current to obtain leaks due to low flow 195 - 205 Amps DC B. fuel injection pumps to develop raise load to > 4.0 MW and hold leaks due to low flow for 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months C. buildup of unburned exhaust raise DC field current to obtain products 195 - 205 Amps DC D. buildup of unburned exhaust raise load to > 4.0 MW and hold products for 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Facility: Waterford 3 Page 27 of 50

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

A. Incorrect: Describes the impact as the fuel injection pumps developing leaks due to low flow. There is a precaution that describes the effects of operating at low loads and the effects on the fuel injection pumps. However, there is no limit on load nor duration based on the injection pumps. The action described in part 2 is the reactive load directed for the EDG full load test and does not address the buildup of unburned exhaust products.

B. Incorrect: Describes the impact as the fuel injection pumps developing leaks due to low flow. There is a precaution that describes the effects of operating at low loads and the effects on the fuel injection pumps. However, there is no limit on load nor duration based on the injection pumps.

C. Incorrect: Has the correct impact, but the action described in part 2 is the reactive load directed for the EDG full load test and does not address the buildup of unburned exhaust products.

D. CORRECT: Per the note prior to step 6.5.1 of OP-009-002 and step 6.5.1, if the EDG is operated at < 50% load (2.2 MWe) for > 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , the load on the diesel must be raised in .5 MWe increments every 20 minutes until load is 4.0 - 4.4 MW and load must be held for at least 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

OP-009-002, Emergency Diesel Generator, Rev.

Technical Reference(s): 318, section 6.5 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-EDG00, Objective 2 (As available)

Question Source: Bank # X 2010 SRO Exam #4 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam 2010 SRO Exam #4 Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 5 Comments: Rev 1 - Changed the entire question to an approved question from the 2010 SRO exam. The original question was considered RO level knowledge.

Rev 2 - added direction statement from procedure to stem and editorial for clarity.

Rev 3, removed reference to MVAR and replaced with DC field current.

Facility: Waterford 3 Page 28 of 50

2012 NRC Exam SRO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 103 A2.04 Importance Rating 3.6*

K/A Statement A2.04 Ability to (a) predict the impacts of the following malfunctions or operations on the containment system and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Containment evacuation (including recognition of the alarm)

Proposed Question: SRO 15 Rev: 1 The following plant conditions exist:

Plant is in MODE 6

Full core offload is in progress in preparation to replace Steam Generators

The following alarms and indications are received:

CLASS 1E RAD MONITORING SYSTEM ACTIVITY HI-HI (Cabinet SA & SB)

RAD MONITORING SYSTEM ACTIVITY HI-HI (Cabinet L)

Plant Stack PIG Iodine Channels read 3.54E-02 ci/cc and rising

Plant Stack WRGM Effluent Rate reads 1.62E+06 ci/sec

Containment Purge Monitors ARM-IRE-5024(5026)(5027) are in Alert

Containment Purge Monitor ARM-IRE-5025 is in High Alarm

Refueling SRO reports that a fuel assembly has dropped from the grapple and fallen back into the core with bubbles rising from the area The crew will enter ___(1)___, and action should be directed to evacuate all ___(2)___.

(1) (2)

A. OP-901-403, High Airborne non-essential personnel from the site Activity in Containment only B. OP-901-403, High Airborne non-essential personnel from the site Activity in Containment and OP-901-405, Fuel Handling Incident concurrently C. OP-901-405, Fuel Handling personnel from Containment Incident only D. OP-901-403, High Airborne personnel from Containment Activity in Containment and OP-901-405, Fuel Handling Incident concurrently Facility: Waterford 3 Page 29 of 50

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

A. Incorrect: Step 12 directs concurrent entry into OP-901-405 if the cause of high activity is a leaking Fuel bundle. The Emergency Director determines the need for site evacuation based on their E-Plan classification, not the procedures used here.

The readings given would only result in an Alert classification, which would not require a site evacuation.

B. Incorrect: Alarms and indications match those listed in the Symptoms section of OP-901-403. Step 4 of OP-901-403 directs evacuation of Containment. Step 12 directs concurrent entry into OP-901-405 if the cause of high activity is a leaking Fuel bundle. The Emergency Director determines the need for site evacuation based on their E-Plan classification, not the procedures used here. The readings given would only result in an Alert classification, which would not require a site evacuation.

C. Incorrect: Alarms and indications match those listed in the Symptoms section of OP-901-403. Step 4 of OP-901-403 directs evacuation of Containment. Step 12 directs concurrent entry into OP-901-405 if the cause of high activity is a leaking Fuel bundle.

D. CORRECT: Alarms and indications match those listed in the Symptoms section of OP-901-403. Step 4 of OP-901-403 directs evacuation of Containment. Step 12 directs concurrent entry into OP-901-405 if the cause of high activity is a leaking Fuel bundle. OP-901-405 indications include Bubbles emerging from submerged, dropped, OR damaged Fuel Assembly.

Technical Reference(s): OP-901-403, High Airborne Activity in (Attach if not previously provided) Containment, Rev. 3 (including version/revision number) OP-901-405, Fuel Handling Incident, Rev. 5 Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PPO40, Obj. 1 & 4 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 5 Comments: (1) Combined the last two bullets in the conditions. (2) Made editorial corrections to the stem as requested.

Facility: Waterford 3 Page 30 of 50

2012 NRC Exam SRO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 2 K/A # 015 G2.1.7 Importance Rating 4.7 K/A Statement Conduct of Operations: 2.1.7 Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation.

Proposed Question: SRO 16 Rev: 2 The following plant conditions exist:

Plant was operating at 80% when CEA 32 dropped into the core

Crew has entered OP-901-102, CEA or CEDMCS Malfunction

Crew is lowering power to comply with TS 3.1.3.1 The CRS can stop the power reduction at ___(1)___ percent. The appropriate procedure for the power reduction is ______(2)______.

(1) (2)

A. 60 OP-901-212, Rapid Plant Power Reduction B. 60 OP-010-005, Plant Shutdown C. 50 OP-901-212, Rapid Plant Power Reduction D. 50 OP-010-005, Plant Shutdown Facility: Waterford 3 Page 31 of 50

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

A. CORRECT: COLR Figure 3 requires a power reduction of 30% after a single CEA deviation, unless power is reduced to 60%, then no further reduction is required.

Initial reactor power is 80%, therefore, the power reduction can be stopped at 60%. A rapid plant downpower is required for a dropped CEA, therefore, OP-901-212, Rapid Plant Power Reduction would be the correct procedure to use.

B. Incorrect. COLR Figure 3 requires a power reduction of 30% after a single CEA deviation, unless power is reduced to 60%, then no further reduction is required.

Initial reactor power is 80%, therefore, the power reduction can be stopped at 60%. A rapid plant downpower is required for a dropped CEA, therefore, OP-010-005, normal shutdown would not be the correct procedure to use.

C. Incorrect. The 50% value is credible for reactor power since reactor power is initially 80% and a 30% reduction per the curve would be 50%. Per the note, power reduction can be stopped at 60%. OP-901-212, Rapid Plant Power Reduction, is the correct procedure.

D. Incorrect. The 50% value is credible for reactor power since reactor power is initially 80% and a 30% reduction per the curve would be 50%. Per the note, power reduction can be stopped at 60%. OP-010-005, Plant Shutdown, is the wrong procedure selection.

Technical Reference(s): TS 3.1.3.1 and COLR Figure 3 OP-901-212, Rapid Plant Power Reduction, Rev (Attach if not previously provided) 004 OP-901-102, CEA or CEDMCS Malfunction, Rev.

(including version/revision number) 301 Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PPO10, Objective 3 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 5 Comments: Rev 1 - Changed part 2 of the question such that type of shutdown (rapid or normal) is determined. Changed part 1 such that the amount of power reduction required (20% or 30%) is determined for a dropped CEA when power is initially at 80%.

Rev 2 - Revised explanations to MW with % and corrected 901-212 title in explanations.

Facility: Waterford 3 Page 32 of 50

2012 NRC Exam SRO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 2 K/A # 033 A2.03 Importance Rating 3.5 K/A Statement A2.03 Ability to (a) predict the impacts of the following malfunctions or operations on the Spent Fuel Pool Cooling System ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Abnormal spent fuel pool water level or loss of water level Proposed Question: SRO 17 Rev: 1 The following plant conditions exist:

Plant is in MODE 6

Full core offload to the Spent Fuel Pool is in progress

The following alarms have been received at CP-2:

o Fuel Pool Temperature Hi o Fuel Pool Pumps Discharge Press Lo o Fuel Pool Level Lo Based on these indications, the CRS should enter ___(1)___ and suspend ___(2)___.

(1) (2)

A. OP-901-510, Component Cooling Water spent fuel movement System Malfunction and OP-901-513, Spent Fuel Pool Cooling Malfunction B. OP-901-510, Component Cooling Water core alterations System Malfunction and OP-901-513, Spent Fuel Pool Cooling Malfunction C. OP-901-513, Spent Fuel Pool Cooling spent fuel movement Malfunction only D. OP-901-513, Spent Fuel Pool Cooling core alterations Malfunction only Facility: Waterford 3 Page 33 of 50

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

A. Incorrect: Part 1, OP-901-510 is incorrect because it only addresses a loss of CCW Pump or indication of loss of cooling of the CCW system, which is not supported by the stem. Part 2 is correct based on OP-901-513, step 4 (requiring suspension of fuel movement in the pool).

B. Incorrect: Part 1, OP-901-510 is incorrect because it only addresses a loss of CCW Pump or indication of loss of cooling of the CCW system, which is not supported by the stem. Part 2 is incorrect because there are no corresponding indications of refueling cavity level below requirements or loss of CCW cooling to the Fuel Pool.

C. CORRECT: The combination of the Fuel Pool Level Lo and Fuel Pool Pumps Discharge Press Lo indicates that Fuel Pool level has dropped to the Fuel Pool Cooling Pumps lo level trip setpoint. Makeup is required to restore level and restart Fuel Pool Cooling. Fuel Pool Hi Temperature is due to loss of the cooling mechanism (Fuel Pool Cooling Pump flow lost). Annunciator response procedure directs the operator to OP-901-513. Part 2 is correct based on OP-901-513, step 4 (requiring suspension of fuel movement in the pool).

D. Incorrect: Part 1 is correct but part 2 is incorrect. Part 2 is incorrect because OP-901-513 directs suspending fuel movement in the spent fuel pool.

Technical Reference(s): OP-901-513, Spent Fuel Pool Cooling Malfunction (Attach if not previously provided) Rev. 006 (including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PPO50, Objective 3 (As available)

Question Source: Bank # X WF3-OPS-09491 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam 2010 NRC Exam SRO #7 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 5 Comments: Changed part 1 of the answers by adding OP-901-513 to A and B and added only to C and D. Changed part 2 of the answers by deleting the actions required for the offnormal and leaving only the required actions for fuel movement.

Facility: Waterford 3 Page 34 of 50

2012 NRC Exam SRO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 2 K/A # 016 K4.03 Importance Rating 2.9 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: SRO 18 Rev: 2 Given:

Plant is at 100% power

Annunciator VCT LEVEL LO-LO is received

ATC operator reports VCT level indicates 45% on CP-4

CVC-ILT-0227 indicates 0% on the PMC The CRS will enter ___(1)___. This failure results in ___(2)___.

(1) (2)

A. OP-901-113, Volume Control Tank RCS boration from the RWSP Makeup Control Malfunction B. OP-901-113, Volume Control Tank loss of auto makeup control to the Makeup Control Malfunction VCT C. OP-901-112, Charging or Letdown loss of auto makeup control to the Malfunction VCT D. OP-901-112, Charging or Letdown RCS boration from the RWSP Malfunction Facility: Waterford 3 Page 35 of 50

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

A. CORRECT: If level instrument CVC-ILT-0227 fails low, the CRS will enter OP-901-113 and this failure will cause the charging pump suction source to swap to the RWSP. (CVC-507 will open and CVC-118 will close)

B. Incorrect. The first part is correct. CVC-ILT-0227 has no effect on auto makeup to the VCT. If the other level instrument (CVC-ILT-0226) had failed, auto makeup would have been affected. The failure of CVC-ILT-0226 is also covered in OP-901-113.

C. Incorrect. OP-901-112 Charging or Letdown Malfunction is not the correct offnormal to enter even though the suction of the charging pumps has swapped. If the other level instrument (CVC-ILT-0226) had failed, auto makeup would have been affected.

The failure of CVC-ILT-0226 is also covered in OP-901-113.

D. Incorrect. OP-901-112 Charging or Letdown Malfunction is not the correct offnormal to enter even though the suction of the charging pumps has swapped. The second part is correct.

OP-901-113, Volume Control Tank Makeup Technical Reference(s): Control Malfunction, Rev. 1 (Attach if not previously provided)

(including version/revision number)

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

Objectives 1, 3 Question Source: Bank #

Modified Bank # X 08879 New Question History: Last NRC Exam 2009 SRO Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 5 Comments: Rev 2 - Changed the K/A and question to reduce the number of refuel questions.

Facility: Waterford 3 Page 36 of 50

2012 NRC Exam SRO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 3 Group # 1 K/A # G2.1.5 Importance Rating 3.9 K/A Statement Conduct of Operations: Ability to use procedures related to shift staffing, such as minimum crew complement, overtime limitations, etc.

Proposed Question: SRO 19 Rev: 1 The following plant conditions exist:

Reactor is shutdown with all CEAs fully inserted

RCS Boron concentration is 2100 ppm

Keff is 0.945

RCS temperature is 195°F Which of the following correctly describes the minimum required shift staffing in accordance with Tech Spec 6.2.2?

A. 1 SM, 1 SRO, 2 ROs, 1 STA and 2 NAOs B. 1 SM, 1 RO, and 1 NAO C. 1 SM, 1 RO, 1 STA and 1 NAO D. 1 SM, 1 SRO, 2 ROs and 2 NAOs Facility: Waterford 3 Page 37 of 50

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

A. Incorrect: This crew composition exceeds the requirement of T.S. 6.2.2, Table 6.2-1 for MODE 5 by 1 SRO, 1 RO and 1 STA.

B. CORRECT: This crew composition meets the requirement of T.S. 6.2.2, Table 6.2-1 for MODE 5.

C. Incorrect: This crew composition exceeds the requirement of T.S. 6.2.2, Table 6.2-1 for MODE 5 by 1 STA.

D. Incorrect: This crew composition exceeds the requirement of T.S. 6.2.2, Table 6.2-1 for MODE 5 by 1 SRO and 1 NAO.

Technical Reference(s): Technical Specification 6.2.2 and associated (Attach if not previously provided) Table 6.2-1 Minimum Shift Crew Composition (including version/revision number) Technical Specification table 1.2 Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-TS03, Objective 6 (As available)

Question Source: Bank # X WF3-OPS-7256-A Modified Bank # See Comments below.

New Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 5 Comments: (1) Changed Shift Manager to SM in the answers.

(2) Changed A and D to 2 ROs vice 1 RO. (3) Added TS table 1.2 to the references.

Facility: Waterford 3 Page 38 of 50

2012 NRC Exam SRO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 3 Group # 1 K/A # G2.1.37 Importance Rating 4.6 K/A Statement Conduct of Operations: Knowledge of procedures, guidelines, or limitations associated with reactivity management.

Proposed Question: SRO 20 Rev: 1 In accordance with EN-OP-115, Conduct of Operations, a second individual with no concurrent activities should be present during manual control rod insertion or withdrawal.

The CRS may waive the second individual requirement for control rod manipulations for

___(1)___ during ___(2)___.

(1) (2)

A. control rod insertion and withdrawal control of Axial Xenon oscillations during steady state operations B. control rod insertion and withdrawal implementation of abnormal or emergency operating procedures C. control rod insertion only control of Axial Xenon oscillations during steady state operations D. control rod insertion only implementation of abnormal or emergency operating procedures Facility: Waterford 3 Page 39 of 50

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

A. Incorrect: EN-OP-115 section 5.5 step 29 states that the CRS may waive the second individual requirements for control rod insertions during the implementation of abnormal or emergency operating procedures.

B. Incorrect: EN-OP-115 section 5.5 step 29 states that the CRS may waive the second individual requirements for control rod insertions (not for control rod withdrawal). The second portion of the answer is correct.

C. Incorrect: The first portion is correct but the second part is not correct because there is not a waiver allowed for Axial Xenon oscillations during steady state operations.

D. CORRECT: EN-OP-115 section 5.5 step 29 states that the CRS may waive the second individual requirements for control rod insertions during the implementation of abnormal or emergency operating procedures.

Technical Reference(s): EN-OP-115 Section 5.5 step 29, Rev. 13 (Attach if not previously provided) OP-010-004 section 9.4.2 Rev. 316 (including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-EXP00, Objective 3 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 6 Comments: Changed the question entirely due to the original question identified as not being a Tier 3 question.

Facility: Waterford 3 Page 40 of 50

2012 NRC Exam SRO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 3 Group # 2 K/A # G2.2.7 Importance Rating 3.6 K/A Statement Equipment Control: Knowledge of the process for conducting special or infrequent tests.

Proposed Question: SRO 21 Rev: 1 In accordance with EN-OP-116, Infrequently Performed Tests or Evolutions, an IPTE briefing is required to perform ___(1)___. The ___(2)___is responsible for ensuring proper performance of the IPTE Pre-job Brief.

(1) (2)

A. Reactor Coolant System Drain Down to Shift Manager 18 feet MSL B. Reactor Coolant System Drain Down to Senior Line Manager 18 feet MSL C. OP-903-007, Turbine Inlet Valve Shift Manager Cycling Test D. OP-903-007, Turbine Inlet Valve Senior Line Manager Cycling Test Facility: Waterford 3 Page 41 of 50

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

A. Incorrect: EN-OP-116, Attachment 9.1, Identified IPTEs, states that Reactor Coolant System Drain Down to Lowered Inventory is an IPTE. Lowered inventory is less than 20 feet, therefore, drain down to 18 feet is an IPTE. EN-OP-116 Responsibilities section 4.0 [8] states that the Senior Line Manager is responsible for ensuring proper performance of the IPTE Pre-job Brief, not the Shift Manager.

B. CORRECT: Lowered inventory to less than 20 feet is an IPTE. EN-OP-116 states that the Senior Line Manager is responsible for ensuring proper performance of the IPTE Pre-job Brief.

C. Incorrect: EN-OP-116 states that any test that actually over-speeds a turbine or Emergency Diesel Generator is an IPTE. OP-903-007 tests the cycling of Turbine valves, not the overspeed protection system. EN-OP-116 Responsibilities section 4.0

[8] states that the Senior Line Manager is responsible for ensuring proper performance of the IPTE Pre-job Brief, not the Shift Manager.

D. Incorrect: OP-903-007 does not overspeed the Main Turbine, and therefore is not an IPTE. EN-OP-116 Responsibilities section states that the Senior Line Manager is responsible for ensuring proper performance of the IPTE Pre-job Brief.

Technical Reference(s): EN-OP-116, Infrequently Performed Tests or (Attach if not previously provided) Evolutions, Rev. 009 OP-001-003, Reactor Coolant System Draindown, (including version/revision number) Rev. 312 limitation 3.2.8.

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-IPTE, Revision 2 (As available)

Objectives 1 and 4 Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 3 Comments: (1) Added in accordance with EN-OP-116 to the question. (2) Added OP-001-003, Reactor Coolant System Draindown, Rev. 312 limitation 3.2.8 as a reference.

(3) Deleted the initial from part 1 of answers A and B. Rev. 2 enhanced explanations.

Facility: Waterford 3 Page 42 of 50

2012 NRC Exam SRO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 3 Group # 3 K/A # G2.2.17 Importance Rating 3.8 K/A Statement Equipment Control: Knowledge of the process for managing maintenance activities during power operations, such as risk assessments, work prioritization, and coordination with the transmission system operator.

Proposed Question: SRO 22 Rev: 1 The following plant conditions exist:

Plant is in MODE 5

Battery Charger A1 must be removed from service for emergent work

Shutdown EOOS assigned an ORANGE risk level to the configuration resulting from removal of Battery Charger A1 In accordance with OI-037-000, Operations Risk Assessment Guideline, the ___(1)___

can authorize the request to remove Battery Charger A1 from service, and a qualitative assessment is ___(2)___.

(1) (2)

A. Duty Plant Manager required B. Duty Plant Manager NOT required C. Shift Manager required D. Shift Manager NOT required Facility: Waterford 3 Page 43 of 50

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

A. CORRECT: OI-037-000, Page 20, step 5.7.2.3 and bullet 1 state: Orange is considered high risk, with the following requirements: Duty Plant Manager approval for voluntary entry, or notification upon entry into emergent activities. Page 10, Step 5.2.1 states, Therefore, when one or more of these SSCs becomes unavailable due to emergent activities, an assessment which considers both the quantitative (EOOS tool) and qualitative (i.e., Level 2 and External Events) aspects of risk is performed.

B. Incorrect. DPM is the correct authority level, but a qualitative risk assessment is required per Step 5.2.1 above.

C. Incorrect. The Shift Manager is not the correct level of authority for Orange level risk assessment determination.

D. Incorrect. The Shift Manager is not the correct level of authority for Orange level risk assessment determination, and a qualitative risk assessment is required per Step 5.2.1 above.

Technical Reference(s): OI-037-000, Operations Risk Assessment (Attach if not previously provided) Guideline, Rev. 304 (including version/revision number)

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

Objectives 2 & 3 Question Source: Bank # X WF3-OPS-08384 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam 2008 NRC Exam Question #96 SRO Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 3 Comments: (1) Changed the first bullet in the conditions to emergent work vice scheduled work. (2) Added In accordance with OI-037-000, Operations Risk Assessment Guideline to the stem.

Facility: Waterford 3 Page 44 of 50

2012 NRC Exam SRO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 3 Group # 3 K/A # G2.3.14 Importance Rating 3.8 K/A Statement Radiation Control: Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or emergency conditions or activities.

Proposed Question: SRO 23 Rev: 3 Per TS 3.9.10.1, Water Level - Reactor Vessel, at least 23 feet of water shall be maintained over the top of the ___(1)___ during movement of irradiated fuel within the reactor pressure vessel.

This restriction ensures that in the event of a rupture of an irradiated fuel assembly, sufficient water depth is available such that ___(2)___.

(1) (2)

A. fuel seated in the reactor pressure sufficient cooling capacity is maintained vessel B. reactor pressure vessel flange iodine released is within limits C. reactor pressure vessel flange sufficient cooling capacity is maintained D. fuel seated in the reactor pressure iodine released is within limits vessel Facility: Waterford 3 Page 45 of 50

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

A. Incorrect: Twenty three feet of water above the fuel is credible since this is the required levels for determining the amount of shutdown cooling trains required per TS 3.9.8. Additionally, Tech Spec 3.9.10.2 states At least 23 feet of water shall be maintained over the top of the fuel seated in the reactor pressure vessel. This is applicable for CEA movement, not fuel movement. The basis for TS 3.9.10.1 is to ensure sufficient depth such that the iodine released is by a factor of at least 200.

Sufficient cooling capacity maintained is the basis for the 32 foot requirement in TS 3.9.8 (Shutdown Cooling)

B. CORRECT: Tech Spec 3.9.10.1 states At least 23 feet of water shall be maintained over the top of the reactor pressure vessel flange. The basis for TS 3.9.10.1 is to ensure sufficient depth such that the iodine released is reduced by a factor of at least 200.

C. Incorrect: Part 1 is correct but the basis for TS 3.9.10.1 is to ensure sufficient depth such that the iodine released is reduced by a factor of at least 200. Sufficient cooling capacity maintained is the basis for the 32 foot requirement in TS 3.9.8 (Shutdown Cooling)

D. Incorrect. Twenty three feet of water above the fuel is credible since this is the required levels for determining the amount of shutdown cooling trains required per TS 3.9.8. Additionally, Tech Spec 3.9.10.2 states At least 23 feet of water shall be maintained over the top of the fuel seated in the reactor pressure vessel. This is applicable for CEA movement, not fuel movement. Part 2 is correct.

Technical Reference(s): Technical Specification 3.9.10.1, Water Level -

(Attach if not previously provided) Reactor Vessel (including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-REQ04, Objective 2 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 4 Comments: Rev 1 - Replaced the entire question because the original question was not Tier 3.

Rev 2 - typographical and editorial for clarity.

Rev 3 - typographical Facility: Waterford 3 Page 46 of 50

2012 NRC Exam SRO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 3 Group # 4 K/A # G2.4.9 Importance Rating 4.2 K/A Statement Emergency Procedures / Plan: Knowledge of low power/shutdown implications in accident (e.g., loss of coolant accident or loss of residual heat removal) mitigation strategies.

Proposed Question: SRO 24 Rev: 2 In accordance with OP-010-006, Outage Operations, the most preferred location to place an irradiated fuel assembly that is in transit during a lowering of Refuel Cavity level is the _______(1)_______.

A. Reactor Vessel B. upender baskets with the upender horizontal C. Fuel Assembly Storage Rack in the Refuel Cavity D. Refuel Cavity deep end, with assembly suspended from the Refueling Machine Facility: Waterford 3 Page 47 of 50

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

A. CORRECT: Per OP-010-006, Attachment 9.23, step 9.23.2.4, the preferred location to place an irradiated fuel assembly in transit is listed in order of decreasing preference with the Reactor Vessel listed as the most preferred.

B. Incorrect: The Upender baskets with the Upender horizontal is listed as the second most preferred location.

C. Incorrect: The Fuel Assembly Storage Rack in the Refuel Cavity is listed as the third most preferred location.

D. Incorrect. The Cavity Deep end, with assembly suspended from the Refueling Machine is listed as the last most preferred location.

Technical Reference(s): OP-010-006, Outage Operations, Attachment (Attach if not previously provided) 9.23, Loss of Refuel Cavity Water Level (including version/revision number) Guidelines, step 9.23.2.4 Rev. 320 Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-REQ04 Objective 5 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 5 Comments: Changed the question completely since the original question was considered not a Tier 3 question.

Rev 2, typographical.

Facility: Waterford 3 Page 48 of 50

2012 NRC Exam SRO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 3 Group # 4 K/A # G2.4.28 Importance Rating 4.1 K/A Statement Emergency Procedures / Plan: Knowledge of procedures relating to a security event (non-safeguards information).

Proposed Question: SRO 25 Rev: 1 In accordance with OP-901-523, Security Events, the NRC shall be notified within 15 minutes from ______(1)_______ of a security-based event so that the NRC can provide prompt ________(2)_______.

(1) (2)

A. discovery notifications to other licensees and federal agencies B. discovery assistance in obtaining necessary resources C. Emergency Classification notifications to other licensees and federal agencies D. Emergency Classification assistance in obtaining necessary resources Facility: Waterford 3 Page 49 of 50

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

A. CORRECT: Per Note on Page 8 of OP-901-523, Security Events, The NRC Headquarters Operations Center shall be notified within 15 minutes of discovery of a security-based event to allow the NRC to more quickly notify other licensees and Federal agencies.

B. Incorrect: Within 15 minute of discovery is the required notification time, but the NRC is not responsible for obtaining necessary resources for security-based events.

C. Incorrect: The NRC Headquarters Operations Center (HOC) shall be notified within 15 minutes of discovery of a security-based event to allow the NRC to more quickly notify other licensees and Federal agencies. Delaying until E-Plan classification and notifications are complete could delay NRC HOC notifications by as much as 60 minutes.

D. Incorrect: The NRC Headquarters Operations Center (HOC) shall be notified within 15 minutes of discovery of a security-based event to allow the NRC to more quickly notify other licensees and Federal agencies. Delaying until E-Plan classification and notifications are complete could delay NRC HOC notifications by as much as 60 minutes. The NRC is not responsible for obtaining necessary resources for security-based events.

Technical Reference(s): OP-901-523, Security Events, Rev. 010 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PPO52, (As available)

Objectives 2 & 3 Question Source: Bank # X WF3-OPS-06733 Modified Bank #

New Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 5 Comments: Added In accordance with OP-901-523, Security Events to the stem.

Changed answers C and D from E-Plan to Emergency.

Facility: Waterford 3 Page 50 of 50

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