ML120880630

2012-03-Final-Written Exam

ML120880630
Person / Time
Site:	Diablo Canyon
Issue date:	03/09/2012
From:	Clyde Osterholtz Operations Branch IV
To:	Pacific Gas & Electric Co
laura hurley
References
Download: ML120880630 (131)

Text

U.S. Nuclear Regulatory Commission Diablo Canyon SRO Written Examination Applicant Information Name: KEY Date: 9 March, 2012 Facility/Unit: Diablo Canyon Region: I II III IV Reactor Type: W CE BW GE Start Time: Finish Time:

Instructions Use the answer sheets provided to document your answers. Staple this cover sheet on top of the answer sheets.

To pass the examination you must achieve a final grade of at least 80.00 percent overall, with 70.00 percent or better on the SRO-only items if given in conjunction with the RO exam; SRO-only exams given alone require a final grade of 80.00 percent to pass.

You have 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> to complete the combined examination, and 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> if you are only taking the SRO portion.

Applicant Certification All work done on this examination is my own. I have neither given nor received aid.

Applicants Signature Results RO/SRO-Only/Total Examination Values 75_ / 25_ / 100 Points Applicants Scores / / Points Applicants Grade / / Percent

DCPP NRC Exam 9 March 2012 Multiple Choice (Fill in your choice) NAME:___ ______________

If you change your answer, write your selection in the blank and initial.

A B C D 26. A B C D 1.

A B C D 27. A B C D 2.

A B C D 28. A B C D 3.

4. A B C D 29. A B C D

5. A B C D 30. A B C D

6. A B C D 31. A B C D A B C D 32. A B C D 7.

A B C D 33. A B C D 8.

A B C D 34. A B C D 9.

10. A B C D 35. A B C D

11. A B C D 36. A B C D

12. A B C D 37. A B C D A B C D 38. A B C D 13.

A B C D 39. A B C D 14.

A B C D 40. A B C D 15.

16. A B C D 41. A B C D

17. A B C D 42. A B C D

18. A B C D 43. A B C D A B C D 44. A B C D 19.

A B C D 45. A B C D 20.

A B C D 46. A B C D 21.

22. A B C D 47. A B C D

23. A B C D 48. A B C D

24. A B C D 49. A B C D A B C D 50. A B C D 25.

i

DCPP NRC Exam 9 March 2012 Multiple Choice (Fill in your choice) NAME:___ ______________

If you change your answer, write your selection in the blank and initial.

A B C D 76. A B C D 51.

A B C D 77. A B C D 52.

A B C D 78. A B C D 53.

54. A B C D 79. A B C D

55. A B C D 80. A B C D

56. A B C D 81. A B C D A B C D 82. A B C D 57.

A B C D 83. A B C D 58.

A B C D 84. A B C D 59.

60. A B C D 85. A B C D

61. A B C D 86. A B C D

62. A B C D 87. A B C D A B C D 88. A B C D 63.

A B C D 89. A B C D 64.

A B C D 90. A B C D 65.

66. A B C D 91. A B C D

67. A B C D 92. A B C D

68. A B C D 93. A B C D A B C D 94. A B C D 69.

A B C D 95. A B C D 70.

A B C D 96. A B C D 71.

72. A B C D 97. A B C D

73. A B C D 98. A B C D

74. A B C D 99. A B C D A B C D 100. A B C D 75.

ii

Examination Outline Cross-Reference Level RO Tier # 2 Knowledge of the operational implications of the following Group # 1 concepts as they apply to the RCPS: The dependency of RCS flow K/A # 003 K5.05 rates upon the number of operating RCPs Rating 2.8 Question 1 GIVEN:

• Unit 1 is in MODE 4

• Three RCPs are running Which of the following describes the expected RCS flow indication?

Idle Loop Flow Operating Loop Flow Indication Indication A. approximately 0% approximately 100%.

B. approximately 0% approximately 108%.

C. approximately 32% approximately 100%.

D. approximately 32% approximately 108%.

Proposed Answer: D. approximately 32%; approximately 108%

Explanation:

Flow in the operating loops increases to 108%. Flow in the idled loop indicates 32%, which is actually -32% (reverse flow). If an RCP is stopped, the high pressure discharge of the other three pumps forces the coolant to flow backwards through the newly idled loop.

Due to the reduced backpressure of the idle loop, the flow rate in each of the operating loops increases from 100% to about 108% of rated flow. The idle loop has an equivalent flow of about 32%, in the reverse direction. Total core flow will decrease since only three RCPs are operating, and some of their flow is bypassing the core through the idle loop.

A and B incorrect. If candidate knows there is reverse flow and bottom of scale is zero, then it would be plausible to think that the meter would read downscale, or zero.

C incorrect. Indicated flow in the loops with running pumps is greater than 100%.

D is correct. Indicated flow in the idle loop, is the reverse flow indication of approximately 32% and flow in the operating loops is greater than 100% or approximately 108%

Technical

References:

TH18T, Pages 28 and 29 and Table 18-3 and LTH-18 (note, lesson states: "lesson guide will be used with fundamentals text TH18T, Transient Analysis.)

References to be provided to applicants during exam: None Learning Objective: 10583 - DESCRIBE the reactor, RCS and Secondary System responses to each of the following transients:

e. Stopping a Reactor Coolant Pump (RCP) with no resultant reactor trip.

Question Source: Bank #

(note changes; attach parent) Modified Bank #

DCPP L091C Exam Rev 1

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.41.14 - Principals of heat transfer, thermodynamics and fluid mechanics.

Modified justification to clearly identify why A, B and C is incorrect.

DCPP L091C Exam Rev 1

Examination Outline Cross-Reference Level RO Tier # 2 Knowledge of the effect that a loss or malfunction of the RCPS Group # 1 will have on the following: Feedwater and emergency feedwater K/A # 003 K3.03 Rating 2.8 Question 02 Unit 1 is at 25% power.

The breaker for RCP 1-1 trips. The reactor does not trip.

Which of the following describes the effect on feed flow to the Steam Generators?

A. Feed flow goes down in all Steam Generators.

B. Feed flow in the Steam Generators 1-2, 1-3 and 1-4 is unchanged; feed flow in Steam Generator 1-1 goes down.

C. Feed flow in the Steam Generators 1-2, 1-3 and 1-4 goes up; feed flow in Steam Generator 1-1 goes down.

D. Feed flow in the Steam Generators 1-2, 1-3 and 1-4 goes down; feed flow in Steam Generator 1-1 goes up.

Proposed Answer: C. Feed flow in the Steam Generators 1-2, 1-3 and 1-4 goes up; feed flow in Steam Generator 1-1 goes down Explanation:

A. Incorrect. increased steaming from the unaffected loop will cause feed flow to rise in those loops. Conversely, decreased steaming in the affected loop causes feed flow to go down. If the candidate believes power will go down, then it would be reasonable to assume feed flow will go down in all loops.

B. Incorrect. power in the unaffected loops goes up, but overall reactor power does not change. Failure to realize this makes the answer plausible.

C. Correct.

D. Incorrect. opposite of what actually happens.

Technical

References:

TH18T, Pages 28 and 29 and Table 18-3 and LTH-18 (note, lesson states: "lesson guide will be used with fundamentals text TH18T, Transient Analysis.)

References to be provided to applicants during exam: None Learning Objective: 10583 - DESCRIBE the reactor, RCS and Secondary System responses to each of the following transients:

e. Stopping a Reactor Coolant Pump (RCP) with no resultant reactor trip.

Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.41.14 - Principals of heat transfer, thermodynamics and fluid mechanics.

DCPP L091C Exam Rev 1

Removed statement "and power remains at 25%" from question setup.

DCPP L091C Exam Rev 1

Examination Outline Cross-Reference Level RO Tier # 2 004 K5.50 - Knowledge of the operational implications of the Group # 1 following concepts as they apply to the CVCS: Design basis K/A # 004 K5.50 letdown system temperatures: resin integrity Rating 2.6 Question 03 Unit 1 is at full power.

Temperature at the outlet of the Letdown Heat Exchanger begins to rise.

If letdown temperature continues to rise, an automatic action will occur to prevent which of the following?

A. Positive reactivity excursion B. Overheating of the CCW system C. Damage to the Letdown Demineralizer resin D. Flashing at the outlet of the Regenerative Heat Exchanger Proposed Answer: C. Damage to the Letdown Demineralizer resin Explanation:

A. Incorrect. High temperature would release Boron from a saturated bed and lower power.

B. Incorrect. CCW cools the letdown passing thru the Letdown heat exchanger. There are no automatic actions associated with CCW and letdown.

C. Correct. The purpose of the Letdown Temperature Divert Valve, TCV-149, is to protect the demineralizer resin from damage when letdown temperature exceeds (136F).

D. Incorrect. Flashing at the outlet of the Regen Heat Exchanger results from too little charging flow thru the heat exchanger. There are no automatic actions that occur, only operator action.

Technical

References:

LB-1A References to be provided to applicants during exam: None Learning Objective: 41114 - Describe CVCS components.

Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.41.7 - Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Added "letdown" to question to specify what temperature the question addresses.

DCPP L091C Exam Rev 1

Examination Outline Cross-Reference Level RO 005 K4.05 - Knowledge of RHRS design feature(s) and/or Tier # 2 interlock(s) which provide or the following: Relation between Group # 1 RHR flowpath and refueling cavity K/A # 005 K4.05 Rating 2.5 Question 04 The crew is preparing to fill the refueling cavity using Containment Spray pump 1-1 thru CS 9003A, RHR Pp 1 to Spray Hdr "A", to the RCS hot legs in accordance with OP B-2:II, RHR -

Filling the Refueling Cavity.

Which of the following is a reason a jumper must be installed to allow opening of CS-1-9003A?

A. Containment recirc sump valve, SI-1-8982A, is closed.

B. Spray additive tank outlet valve, CS-1-8992, is closed.

C. Reactor Coolant Loop 4 Outlet valves to RHR, RHR-1-8701 and RHR-1-8702, are closed.

D. RHR System Return to RCS Hot Legs Loops 1 and 2, RHR-1-8703, is open.

Proposed Answer: A. Containment recirc sump valve, SI-1-8982A is closed.

Explanation:

A. Correct. As a permissive to open CS 9003A, either RHR 8701 or RHR 8702 must be CLOSED, SI 8982A must be OPEN. In this lineup, 8701 and 8702 are open, and, the sump recirc valve is closed. Therefore, in order to open 9003A, a jumper to bypass the interlock is required.

B. Incorrect. The spray additive tank outlet valve does not impact the operation of the containment spray valve.

C. Incorrect. the valves being closed is the normally required position.

D. Incorrect. not part of the valve interlock, but is opened for this lineup.

Technical

References:

OP B-2:II, section 6.9.

References to be provided to applicants during exam: None Learning Objective: 35317 - Analyze automatic features and interlocks associated with the RHR system Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.41.7 Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Added valve name to 9003A. Reformatted D to match A, B and C. Corrected justification order.

DCPP L091C Exam Rev 1

Examination Outline Cross-Reference Level RO Tier # 2 006 A1.19 - Ability to predict and/or monitor changes in Group # 1 parameters (to prevent exceeding design limits) associated with K/A # 006 A1.19 operating the ECCS controls including: subcooling Rating 4.0 Question 05 The crew is performing the actions of E-1.2, Post LOCA Cooldown and Depressurization. RCS temperature is 300°F.

When evaluating conditions to stop the first SI pump, there is less subcooling than required. In accordance with the RNO step in E-1.2, the operator starts an RHR pump and then stops the SI pump.

According to the background document for E-1.2, what is accomplished by starting an RHR pump prior to stopping the SI pump?

A. Ensures the RCS will remain subcooled as RCS pressure lowers.

B. Contributes to the ability to refill the Pressurizer.

C. Prevents a severe challenge to the Core Cooling critical safety function.

D. Maintains conditions for continued operation of at least one RCP.

Proposed Answer: A. Ensures the RCS will remain subcooled as RCS pressure lowers.

Explanation:

A. Correct. The background document states: If the RCS subcooling criterion is not satisfied, but the RCS hot leg temperatures are less than the saturation temperature corresponding to the low-head (RHR) SI pump head at minimum pump recirculation flow, the charging/SI pump can be stopped if a low-head SI pump is running or can be started. Starting a low-head pump for this case ensures that RCS subcooling will be maintained after the charging/SI pump is stopped.

Technical

References:

E-1.2 background, E-1.2 step 16, LPE-1B References to be provided to applicants during exam: None Learning Objective: 6855 - Explain the effect of SI termination and/or reinitiation on control of the plant Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.41.5 - Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating DCPP L091C Exam Rev 1

limitations and reasons for these operating characteristics.

Added "According to the background document for E-1.2" to question for added focus.

DCPP L091C Exam Rev 1

Examination Outline Cross-Reference Level RO Tier # 2 007 K4.01 - Knowledge of PRTS design feature(s) and/or Group # 1 interlock(s) which provide for the following: Quench tank cooling K/A # 007 K4.01 Rating 2.6 Question 06 PK05-25, PRT Press/Lvl/Temp alarms. The operator reports it is due to high PRT temperature.

Level and pressure are within the normal range.

In accordance with PK05-25, which of the following is the initial action to take to lower PRT temperature?

A. Lower RCS pressure by opening the Pressurizer spray valves.

B. Manually opening the Primary Water system supply valve.

C. Vent the PRT to the Waste Gas Header.

D. Drain the PRT to the RCDT.

Proposed Answer: B. Manually opening the Primary Water system supply valve.

Explanation:

A incorrect. With the safeties and PORVs closed, lowering RCS pressure will not lower PRT pressure.

B correct. Temperature is lowered by manually opening RCS-1-8030 on VB2.

C and D incorrect. Pressure and level are normal - draining or venting is not necessary.

Technical

References:

AR PK05 PRT Press/Lvl/Temp References to be provided to applicants during exam: None Learning Objective: 4950 - Explain the operation of PRT system Question Source: Bank # NRC Exam 4/2007 X (note changes; attach parent) Modified Bank #

New Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content: 55.41.7 - Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Reformatted question, added "initial" to question to eliminate draining, which is done if level is high after the PRT is cooled.

Changed A based on NRC feedback.

DCPP L091C Exam Rev 1

Examination Outline Cross-Reference Level RO Tier # 2 008 K1.04 - Knowledge of the physical connections and/or cause- Group # 1 effect relationships between the CCWS and the following K/A # 008 K1.04 systems: RCS, in order to determine source(s) of RCS leakage Rating 3.3 into the CCWS Question 07 Unit 1 is at full power.

CCW surge tank level is rising. High radiation has caused RCV-16, the surge tank vent, to close.

Which of the following components is the most likely source of leakage into the CCW system?

A. ECCS Charging pumps B. Letdown heat exchanger C. Seal water heat exchanger D. Spent Fuel Pool heat exchanger Proposed Answer: B. Letdown heat exchanger Explanation:

A. Incorrect. AP-11 lists the following as possible sources:

Letdown heat exchangers, RHR heat exchangers, RCP thermal barrier, excess letdown heat exchanger, RHR pump seal coolers, Pressurizer sample coolers. The portion of the charging pumps cooled by CCW is the motor cooler.

B. Correct. The letdown heat exchanger is at a higher pressure than CCW and will cause the radiation monitor to alarm.

C. Incorrect. Seal water heat exchanger would not flow into the CCW system, it is at VCT pressure.

D. Incorrect. The spent fuel pool is not a potential source of in-leakage.

Technical

References:

OP AP-11 References to be provided to applicants during exam: None Learning Objective: 40500 - Describe the major flow path of the CCW system Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.41.7 Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Removed distractor A, replaced w/SFP heat exchanger. Reordered answers DCPP L091C Exam Rev 1

Examination Outline Cross-Reference Level RO Tier # 2 010 K5.01 - Knowledge of the operational implications of the Group # 1 following concepts as the apply to the PZR PCS: Determination K/A # 010 K5.01 of condition of fluid in PZR, using steam tables Rating 3.5 Question 08 GIVEN:

• TI-653, Pressurizer Liquid temperature, on VB2, indicates 645°F

• TI-654, Pressurizer Vapor temperature, on VB2, indicates 655°F

• Pressurizer pressure is 2270 psig Based on the temperature indications, what is the status of the water and steam (vapor) in the Pressurizer at this time?

A. The water is at saturation.

B. The steam is at saturation.

C. The water is subcooled by approximately 5°F.

D. The steam is superheated by approximately 5°F.

Proposed Answer: B. The steam is at saturation.

Explanation:

A. Incorrect. Based on the liquid temperature, the water is subcooled by approximately 10F.

B. Correct. The saturation pressure is based on the 2270 psig (2285 psia) which has a corresponding saturation temperature of 655°F.

C. Inorrect. the water is subcooled but by 10°F.

D. Incorrect. the steam is at saturation, if the assumption is that the state of the fluids is based on the liquid temperature, and an error is made in determing psia, then the steam would be indicating superheated conditions of approximately 5°F.

Technical

References:

steam tables, OVID 106707 sheet 3 References to be provided to applicants during exam: steam tables Learning Objective:

Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.41.14 Principals of heat transfer, thermodynamics and fluid mechanics.

Modified answers to remove implausible distractors.

DCPP L091C Exam Rev 1

Examination Outline Cross-Reference Level RO Tier # 2 012 K2.01 - Knowledge of bus power supplies to the following: Group # 1 RPS channels, components, and interconnections K/A # 012 K2.01 Rating 3.3 Question 09 A loss of which of the following power supplies would result in a failure of SSPS Train B to actuate if necessary?

A. 120 Vital Instrument Bus 12 due to loss of power to the slave relays B. 120 Vital Instrument Bus 14 due to loss of power to the slave relays C. 120 Vital Instrument Bus 12 due to loss of power to the master relays D. 120 Vital Instrument Bus 14 due to loss of power to the master relays Proposed Answer: B. 120 Vital Instrument Bus 14 due to loss of power to the slave relays Explanation:

A. Incorrect. Train A of SSPS slave relays are powered from PY 11, its logical to think PY12 is Train B.

B. Correct. Train B slave relays are powered from PY14. Loss of the PY would prevent the relays from energizing.

C. Incorrect. DC powered from SSPS via PY 13 and 14.

D. Incorrect. DC powered from SSPS via PY 13 and 14, however, because there are two sources, loss of one would not inhibit the master relays from actuating.

Technical

References:

OIM B-6-1b References to be provided to applicants during exam: None Learning Objective: 3291 - State the power supplies to Eagle 21 and Solid State Protection System components.

Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.41.7 Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features Replaced original question, which was moved to question 34.

DCPP L091C Exam Rev 1

Examination Outline Cross-Reference Level RO Tier # 2 013 K6.01 - Knowledge of the effect of a loss or malfunction of Group # 1 the following will have on the ESFAS: Sensors and detectors K/A # 013 K6.01 Rating 2.7 Question 10 GIVEN:

• Unit 2 is at 100% power

• Containment pressure channel IV, (PT-934), is declared inoperable

• Containment pressure channels I, II and III, (PT937, 936, and 935), remain OPERABLE

• Required actions per OP AP-5, U1&2, Malfunction of Eagle 21 Protection, have been completed

• To meet the Technical Specification LCO for Channel PT-934:

o "High Containment" pressure bistable is placed in "TRIP" o "High-High Containment Pressure" bistable is placed in "BYPASS" The MINIMUM remaining coincidences for ESF actuations from the OPERABLE Containment pressure channels are:

A. 1/2 for SI actuation and 2/2 for Containment Spray actuation B. 1/2 for SI actuation and 2/3 for Containment Spray actuation C. 2/3 for SI actuation and 1/2 for Containment Spray actuation D. 2/3 for SI actuation and 2/3 for Containment Spray actuation Proposed Answer: B. 1/2 for SI actuation and 2/3 for Containment Spray actuation Explanation:

A. Incorrect. High Containment Pressure SI actuation is 2 of 3 channels and Containment Spray is 2 of 4 channels. The remaining coincidence is then 1 of 2 for SI and 2 of 3 for Containment Spray because the bistable for the failed channel is in bypass, not trip. This answer could be true if it is thought that both are 2 of 3, with one in bypass, it would require the remaining 2 to cause actuation.

B. Correct. High pressure is the SI bistable, and in trip. the remaining coincidence is 1 of the 2 remaining channels. High-high is normally 2 of 4. With one channel in BYPASS, the coincidence is reduced to 2 of 3.

C. Incorrect. SI coincidence is normally 2 of 3, while Containment Spray is 2 of 4. If believes the logic is 2/4 and 2/3, (the logic is reversed), this answer is correct..

D. Incorrect. This would be correct if the candidate assumes the 4 channels input to both signals. This would cause actuation, but is not the minimum coincidence.

Technical

References:

OIM pages B-6-5 and B-6-8, Technical Specification 3.3.2 References to be provided to applicants during exam: None Learning Objective: 41313 - Analyze automatic features and interlocks associated with the Eagle-21/SSPS Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X DCPP L091C Exam Rev 1

Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.41.7 Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Removed 1/3 as a distractor. Minimum all caps.

DCPP L091C Exam Rev 1

Examination Outline Cross-Reference Level RO Tier # 2 013 G2.4.9 - ESFAS: Knowledge of low power/shutdown Group # 1 implications in accident (e.g., loss of coolant accident or loss of K/A # 013 G2.4.9 residual heat removal) mitigation strategies. Rating 3.8 Question 11 GIVEN:

• Unit 1 is performing a heatup in accordance with OP L-1, Plant Heatup From Hot Shutdown to Hot Standby

• Electrical power is aligned to backfeed from the 500 kV switchyard

• Startup Transformer 1-1 is cleared

• All RCPs are running

• RCS pressure is 1900 psig

• RCS temperature is 525°F A steam break, upstream of the MSIV occurs on the 1-1 Steam Generator.

Which of the following describes how Safety Injection will be actuated and the response of the motor driven AFW pumps?

A. SI actuates on low RCS pressure; AFW pumps will remain running.

B. SI actuates on low steam generator pressure; AFW pumps stop and restart when sequenced on to their respective Emergency Diesel Generator.

C. SI will not actuate automatically and will have to be manually actuated by the operator; AFW pumps stop and restart when sequenced on to their respective Emergency Diesel Generator.

D. SI will not actuate automatically and will have to be manually actuated by the operator; AFW will remain running.

Proposed Answer: C. SI will not actuate automatically and will have to be manually actuated by the operator; AFW pumps stop and restart when sequenced on to their respective Emergency Diesel Generator.

Explanation:

A. Incorrect. SI on low RCS pressure is blocked below P-11 (1915 psig).

B. Incorrect. SI on low steam generator pressure is blocked below P-11 (1915 psig).

C. Correct. SI will have to be manually actuated, when the SI causes a transfer to diesel, the motor driven AFW pumps stop and then are sequenced on to the Emergency Diesel Generators.

D. Incorrect. The AFW pumps will stop on the transfer to diesel. This would be true if the SI did not cause a transfer. Normally, the SI would cause a transfer to startup (and the pumps would remain running). However, with Startup cleared, a transfer to diesel will occur.

Technical

References:

OIM B-6-2, B-6-5 and C-2-1 References to be provided to applicants during exam: None Learning Objective: 41313 - Analyze automatic features and interlocks associated with the Eagle-21/SSPS Question Source: Bank #

(note changes; attach parent) Modified Bank #

DCPP L091C Exam Rev 1

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.41.8 Components, capacity, and functions of emergency systems.

Added that Startup power is cleared to cause a transfer to diesel to occur. This causes the AFW pumps to stop and wait for a transfer to diesel. Adjusted distractors accordingly for balance.

DCPP L091C Exam Rev 1

Examination Outline Cross-Reference Level RO Tier # 2 022 A2.04 Ability to (a) predict the impacts of the following malfunctions or operations on the CCS; and (b) based on those Group # 1 predictions, use procedures to correct, control, or mitigate the K/A # 022 A2.04 consequences of those malfunctions or operations: Loss of service Rating 2.9 water (Note: At DCPP, ASW is the equivalent system to SWS)

Question 12 The crew is aligning Unit 1 for Cold Leg Recirculation in accordance with E-1.3, Transfer to Cold Leg Recirculation. All CFCUs are running in "LOW".

Only one train of ASW is available.

In accordance with E-1.3, what action should the crew take regarding the CFCUs?

A. Leave FOUR of the CFCUs running in LOW and stop ONE of the CFCUs to prevent exceeding CCW system temperature design limit.

B. Leave THREE of the CFCUs running in LOW and stop TWO of the CFCUs to prevent exceeding CCW system temperature design limit.

C. Shift THREE of the CFCUs from LOW to HIGH to increase heat removal from the CCW system.

D. Shift ALL of the CFCUs from LOW to HIGH to increase heat removal from the CCW system.

Proposed Answer: B. Leave THREE of the CFCUs running in LOW and stop TWO of the CFCUs to prevent exceeding CCW system temperature design limit.

Explanation:

A. Incorrect. Two CFCUs are secured.

B. Correct. The heat load of all the loads on the CCW with only one ASW train could cause the system to not be able to meet its purpose to remove heat from the CCW system and the CCW system could exceed its design temperature Three remain running to remove heat from Containment and two are stopped.

C. Incorrect. None of the CFCUs are shifted to HIGH.The CFCUs are left in LOW to prevent possibly tripping on overcurrent.

D. Incorrect. None of the CFCUs are shifted to HIGH.The CFCUs are left in LOW to prevent possibly tripping on overcurrent.

Technical

References:

E-1.3, LF-2, LPE-1C References to be provided to applicants during exam: None Learning Objective: 35490 - Discuss abnormal conditions associated with the CCW System.

Question Source: Bank #

(note changes; attach parent) Modified Bank # DCPP NRC Exam 7/2011 X New Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental DCPP L091C Exam Rev 1

Comprehensive/Analysis X 10CFR Part 55 Content: 55.41.5 Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.

Reworded answer and distractors. Added IAW E-1.3 to question.

DCPP L091C Exam Rev 1

Examination Outline Cross-Reference Level RO Tier # 2 022 G2.4.34 Containment Cooling: Knowledge of RO tasks Group # 1 performed outside the main control room during an emergency K/A # 022 G2.4.34 and the resultant operational effects. Rating 4.2 Question 13 GIVEN:

• The Control Room has been evacuated

• The crew has entered OP AP-8A, Control Room Inaccessibility - Establishing Hot Standby

• The electrical buses have transferred to start-up

• The crew has completed Appendix F, 480V Bus Alignment The operator is placing the Transfer Switches at the Hot Shutdown Panel in LOCAL.

Which of the following will occur when the operator places the Transfer Switch for a running CFCU in LOCAL?

A. The CFCU stops and then AUTOMATICALLY restart in LOW speed.

B. The CFCU stops and then AUTOMATICALLY restart in HIGH speed.

C. The CFCU stops and now can ONLY be started in LOW at the Hot Shutdown Panel.

D. The CFCU stops and now can ONLY be started in HIGH at the Hot Shutdown Panel.

Proposed Answer: D. The CFCU stops and now can ONLY be started in HIGH at the Hot Shutdown Panel.

Explanation:

A. Incorrect. The CFCU only operates in HIGH from the HSDP and when the operator takes the switch to LOCAL, the fan will stop.

B. Incorrect. The fan can be started and will run in HIGH but going to LOCAL stops the fan.

C. Incorrect. The fan stops but will only run in HIGH when started by the operator.

D. Correct. The fan stops and can be started by taking the control switch to ON, which starts the fan in HIGH.

Technical

References:

LH-2, OP AP-8A References to be provided to applicants during exam: None Learning Objective: 6143 - Analyze automatic features and interlocks associated with the CFCUs Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content: 55.41.7 Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features Editorial changes to answer and distractors.

DCPP L091C Exam Rev 1

Examination Outline Cross-Reference Level RO Tier # 2 026 A4.05 Ability to manually operate and/or monitor in the Group # 1 control room: Containment Spray Reset switches K/A # 026 A4.05 Rating 3.5 Question 14 GIVEN:

• A LOCA occurs causing Containment pressure to peak at 25 psig

• Containment Spray and Phase B have actuated

• PK01-18, CONTMT SPRAY ACTUATION is in alarm

• Current Containment pressure is 18 psig The operator presses both Containment Spray Reset pushbuttons.

Which of the following describes the status of PK01-18 and the red Phase B lights?

A. PK01-18 remains in alarm; the Phase B red lights remain lit.

B. PK01-18 resets and immediately reflashes when the operator releases the pushbuttons; the Phase B red lights remain lit.

C. PK01-18 resets and remains out; the Phase B red lights remain lit.

D. PK01-18 resets and remains out; the Phase B red lights go out.

Proposed Answer: C. PK01-18 resets and remains out; the Phase B red lights remain lit.

Explanation:

A. Incorrect. Spray will reset and alarm will reset, (even if above the setpoint).

B. Incorrect. Spray will reset and alarm will reset, (even if above the setpoint).

C. Correct. Containment spray reset is a latch, which will reset the spray alarm and clear the alarm. Phase B has the same setpoint as Containment Spray but has its own reset and is not affected by resetting spray. Components could be stopped. Manual actuation will still cause spray to actuate.

D. Incorrect. While the containment spray components are on the Monitor Light box for Phase B, the Phase B red lights only come on/off when Phase B is actuated/reset.

Technical

References:

LB-6A, LI-2 References to be provided to applicants during exam: None Learning Objective: 37578 - Describe controls, indications, and alarms associated with the CSS Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.41.7 Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features Editorial changes to remove "if any" from question.

DCPP L091C Exam Rev 1

Examination Outline Cross-Reference Level RO Tier # 2 039 K4.02 Knowledge of MRSS design feature(s) and/or Group # 1 interlock(s) which provide for the following: Utilization of T-ave. K/A # 039 K4.02 program control when steam dumping through atmospheric Rating 3.1 relief/dump valves, including T-ave. limits Question 15 GIVEN:

• Unit 1 trips from full power

• The Steam Dump system responds as designed

• All electrical buses are energized from startup power Which of the following causes the Group IV steam dump valves to open and what is the approximate temperature the Steam Dump System should stabilize the RCS?

A. Reactor trip controller; 547°F B. Individual pressure controller; 547°F C. Reactor trip controller; 551°F D. Individual pressure controller; 551°F Proposed Answer: B. Individual pressure controller; 547°F Explanation:

A. Incorrect. Reactor trip controller blocks 10% valves from opening. They will operate from steam pressure.

B. Correct. The individual pressure controllers will operate, the reactor trip controller is blocked. Dumps will close at approximately 1000 psig, (547F)

C. Incorrect. For a load rejection controller RCS Tave steam dumps are closed 4°F above program Tave. Note: Tave is programmed from 547 to 570.5°F. Final Tave (program) would be 551°F.

D. Incorrect. The individual pressure controllers are set to control at no load Tave Technical

References:

LC-2B, OIM C-2-5 References to be provided to applicants during exam: None Learning Objective: 9993 - Explain the operation of the Steam Dump System Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.41.6 Design, components, and function of reactivity control mechanisms and instrumentation Editorial change to question, answer and distractors DCPP L091C Exam Rev 1

Examination Outline Cross-Reference Level RO Tier # 2 059 A4.08 Ability to manually operate and monitor in the control Group # 1 room: Feed regulating valve controller K/A # 059 A4.08 Rating 3.0 Question 16 Unit 1 is at 70% power.

MFW Pump 1-2 trips.

Which of the following actions initially occur?

A. Feed Reg Valve controller will throttle the valves open and running MFW pump controller increases pump speed.

B. Feed Reg Valve controller will throttle the valves closed and running MFW pump controller decreases pump speed.

C. Running MFW pump controller decreases pump speed causing the Feed Reg Valve controller to throttle the valves open.

D. Running MFW pump controller increases pump speed causing the Feed Reg Valve controller to throttle the valves closed.

Proposed Answer: A. Feed Reg Valve controller will throttle the valves open and running MFW pump controller increases pump speed Explanation:

A. Correct. The resulting shrink causes the valves to open. DP will lower and the loss of flow due to the loss of the MFP, causes the speed of the remaining pump to increase.

B. Incorrect. power lowers and at steady state, there would be less demand for flow, but the shrink due to the ramp causes the valves to open and pump speed to increase.

C. Incorrect. Pump speed increases as a result of lowering DP.

D. Incorrect. this would occur at steady state, an increase in pump speed would cause the reg valves to close down to maintain level.

Technical

References:

LPA25 References to be provided to applicants during exam: None Learning Objective: 3477 - Given an abnormal condition, summarize the major actions of the abnormal operating procedure to mitigate an event in progress.

Question Source: Bank #

(note changes; attach parent) Modified Bank # P-0103 X New Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content: 55.41.5 Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, DCPP L091C Exam Rev 1

effects of load changes, and operating limitations and reasons for these operating characteristics.

Removed "program ramp initiates" from setup.

DCPP L091C Exam Rev 1

Examination Outline Cross-Reference Level RO Tier # 2 059 A2.11 Ability to (a) predict the impacts of the following Group # 1 malfunctions or operations on the MFW; and (b) based on those K/A # 059 A2.11 predictions, use procedures to correct, control, or mitigate the Rating 3.0 consequences of those malfunctions or operations: Failure of feedwater control system Question 17 GIVEN:

• Unit 1 at 100% power

• Loop 2 Feedwater Temperature Transmitter, TT-1191, has been removed from service in the Digital Feedwater Control System (DFWCS)

Loop 4 Feedwater Temperature Transmitter, TT-1190, fails low Which of the following occurs and what action, if any, will be taken by the operator to control steam generator levels?

A. The DFWCS will use the remaining good input to automatically control level; no operator action required.

B. All main feed reg valves and bypass valves switch to MANUAL; the operators will manually control steam generator levels until at least one of the channels is returned to service.

C. The DFWCS will use the highest of the two remaining good inputs to automatically control level; no operator action required.

D. The main feed pumps switch to MANUAL; the operators will manually control steam generator levels until at least one of the channels is returned to service.

Proposed Answer: B. All main feed reg valves and bypass valves switch to MANUAL; the operators will manually control steam generator levels until at least one of the channels is returned to service.

Explanation:

A. Incorrect. If 2 or more of the three inputs are lost, the main feed valves fail to manual.

B. Correct. Loss of 2 or more channels causes the system to switch to manual, manual operator action will be required.

C. Incorrect. While there are four loops, only 3 have temperature instrumentation, loops 2, 3, and 4.

D. Incorrect. The temperature input is not an input into the feed pump control.

Technical

References:

LC-8B References to be provided to applicants during exam: None Learning Objective: 37642 - Discuss abnormal conditions associated with the DFWCS.

Question Source: Bank # S-27024 X (note changes; attach parent) Modified Bank #

New Question History: Last NRC Exam No DCPP L091C Exam Rev 1

Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.41.7 Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Removed loops in B and C (not necessary) and editorial change to second bullet (the).

DCPP L091C Exam Rev 1

Examination Outline Cross-Reference Level RO Tier # 2 061 K2.01 Knowledge of bus power supplies to the following: Group # 1 AFW system MOVs K/A # 061 K2.01 Rating 3.2 Question 18 A loss of which of the following power supplies would prevent the 1-1 Turbine Driven AFW pump from automatically starting, if required?

A. 480 VAC Bus F B. 480 VAC Bus H C. 125 VDC Bus 1-1 D. 125 VDC Bus 1-2 Proposed Answer: D. 125 VDC Bus 1-2 Explanation:

A. Incorrect. AFW pumps 2, and 3 are powered from Bus F and H (4 kv), Turbine pump steam supply valves, FCV-37 and 38 are powered from 480 VAC buses F and H. However, either one can supply the pump.

B. Incorrect. AFW pumps 2, and 3 are powered from Bus F and H (4 kv), Turbine pump steam supply valves, FCV-37 and 38 are powered from 480 VAC buses F and H. However, either one can supply the pump.

C. Incorrect. FCV-95 is powered from bus 1-2.

D. Correct. FCV-95 is closed and opens upon actuation. A loss of bus 1-2 will prevent it from opening and the pump will not start.

Technical

References:

LD-1 References to be provided to applicants during exam: None Learning Objective: 8405 - State the power supplies to AFW system components Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content: 55.41.4 4. Secondary coolant and auxiliary systems that affect the facility DCPP L091C Exam Rev 0

Examination Outline Cross-Reference Level RO Tier # 2 061 A3.01 Ability to monitor automatic operation of the AFW, Group # 1 including: AFW startups and flows K/A # 061 A3.01 Rating 4.2 Question 19 Unit 1 is at 18%. Motor Driven AFW pump 1-2 is out of service.

The operator trips the reactor due to loss of 12 kV bus D.

When the operator checks AFW status, steam generator narrow range levels have lowered to 58%.

Without operator action, what is the expected status of the AFW pumps and AFW flow indication?

A. None of the AFW pumps started; no AFW indicated to all four steam generators.

B. Only the Motor Driven AFW pump 1-3 started; AFW indicated to only the 1-1 and 1-2 Steam Generators.

C. The Turbine Drive AFW pump and Motor Driven AFW pump 1-3 started; max AFW indicated to all four steam generators.

D. Only the Turbine Driven AFW pumps started; max AFW indicated to all four Steam Generators.

Proposed Answer: A. None of the AFW pumps started; no AFW indicated to all four steam generators.

Explanation:

A. Correct. The motor driven pumps start for SI, loss of both MFPs, Transfer to diesel, less than 15% on 1 of steam generators or AMSAC. The Turbine AFW pump starts for AMSAC, less than 15% in 2 of 4 steam generators or loss of BOTH 12 kV buses.

B. Incorrect. None of the AFW pumps will be running.

C. Incorrect. None of the AFW pumps start.

D. Incorrect. This would be correct if the other 12 kV bus de-energized.

Technical

References:

LD-1, OIM D-1-2 References to be provided to applicants during exam: None Learning Objective: 37637 - Analyze automatic features and interlocks associated with the AFW system Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.41.4 Secondary coolant and auxiliary systems that affect the facility.

Editorial change to question.

DCPP L091C Exam Rev 1

Examination Outline Cross-Reference Level RO Tier # 2 062 A2.12 Ability to (a) predict the impacts of the following Group # 1 malfunctions or operations on the ac distribution system; and (b) K/A # 062 A2.12 based on those predictions, use procedures to correct, control, or Rating 3.2 mitigate the consequences of those malfunctions or operations:

Restoration of power to a system with a fault on it Question 20 Unit 1 is in MODE 3 following a reactor trip, with Startup power supplying all AC buses.

A fault on the 230 kV Morro Bay line causes a loss of Startup power. The crew is now taking action to backfeed from the 500 kV system in accordance with OP J-2:V, Backfeeding the Unit From the 500kV System.

In accordance with OP J-2:V, which of the following actions must be taken by the operator before circuit breakers CB-532 and CB-632 can be closed?

A. Reset the 4 kV and 12 kV auto transfers.

B. Close the Motor Operated Disconnect from CC3.

C. Reset the Unit Lockout Relays, 86G1 and 86G11 on VB4.

D. Switch all device CUTOUT switches on VB4 with Blue lamacoids from CUTOUT to CUTIN.

Proposed Answer: C. Reset the Unit Lockout Relays, 86G1 and 86G11 on VB4.

Explanation:

A. Incorrect. the auto transfer relays do not input into the operation of the 500 kV breakers.

B. Incorrect. The MOD must be OPEN, not closed.

C. Correct. Tripping the 86G1 and G11 relays trips 532 and 632. they must be reset before the breakers can be closed.

D. Incorrect. The CUTOUT switches are taken to CUTOUT. The normal (pretrip) position of these switches is CUTIN.

Technical

References:

OP J-2:V, LJ-4A References to be provided to applicants during exam: None Learning Objective: 5280 - Analyze automatic features and interlocks associated with the Main Generator.

Question Source: Bank #

(note changes; attach parent) Modified Bank # A-0915 X New Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content: 55.41.10 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Replaced D.

DCPP L091C Exam Rev 1

Examination Outline Cross-Reference Level RO Tier # 2 063 K1.03 Knowledge of the physical connections and/or cause- Group # 1 effect relationships between the DC electrical system and the K/A # 063 K1.03 following systems: Battery charger and battery Rating 2.9 Question 21 A battery equalizing charge has just been started.

Which of the following are indications the equalizing charge is in progress?

A. A drop in battery cell level and temperature and a rise in battery charger voltage.

B. A rise in battery cell level and temperature and a drop in battery charger voltage.

C. A drop in battery cell level and temperature and a drop in battery charger voltage.

D. A rise in battery cell level and temperature and a rise in battery charger voltage.

Proposed Answer: D. A rise in battery cell level and temperature and a rise in battery charger voltage.

Explanation:

A. Incorrect. The equalizing charge causes gas formation and cell temperature (and level) to rise.

B. Incorrect. Battery charger voltage rises for the equalization charge. Its not an equalization of battery and charger voltages.

C. Incorrect. Both rise.

D. Correct. Both rise.

Technical

References:

OP J-9:IV References to be provided to applicants during exam: None Learning Objective: 7115 - Describe the operation of the DC Power System Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content: 55.41.5 Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.

DCPP L091C Exam Rev 0

Examination Outline Cross-Reference Level RO Tier # 2 064 G2.1.30 EDG: Ability to locate and operate components, Group # 1 including local controls. K/A # 064 G2.1.30 Rating 4.4 Question 22 An operator has been dispatched to start an emergency diesel generator locally.

The operator takes the START/STOP switch on the LOCAL panel to START.

The diesel will ONLY start if the Control Selector switch (LOCAL/REMOTE) on the Excitation Cubicle is in LOCAL AND:

A. the Mode Control switch (TEST/AUTO) on the Local panel is in AUTO.

B. the Mode Control switch (TEST/AUTO) on the Local panel is in TEST.

C. the Droop Switch (DROOP/ISOC) on the Excitation Cubicle is in ISOC.

D. the Droop Switch (DROOP/ISOC) on the Excitation Cubicle is in DROOP.

Proposed Answer: B. the Mode Control switch (TEST/AUTO) on the Local panel is in TEST.

Explanation:

A. Incorrect. In AUTO, the start/stop circuit is not aligned. The Mode Select switch must be in TEST.

B. Correct. Placing the Control Selector in LOCAL and the Mode Control switch in TEST aligns the Start/Stop switch and the diesel would start.

C. Incorrect. the Droop switch in aligned in Local and Test however, it is not part of the start circuit.

D. Incorrect. the Droop switch in aligned in Local and Test however, it is not part of the start circuit.

Technical

References:

LJ-6B References to be provided to applicants during exam: None Learning Objective: 6431 - State the purpose of Diesel Generator System components Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content: 55.41.8 Components, capacity, and functions of emergency systems.

Changed A. As a result of change to A, moved redundant wording, (Control Selector switch (LOCAL/REMOTE) on the Excitation Cubicle is in LOCAL AND: ) to question.

DCPP L091C Exam Rev 1

Examination Outline Cross-Reference Level RO Tier # 2 064 K6.08 Knowledge of the effect of a loss or malfunction of the Group # 1 following will have on the ED/G system: Fuel oil storage tanks K/A # 064 K6.08 Rating 3.2 Question 23 If there is less than the required level in the diesel fuel oil storage tanks, the emergency diesel generators may not operate for the required Engineered Safeguards MINIMUM assumed time of:

A. 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> B. 7 days C. 14 days D. 30 days Proposed Answer: B. 7 days.

Explanation:

A. Incorrect. 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is a frequently used tech spec action time.

B. Correct. Calculated fuel oil consumption necessary to support the operation of the DGs to power the minimum engineered safety feature (ESF) systems required to mitigate a design basis accident (LOCA) in one unit and those minimum required systems for a concurrent non-LOCA safe shutdown in the remaining unit (both units initially in MODE 1 operation).

The fuel oil consumption is calculated for a period of 7 days operation of minimum ESF systems.

C. Incorrect. 14 days is a frequent tech spec action time.

D. Incorrect. 30 days is a frequent tech spec action time.

Technical

References:

LJ-6B References to be provided to applicants during exam: None Learning Objective: 41342 - Explain significant Diesel Generator System design features and the importance to nuclear safety.

Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content: 55.41.8 Components, capacity, and functions of emergency systems Editorial change to question DCPP L091C Exam Rev 1

Examination Outline Cross-Reference Level RO Tier # 2 073K3.01 Knowledge of the effect that a loss or malfunction of Group # 1 the PRM system will have on the following: Radioactive effluent K/A # 073 K3.01 releases Rating 3.6 Question 24 RE-23, Steam Generator Blowdown (SGBD) Effluent Radiation Monitor, has lost power.

What combination, if any, of the following SGBD valves will close?

1. Outside Containment Sample Isolation Valves

2. Outside Containment Blowdown Isolation Valves

3. Inside Containment Blowdown Isolation Valves A. None; loss of power to the radiation monitor will not cause the valves to reposition.

B. 2 and 3 C. 1 and 3 D. 1 and 2 Proposed Answer: D. 1 and 2 Explanation:

A. Incorrect. Loss of power will cause the rad monitor to fail low, however, auto actions occur as a result of the loss of power.

B. Incorrect. The inside containment valves, unlike other isolation signals, (ie Phase A), are not affected, only the outside isolation valves are affected.

C. Incorrect. The inside containment valves, unlike other isolation signals, (ie Phase A), are not affected, only the outside isolation valves are affected.

D. Correct. The outside sample and blowdown isolation valves go closed.

Technical

References:

LD-2 References to be provided to applicants during exam: None Learning Objective: 35705 - Describe controls, indications, and alarms associated with the SGBD System Question Source: Bank # S-41588 X (note changes; attach parent) Modified Bank #

New Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content: 55.41.7 Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

DCPP L091C Exam Rev 0

Examination Outline Cross-Reference Level RO Tier # 2 076 K3.07 Knowledge of the effect that a loss or malfunction of Group # 1 the SWS will have on the following: ESF Loads K/A # 076 K3.07 Rating 3.7 Question 25 A total loss of ASW has just occurred on Unit 2.

In accordance with OP AP-11, Malfunction of Component Cooling Water System, what backup cooling can be aligned to the ECCS CCPs?

A. Fire water, but only to one ECCS CCP at a time.

B. Fire water to both ECCS CCPs simultaneously.

C. Temporary portable industrial pumps, but only to one ECCS CCP at a time.

D. Temporary portable industrial pumps to both ECCS CCPs simultaneously.

Proposed Answer: A. Fire water, but only to one ECCS CCP at a time.

Explanation:

A. Correct. Per Appendix C of OP AP-11, the Backup Cooling System can be connected to either ECCS centrifugal charging pump, however, this system can only be used to supply cooling for one charging pump at a time.

B. Incorrect. Only one CCP at a time may be connected.

C. Incorrect. The temporary pumps are aligned in Appendix D to as a backup to the ASW system (aligned thru the ASW heat exchangers), it is not connected to the CCPs.

D. Incorrect The temporary pumps are aligned in Appendix D to as a backup to the ASW system (aligned thru the ASW heat exchangers), it is not connected to the CCPs.

Technical

References:

OP AP-11 appendix C and D References to be provided to applicants during exam: None Learning Objective: 5664 - State the alternate cooling lineup for the CCPs using fire water and explain how it affects CCP cooling operations.

Question Source: Bank #

(note changes; attach parent) Modified Bank # P-36552 X New Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content: 55.41.10 Administrative, normal, abnormal, and emergency operating procedures for the facility Added IAW OP AP-11. removed "can be connected" from distractors and answer.

DCPP L091C Exam Rev 1

Examination Outline Cross-Reference Level RO Tier # 2 078 K1.05 Knowledge of the physical connections and/or cause- Group # 1 effect relationships between the IAS and the following systems: K/A # 078 K1.05 MSIV air Rating 3.4 Question 26 A main steam isolation valve (MSIV) is open and will not close from the Control Room.

To close it, an operator will close the valve locally by isolating the instrument air supply and:

A. venting the accumulators.

B. isolating the backup nitrogen bottles.

C. closing the MSIV with the handwheel.

D. opening the breakers for the DC solenoids.

Proposed Answer: A. venting the accumulators.

Explanation:

A. Correct. Both air supplies are isolated/vented to fail the MSIVs closed.

B. Incorrect. Unlike the 10% steam dumps, the backup is air tanks, not nitrogen.

C. Incorrect. These valves do not have handwheels (unlike other secondary valves, such as the Feedwater isolation valves).

D. Incorrect. The DC solenoids energize to close the MSIVs (normally de-energized).

Technical

References:

E-2, LC-2A, LPE2 References to be provided to applicants during exam: None Learning Objective: 7254 Describe the operation of the Main Steam System.

Question Source: Bank # S-27006 X (note changes; attach parent) Modified Bank #

New Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content: 55.41.7 Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features Moved redundant wording into question and reordered answers.

DCPP L091C Exam Rev 1

Examination Outline Cross-Reference Level RO Tier # 2 078 A3.01 Ability to monitor automatic operation of the IAS, Group # 1 including: Air pressure K/A # 078 A3.01 Rating 3.1 Question 27 All air compressors are available and are aligned in a normal system alignment.

Instrument air header pressure lowers from 106 to 99 psig.

What would be the status of the instrument air compressors?

A. All compressors are running and loaded.

B. Only Rotary compressors 0-5, 0-6 and 0-7 are running and loaded.

C. Only Reciprocating compressors 0-1 through 0-4 are running and loaded.

D. All compressors running but only Rotary compressors 0-5 and 0-6 are loaded.

Proposed Answer: B. Only Rotary compressors 0-5, 0-6 and 0-7 are running and loaded.

Explanation:

A. Incorrect. For a normal system configuration, the reciprocating air compressors will start when pressure is 93 psig and the rotary air compressors start at 103 psig.

B. Correct. The rotary air compressors (0-5, 6 and 7) start at 103 psig and the reciprocating compressors will start at 93 psig.

C. Incorrect. The rotary air compressors start first, followed by the reciprocating.

D. Incorrect. At this pressure, all compressors will be loaded.

Technical

References:

OIM K-1-2 References to be provided to applicants during exam: None Learning Objective: 37565 Analyze automatic features and interlocks associated with the Compressed Air System.

Question Source: Bank #

(note changes; attach parent) Modified Bank # S-35036 X New Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.41.7 Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Raised pressure to 99 psig and changed answer to B. Changed to Modified Bank. Added 0-7 to B as it is a rotary AC and would be running loaded.

DCPP L091C Exam Rev 1

Examination Outline Cross-Reference Level RO Tier # 2 103 A1.01 Ability to predict and/or monitor changes in Group # 1 parameters (to prevent exceeding design limits) associated with K/A # 103 A1.01 operating the containment system controls including: Rating 3.7 Containment pressure, temperature, and humidity Question 28 A large LOCA has occurred.

Containment pressure is rising rapidly.

Which of the following describes the MINIMUM equipment that is assumed to be in operation in order to maintain Containment pressure below design?

A. 3 Containment fan coolers OR both trains of Containment spray.

B. 3 Containment fan coolers AND both trains of Containment spray.

C. 2 Containment fan coolers OR one train of Containment spray.

D. 2 Containment fan coolers AND one train of Containment spray.

Proposed Answer: D. 2 Containment fan coolers AND one train of Containment spray.

Explanation:

A incorrect. Both spray and fans are assumed.

B incorrect. Only 2 fans are assumed.

C incorrect. Both required.

D Correct. Two CFCUs and one train of spray is the minimum.

Technical

References:

LMCD FRZ References to be provided to applicants during exam: None Learning Objective: 40808 Explain significant CFCU design features and the importance to nuclear safety Question Source: Bank # P-29834 X (note changes; attach parent) Modified Bank #

New Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content: 55.41.5 Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.

Caps for MINIMUM DCPP L091C Exam Rev 0

Examination Outline Cross-Reference Level RO Tier # 2 001 A4.02 Ability to manually operate and/or monitor in the Group # 2 control room: boration/dilution K/A # 001 A4.02 Rating 4.1 Question 29 GIVEN:

• When performing a reactor startup following a refueling, the reactor goes critical below the Rod Insertion Limit

• The crew terminates the startup, fully inserts the Control Rods and initiates Emergency Boration in accordance with OP AP-6, Emergency Boration, to raise RCS boron concentration 100 ppm using the normal boration flowpath In accordance with OP AP-6, which of the following is the MINIMUM flowrate and MINIMUM amount of 4% boric acid the crew must add?

A. 30 gpm, 100 gallons B. 30 gpm, 900 gallons C. 90 gpm, 900 gallons D. 90 gpm, 2700 gallons Proposed Answer: B. 30 gpm, 900 gallons Explanation:

A. Incorrect. 30 gpm is correct, but the amount is 900 gallons for 100 ppm.

B. Correct. OP AP-6 states the minimum flowrate is 30 gpm and 900 gallons increases RCS boron concentration 100 ppm at BOL (current plant condition)

C. Incorrect. 90 gpm is the flowrate if using the RWST.

D. Incorrect. 90 gpm is the flowrate if using the RWST, 2700 gallons would be the amount.

Technical

References:

OP AP-6, L-3 References to be provided to applicants during exam: None Learning Objective: 3477F Given an abnormal condition, summarize the major actions of OP AP-6 to mitigate an event in progress.

Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.41.10 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Add MINIMUM and IAW OP AP-6 DCPP L091C Exam Rev 1

Examination Outline Cross-Reference Level RO Tier # 2 002 K1.05 Knowledge of the physical connections and/or cause- Group # 2 effect relationships between the RCS and the following systems: K/A # 002 K1.05 PRT Rating 3.2 Question 30 Unit 2 is at full power.

Which of the following would cause RCS inventory to be directed to the PRT?

A. Phase A actuation B. Phase B actuation C. RCS pressure rises to 2300 psig D. The open Letdown orifice isolation valve closes Proposed Answer: A. Phase A actuation Explanation:

A. Correct. Phase A will cause RCP seal return to close and the letdown containment isolation valve to close, both will cause a relief to lift and direct RCS to the PRT.

B. Incorrect. Phase B will isolate CCW but will not cause any flow to the PRT C. Incorrect. PORV lift setpoint is 2335 psig.

D. Incorrect. Closing the letdown orifice valve will stop RCS flow to letdown, but because it is upstream of the containment isolation, no flow to the PRT will occur.

Technical

References:

LB-1A pages 54 and 70 References to be provided to applicants during exam: None Learning Objective: 40449 Discuss abnormal conditions associated with the CVCS Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.41.7 Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

DCPP L091C Exam Rev 0

Examination Outline Cross-Reference Level RO Tier # 2 011 K6.04 Knowledge of the effect of a loss or malfunction on the Group # 2 following will have on the PZR LCS: Operation of PZR level K/A # 011 K6.04 controllers Rating 3.1 Question 31 The controlling Pressurizer level channel fails at its current value.

The crew begins a downpower from 100% power.

As power is reduced, charging flow will:

A. rise. Actual pressurizer level will rise but then charging flow will lower to maintain level at the failed reference level.

B. rise. Actual pressurizer level will rise and eventually the reactor will trip on high pressurizer level.

C. lower. Actual pressurizer level will lower but then charging flow will rise to maintain level at the failed reference level.

D. lower. Actual pressurizer level will lower and eventually letdown will isolate.

Proposed Answer: D. lower. Actual pressurizer level will lower and eventually letdown will isolate.

Explanation:

A. Incorrect. Charging flow lowers.

B. Incorrect. Charging flow lowers.

C. Incorrect. Actual level will lower, but charging flow is referencing reference level (which appears high) to actual level.

D. Correct. As power lowers, Tave lowers. Operating properly, charging will adjust to maintain level on program ramp. If the reference does not change, then as temperature lowers, charging will lower to attempt to bring the reference level down. Eventually letdown will isolate.

Technical

References:

OIM A-4-3 References to be provided to applicants during exam: None Learning Objective: 36926 Discuss abnormal conditions associated with the Pzr, Pzr Pressure and Level Control System Question Source: Bank # A-1067 X (note changes; attach parent) Modified Bank #

New Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.41.5 Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure DCPP L091C Exam Rev 0

and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.

DCPP L091C Exam Rev 0

Examination Outline Cross-Reference Level RO Tier # 2 014 K4.03 Knowledge of design feature(s) and/or interlock(s) Group # 2 which provide for the following: Maintenance of spent fuel level K/A # 033 K4.01 Rating 2.9 Question 32 A leak develops in the Spent Fuel Pool pump discharge line to the Spent Fuel Pool.

Which of the following describes approximately how low the Spent Fuel Pool level could drop?

A. Until level is below the anti-siphon hole in the discharge line, approximately two feet less than normal water level.

B. Until level is below the anti-siphon hole in the discharge line, approximately twenty three feet above the bottom of the Spent Fuel Pool.

C. Until level is below the discharge piping, approximately eight feet above the spent fuel assemblies.

D. Until level is below the discharge piping, approximately twenty three feet above the top of the spent fuel assemblies.

Proposed Answer: A. Until level is below the anti-siphon hole in the discharge line, approximately two feet less than normal water level.

Explanation:

A. Correct. The anti-sipon hole is at elevation 136 feet 3 inches, normal level is approximately 138 feet, therefore, level would drop approximately two feet.

B. Incorrect. The requirement is to maintain 23 feet above the fuel, (approximately 134 foot elevation), not above the bottom of the pool.

C. Incorrect. This is the point at which the discharge line is at in the SFP.

D. Incorrect. This is close to where the anti-siphon hole is (approximately 135 foot elevation).

Technical

References:

LB-7, pages 8 and 26 References to be provided to applicants during exam: None Learning Objective: 5275 - Identify the location of components associated with the Spent Fuel Pool Cooling System.

Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content: 55.41.2 General design features of the core, including core structure, fuel elements, control rods, core instrumentation, and coolant flow.

Replaced original KA, new question.

DCPP L091C Exam Rev 1

Examination Outline Cross-Reference Level RO Tier # 2 016 A3.01 Ability to monitor automatic operation of the NNIS, Group # 2 including: Automatic selection of NNIS inputs to control systems K/A # 016 A3.01 Rating 2.9 Question 33 What signal from Steam Generator 11 pressure channels (PT-514, 515, 516) will be used as the signal for use in Digital Feedwater Control to develop the Steam/Feed DP for Main Feedwater Pump speed control?

A. The median of the three channels B. The auctioneered high of the three channels C. The auctioneered low of the three channels D. The average of the three channels Proposed Answer: A. The median of the three channels.

Explanation:

A. Correct. DFWCS uses the median from each steam generator as input to the Steam/Feed DP signal.

B. Incorrect. Unlike other circuits, such as Pressurizer level, DFWCS does not use auctioneering to select inputs.

C. Incorrect. Its plausible that the low would be used as a conservative input.

D. Incorrect. The average is not used, its the median (middle, which usually will not be the average).

Technical

References:

OIM C-8-4a, d, LC-8B References to be provided to applicants during exam: None Learning Objective: 4962 Describe DFWCS components Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content: 55.41.7 Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Replaced D DCPP L091C Exam Rev 1

Examination Outline Cross-Reference Level RO Tier # 2 012 K2.01 - Knowledge of bus power supplies to the following: Group # 1 NIS channels, components, and interconnections K/A # 015 K2.01 Rating 3.3 Question 34 The crew has taken the reactor critical and is now taking Critical Data at 10-8 amps.

A loss of 120 VAC Vital Instrument Panel, PY-12, occurs.

The reactor automatically trips due to a loss of power to which of the following components?

A. Source Range N32 B. Intermediate Range N36 C. The UV coil for Reactor Trip Breaker A D. The UV coil for Reactor Trip Breaker B Proposed Answer: B. Intermediate Range N36 Explanation:

A. Incorrect. Critical Rod Height data is taken in the IR, the source ranges are de-energized.

B. Correct. Loss of a single Intermediate Range channel causes an IR high flux trip.

C. Incorrect. DC powered from SSPS.

D. Incorrect. DC powered from SSPS.

Technical

References:

OIM B-6-1b, OP L-2, LPA-4 References to be provided to applicants during exam: None Learning Objective: 4274 Explain the consequences of loss of vital instrument bus Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.41.7 Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features Replaced KA 028 K2.01. Moved from question 09 (KA is 012 K2.01 - new question for this KA).

DCPP L091C Exam Rev 1

Examination Outline Cross-Reference Level RO Tier # 2 029 A1.02 Ability to predict and/or monitor changes in Group # 2 parameters to prevent exceeding design limits) associated with K/A # 029 A1.02 operating the Containment Purge System controls including: Rating 3.4 Radiation levels Question 35 Unit 1 is at full power.

The crew is planning to perform a routine discharge of the Containment atmosphere in accordance with OP H-4:I, Containment Ventilation Make Available and Place in Service.

In accordance with OP H-4:I, which of the following must be OPERABLE to monitor the purge and alert the operator to rising radiation levels?

A. Containment Air Particulate monitor, RM-11 B. Containment Rad Gas monitor, RM-12 C. Plant Vent monitors, RM-14 and 14R D. Containment Vent monitors, RM-44A and 44B Proposed Answer: D. Containment Vent monitors, RM-44A and 44B Explanation:

A. Incorrect. RM-11 monitors air particulate in Containment but is not required to be operable for a purge.

B. Incorrect. RM-12 monitors gas particulate in Containment but is not required to be operable for a purge.

C. Incorrect. The release go thru the plant vent but its RM-44A and B that must be OPERABLE to perform a purge.

D. Correct. Both monitors must be OPERABLE to be able to isolate containment ventilation.

Technical

References:

OP H-4:I References to be provided to applicants during exam: None Learning Objective: 5141 Describe the operation of the Containment Purge System Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content: 55.41.11Purpose and operation of radiation monitoring systems, including alarms and survey equipment.

Added procedure name and number to question.

DCPP L091C Exam Rev 1

Examination Outline Cross-Reference Level RO Tier # 2 045 G2.4.49 Main Turbine Generator: Ability to perform without Group # 2 reference to procedures those actions that require immediate K/A # 045 G2.4.49 operation of system components and controls. Rating 4.6 Question 36 One of the two running Main Feedwater pumps trip and the crew enters OP AP-15, Loss of Feedwater Flow, Section A, One MFP Trips With Both MFPs Operating.

In accordance with OP AP-15, which of the following conditions would require the operator to perform an immediate reactor trip?

A. Reactor power at 90%

B. Turbine load of 1000 MW with no program ramp in progress C. Running MFW pump suction pressure of 225 psig and stable D. Control Rods in MANUAL Proposed Answer: B. Turbine load of 1000 MW with no program ramp in progress Explanation:

The first 4 steps (corresponding to A thru D) of AP-15 are immediate actions.

A. Incorrect. Previously required a reactor trip, now the action is to start the motor driven AFW pumps.

B. Correct. Above 650 MW and no ramp in progress requires a reactor trip (step 2 RNO)

C. Incorrect. Step 3 RNO is to trip the reactor if suction pressure is less than 190 psig.

D. Incorrect. This is the RNO for step 4 if rods not controlling auto.

Technical

References:

OP AP-15 References to be provided to applicants during exam: None Learning Objective: 9693 State the steps and transitions in procedures that are considered immediate actions Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content: 55.41.10 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Added IAW OP AP-15 to question for focus. "running" to C and "control" to D DCPP L091C Exam Rev 1

Examination Outline Cross-Reference Level RO Tier # 2 071 K3.05 Knowledge of the effect that a loss or malfunction of Group # 2 the Waste Gas Disposal System will have on the following: ARM K/A # 071 K3.05 and PRM systems Rating 3.2 Question 37 A discharge of a gas decay tank is in progress.

The pressure control valve controlling the flow rate of the gas decay tank discharge malfunctions, causing Waste Gas Vent monitor, RE-22, to detect activity higher than its high alarm setpoint.

Which of the following occurs?

A. RCV-17, GDT to Plant Vent valve, closes ONLY.

B. RCV-17, GDT to Plant Vent valve, AND the GDT Plant Vent Header Isolation valve for the GDT being discharged closes.

C. FCV-410, GDT Outlet to Plant Vent valve, closes ONLY.

D. FCV-410, GDT Outlet to Plant Vent valve, AND the GDT Plant Vent Header Isolation valve for the GDT being discharged closes.

Proposed Answer: A. RCV-17, GDT to Plant Vent valve, closes ONLY.

Explanation:

A. Correct. RE-22 in high alarm closes RCV-17.

B. Incorrect. No auto closure on high radiaton.

C. Incorrect. FCV-410 does not auto close on high radiation.

D. Incorrect. FCV-410 does not auto close on high radiation.

Technical

References:

LF-2 References to be provided to applicants during exam: None Learning Objective: 37706 Describe controls, indications, and alarms associated with the Gaseous Radwaste System.

Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content: 55.41.11 Procedures and equipment available for handling and disposal of radioactive materials and effluents.

Editorial corrections (caps on ONLY, AND)

DCPP L091C Exam Rev 1

Examination Outline Cross-Reference Level RO 075 A2.01 Ability to (a) predict the impacts of the following Tier # 2 malfunctions or operations on the circulating water system; and Group # 2 (b) based on those predictions, use procedures to correct, control, K/A # 075 A2.01 or mitigate the consequences of those malfunctions or operations: Rating 3.0 Loss of intake structure Question 38 GIVEN:

• Unit 1 is at 70% power, ramping down at 5 MW/minute due to screen DP alarms

• The crew has entered OP AP-7, Degraded Condenser, section C, Travelling Screen Problem

• All screens are in HIGH speed in MANUAL

• Circ Water pump 11 screen DP is 80 inches and rising

• Circ Water pump 12 screen DP is 40 inches and rising Which of the following actions should be taken by the crew in accordance with OP AP-7?

A. Trip the reactor and then trip both Circ Water pumps.

B. Trip the Circ Water pump 11 and monitor pump 12 screens for increased loading due to backflow from the 11 pump.

C. Trip Circ Water pump 12 and monitor pump 11 screens for decreased loading due to backflow from the 12 pump.

D. Continue to monitor screen DP and increase the ramp rate to reduce power to less than 50%,

then remove Circ Water pump 11 from service.

Proposed Answer: B. Trip the Circ Water pump 11 and monitor pump 12 screens for increased loading due to backflow from the 11 pump Explanation:

A. Incorrect. The only pump that needs to be secured at this point is the 11 pump.

B. Correct. With DP above 70 inches, the pump is secured. There is the possibility that backflow from the secured pump could cause increased loading on the remaining pump.

C. Incorrect. The backflow will cause increased loading.

D. Incorrect. The 11 pump needs to be removed from service at this time.

Technical

References:

OP AP-7 References to be provided to applicants during exam: None Learning Objective: 3477G Given an abnormal condition, summarize the major actions of the abnormal operating procedure to mitigate an event in progress.

Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.41.10 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Editorial correction "iaw OP AP-7" DCPP L091C Exam Rev 1

Examination Outline Cross-Reference Level RO Tier # 1 EPE 007 EK3.01 Knowledge of the reasons for the following as Group # 1 the apply to a reactor trip: Actions contained in EOP for reactor K/A # EPE007 trip EK3.01 Rating 4.0 Question 39 The crew is performing step 6, CHECK RCS Temperature, in E-0, Reactor Trip or Safety Injection.

RCS temperature is 540°F. The Shift Foreman directs the operator to stabilize RCS temperature at its current value.

What is the basis for stabilizing temperature at 540°F?

A. Lengthens the time to Pressurizer overfill.

B. Mitigates PTS concerns if a repressurization of the RCS occurs.

C. Aids in meeting TCOA time in the event the reason for the Safety Injection is a tube rupture.

D. Prevents the opening of the atmospheric steam dump valves when the main condenser is not available.

Proposed Answer: A. Lengthens the time to Pressurizer overfill.

Explanation:

A. Correct. Controlling RCS heatup rate or stabilizing RCS temperature below 547°F will allow more volume in the Pzr before overfill becomes a problem. This has proven to be an effective strategy in preventing Pzr overfill during inadvertent SI scenarios. This does not contradict the intent of the step which is to check for proper control operation and plant stabilization.

B. Incorrect. Repressurization is a concern with excessive RCS cooldown, but this is not the basis for the step.

C. Incorrect. There are multiple TCOAs for a SGTR, however, this step does not aid in any of them. The TCOAs are concerned with the time to start the cooldown or stop the cooldown and initiate depressurization, not the time to perform the cooldown.

D. Incorrect. If the crew stabilizes at the current temperature, it would be done by adjusting the setpoint of the steam dumps and the atmospherics would lift as necessary to control temperature.

Technical

References:

E-0 step 6 (formally a note prior to step 8)

References to be provided to applicants during exam: None Learning Objective: 7920A Explain basis of emergency procedure steps (E-0, E-0.1) including:

Bases for TCOAs with operator actions of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or less.

Question Source: Bank # R-50632 X (note changes; attach parent) Modified Bank #

New Question History: Last NRC Exam No DCPP L091C Exam Rev 1

Question Cognitive Level: Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content: 55.41.5 Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.

Editorial change, added actual temperature, not "less than 547F" DCPP L091C Exam Rev 1

Examination Outline Cross-Reference Level RO Tier # 1 APE 008 AK2.02 Knowledge of the interrelations between the Group # 1 Pressurizer Vapor Space Accident and the following: Sensors and K/A # APE008 dectectors AK2.02 Rating 2.7 Question 40 The plant is operating at 100% power. PRT pressure is 5 psig.

A Pressurizer PORV begins to leak by. RCS pressure has lowered to 2100 psig and PRT pressure has risen to 15 psig.

Which of the following describes how PORV tailpipe temperature instrument, TI-463, responded to the event?

The temperature indication on TI-463:

A. remained the same as RCS pressure lowered.

B. rose as PRT pressure rose.

C. lowered as RCS pressure lowered.

D. lowered as PRT pressure rose.

Proposed Answer: B. rose as PRT pressure rose.

Explanation:

A. Incorrect. The throttling process results in the tailpipe temperature following PRT pressure.

B. Correct. As long as the conditions remain inside the saturation dome, as PRT pressure rose, Tsat for the PRT rose as well, and the temperature of the tailpipe will reflect the rising tailpipe temperature.

C. Incorrect. Tailpipe temperature will rise.

D. Incorrect. Tailpipe temperature will rise.

Technical

References:

Steam Tables References to be provided to applicants during exam: none Learning Objective:

Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.41.5 Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.

Removed reference (not needed)

DCPP L091C Exam Rev 1

Examination Outline Cross-Reference Level RO Tier # 1 EPE 009 EK1.02 Knowledge of the operational implications of Group # 1 the following concepts as they apply to the small break LOCA: K/A # EPE009 Use of steam tables EK1.02 Rating 3.5 Question 41 GIVEN:

• The crew is performing step 15, CHECK If One ECCS CCP Should Be Stopped, in EOP E-1.2, Post-LOCA Cooldown and Depressurization

• The highest Core Exit Thermocouples indicate 515°F

• RCS Hot Leg Temperatures are 500°F

• RCS wide range pressure is 1115 psig 53°F of subcooling is required to stop one of the ECCS CCPs.

In accordance with E-1.2, is the crew permitted to stop one of the ECCS CCPs?

A. Yes, based on core exit thermocouples, there is sufficient subcooling.

B. Yes, based on hot leg temperatures, there is sufficient subcooling.

C. No, based on core exit thermocouples, there is insufficient subcooling.

D. No, based on hot leg temperatures, there is insufficient subcooling Proposed Answer: C. No, based on core exit thermocouples, there is insufficient subcooling.

Explanation:

A. Incorrect. Tsat for 1130 psia is approximately 560°F. Therefore, subcooling is 45°F, not the required 53°F.

B. Incorrect. If Thot was used, there is sufficient subcooling. Thot is used for temperature indication, however, not to determine subcooling.

C. Correct. Only approximately 45°F of subcooling currently exists.

D. Incorrect. If Thots were used, there is sufficient subcooling.

Technical

References:

steam tables, E-1.2 References to be provided to applicants during exam: steam tables Learning Objective: 65849 Define and describe subcooling margin.

Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.41.14 Principals of heat transfer, thermodynamics and fluid mechanics.

Editorial change to question DCPP L091C Exam Rev 1

Examination Outline Cross-Reference Level RO Tier # 1 APE 015/017 0Reactor Coolant Pump (RCP) Malfunctions: Group # 1 Ability to interpret reference materials, such as graphs, curves, K/A # APE015/017 tables, etc. G2.1.25 Rating 3.9 Question 42 GIVEN:

• The crew is at step 1 of OP AP-28, Reactor Coolant Pump Malfunction, Section B, RCP No.

1 Seal Failure due to a problem with RCP 1-1

• An operator has begun monitoring RCP 1-1 Operating Limits, using attachment 4.1

• RCP 1-1 indications:

o Seal injection is 8 gpm o No. 1 seal return flow on FR-159 on VB2 indicates 0.7 gpm and stable o An operator from inside Containment reports that No. 2 seal leakoff flow indication on FI-169 is 0.5 gpm and stable o Radial Bearing temperature is 185°F and stable o RCP seal outlet temperature is 200°F and stable Which of the following actions will be performed by the crew in accordance with OP AP-28?

A. Trip the reactor and trip RCP 1-1 per the foldout page step 1.0, No. 1 Seal, Trip Criteria 1.

B. Trip the reactor and trip RCP 1-1 per the foldout page step 1.0, No. 1 Seal, Trip Criteria 4.

C. Take the actions, (shutdown the unit per OP AP-25 within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, etc.), per step 1 of Section B for Total No. 1 Seal Leakoff less than 0.8 gpm and stable radial bearing and seal outlet temperature.

D. Take the actions, (Continue pump operation, etc), per step 1 of Section B for Total No. 1 Seal Leakoff of 0.8 gpm to 6 gpm and stable radial bearing and seal outlet temperature.

Proposed Answer: D. Take the actions, (Continue pump operation, etc), per step 1 of Section B for Total No. 1 Seal Leakoff of 0.8 gpm to 6 gpm and stable radial bearing and seal outlet temperature.

Explanation:

A. Incorrect. Seal injection flow, not return flow is greater than 6 gpm.

B. Incorrect. Seal return flow is the sum of both seal flows, 0.7 and 0.5 or 1.2 gpm.

C. Incorrect. The criteria for the first action in the table at step 1 does not apply. Seal return flow is 1.2 gpm (0.7 + 0.5)

D. Correct. Per the table for step 1, continued operation is allowed. All monitored paramters per attachment 4.1 and 4.2 are satisified.

Technical

References:

OP AP-28 References to be provided to applicants during exam: OP AP-28 page 8 and FOP Learning Objective: 7927 - Given initial conditions and assumptions, determine if a reactor trip or safety injection actuation is required Question Source: Bank #

(note changes; attach parent) Modified Bank #

DCPP L091C Exam Rev 0

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.41.10 Administrative, normal, abnormal, and emergency operating procedures for the facility.

DCPP L091C Exam Rev 0

Examination Outline Cross-Reference Level RO Tier # 1 APE 022 AA1.05 Ability to operate and/or monitor the following Group # 1 as they apply to the Loss of Reactor Coolant Makeup: RCP seal K/A # APE022 back pressure regulator valves and flow indicators AA1.05 Rating 2.9 Question 43 Unit 1 is at full power.

HCV-142, RCP Seal Flow Control valve fails closed.

The expected response is that:

A. Charging flow indication on FI-128 and all seal injection flow indications go to zero.

B. Charging flow indication on FI-128 lowers and all seal injection flow indications go high.

C. Charging flow indication on FI-128 and all seal injection flow indications go high.

D. Charging flow indication on FI-128 goes high and all seal injection flow indications go to zero.

Proposed Answer: B. Charging flow indication on FI-128 lowers and all seal injection flow indications go high.

Explanation:

A. Incorrect. HCV-142 is a fail closed valve, used by the operartor to create backpressure and ensure adequate seal injection flow, controlled by the operator at the Center Console.

Opening the valve increases charging and lowers seal injection flow, closing down on the valve does the opposite. Loss of air will cause it to close, stopping charging but forcing more water into the seal injection lines.

B. Correct. The valve fails closed, stopping charging and causing an increase in seal injection.

C. Incorrect. This would occur if it was FCV-128 that failed closed.

D. Incorrect. This would occur if HCV-142 was in line with the seal injection valves.

Technical

References:

LB-1A, OIM B-1-1, OP AP-17 References to be provided to applicants during exam: None Learning Objective: 40448 Describe controls, indications, and alarms associated with the CVCS.

Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.41.4 Secondary coolant and auxiliary systems that affect the facility.

DCPP L091C Exam Rev 0

Examination Outline Cross-Reference Level RO Tier # 1 APE 025 AK2.05 Knowledge of the interrelations between the Group # 1 Loss of Residual Heat Removal System and the following: K/A # APE025 Reactor building sump AK2.05 Rating 2.6 Question 44 Following a LOCA, indications of Containment sump blockage was noted and EOP ECA-1.3, "Sump Blockage Guideline," was entered.

Per ECA-1.3, what is the basis for throttling RHR Flow Control valves, FCV-637 and FCV-638?

A. Maintaining flow less than 1500 gpm.

B. Maintaining less than 57 amps on the running RHR pump.

C. Prevent/stop cavitation of the running RHR pump.

D. Delay depletion of RWST inventory.

Proposed Answer: C. Prevent/stop cavitation of the running RHR pump.

Explanation:

A. Incorrect. 1500 gpm is a limit in some operating procedures, such as draining to mid-loop (A-2:III)

B. Incorrect. 57 amps is a limit in E-1.3 C. Correct. The valves are throttled if there is indication of cavitation or to prevent cavitation when establishing recirculation flow.

D. Incorrect. This is a goal of ECA-1.1 Technical

References:

ECA-1.3, A-2:III, E-1.3 and ECA-1.1 References to be provided to applicants during exam: None Learning Objective: 42460 Explain basis of emergency steps of ECA-1.1 Question Source: Bank # P-45906 X (note changes; attach parent) Modified Bank #

New Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content: 55.41.8 Components, capacity, and functions of emergency systems.

Editorial change to question DCPP L091C Exam Rev 1

Examination Outline Cross-Reference Level RO Tier # 1 APE 026 AK3.04 Knowledge of the reasons for the following Group # 1 responses as they apply to the Loss of Component Cooling K/A # APE 026 Water: Effect on the CCW flow header of a loss of CCW AK3.04 Rating 3.5 Question 45 CCW heat exchanger fouling is causing CCW temperatures to rise slowly.

The crew is about to perform CCW heat load isolation in accordance with OP AP-11 Malfunction of the Component Cooling Water System, Appendix B, CCW Heat Load Isolation.

In accordance with OP AP-11, which of the following describes whether or not the Containment Fan Cooler Units, (CFCUs), should be isolated and why?

A. The CFCUs should not be isolated to prevent steam binding and water hammer in the units due to CCW flashing.

B. The CFCUs should not be isolated because they may be used as a heat sink for the CCW system.

C. The CFCUs should be isolated because this will substantially reduce CCW heat load.

D. The CFCUs should be isolated to prevent steam binding and water hammer in the units due to CCW flashing.

Proposed Answer: B. The CFCUs should not be isolated because they may be used as a heat sink for the CCW system.

Explanation:

A. Incorrect. Steam binding and water hammer are concerns for a LOCA and transfer to diesel scenario and the reason for the nitrogen pressurization system.

B. Correct. Unless leaking or there is the possibility of exceeding containment temperature tech spec limit, the CFCUs are typically a heat sink for the CCW system.

C. Incorrect. They should not be isolated and while a large load, they will be removing heat from the system.

D. Incorrect. Steam binding and water hammer are concerns for a LOCA and transfer to diesel scenario and the reason for the nitrogen pressurization system.

Technical

References:

OP AP-11 References to be provided to applicants during exam: None Learning Objective: 35490 Discuss abnormal conditions associated with the CCW system Question Source: Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content: 55.41.8 Components, capacity, and functions of emergency systems.

Editorial change to question, changed C DCPP L091C Exam Rev 1

Examination Outline Cross-Reference Level RO Tier # 1 APE 027 AK1.02 Knowledge of the operational implications of Group # 1 the following concepts as they apply to Pressurizer Pressure K/A # APE027 Control Malfunctions: Expansion of liquids as temperature AK1.02 increases Rating 2.8 Question 46 GIVEN:

• The unit is at full power

• Master Pressure Controller PC-455K is in MANUAL

• Backup Pressurizer heaters are OFF in AUTO

• Pressurizer spray valves are CLOSED A 100 MWe load rejection occurs.

How is the initial response of the Pressurizer pressure control system to the load rejection affected by having PC-455K in MANUAL?

A. Pressurizer spray valves will not respond to the pressure increase caused by the load rejection.

B. Pressurizer heaters will not respond to the greater than 5% level deviation from program level caused by the load rejection.

C. Pressurizer heaters will not respond to the pressure decrease caused by the load rejection.

D. Pressurizer spray valves will not respond to the pressure increase caused by the heaters automatically energizing due to the load rejection.

Proposed Answer: A. Pressurizer spray valves will not respond to the pressure increase caused by the load rejection.

Explanation: .

A. Correct. The load rejection will cause the RCS temperature to increase and expand. A pressurizer load rejection results in a new, higher, equilibrium pressure and temperature condition. The pressure increase is due to compressing the steam bubble. The temperature increase is due to some of the steam condensing to water, transferring latent heat to the steam/water interface. The magnitude of the pressure increase is minimized by the condensing action of the steam. However, the rise in pressure will not be mitigated by the spray valves if the master pressure controller is in manual. The PORVs would respond.

B. Incorrect. RCS will heatup and raise level, but the heaters will energize because they are turned on by the master level controller in this case.

C. Incorrect. Pressure will begin to rise due to an load rejection.

D. Incorrect. There will be an load rejection, not an outsurge.

Technical

References:

TH14T, OIM References to be provided to applicants during exam: None Learning Objective: 36926 Discuss abnormal conditions associated with the Pzr, Pzr Pressure and Level Control System Question Source: Bank #

DCPP L091C Exam Rev 1

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.41.14 Principals of heat transfer, thermodynamics and fluid mechanics.

Replace insurge/outsurge with load rejection. Add "automatically" energize to D and "deviation from" to B. Added Auto to heater setup to eliminate B as a possible answer.

DCPP L091C Exam Rev 1

Examination Outline Cross-Reference Level RO Tier # 1 APE 054 AA2.04 Ability to determine and interpret the following Group # 1 as they apply to the Loss of Main Feedwater (MFW): Proper K/A # APE054 operation of AFW pumps and regulating valves AA2.04 Rating 4.2 Question 47 GIVEN:

• Unit 2 trips from 20% due to a loss of the running main Feedwater pump

• Steam Generator narrow range levels drop to 58% and are returning to setpoint While in E-0.1, Reactor Trip Response, the operator reports that only the motor driven AFW pumps are running and that total AFW flow is less than 200 gpm because the level control valves, (LCVs) are not fully open.

What action, if any, should the operator recommend to address the AFW status?

A. Fully closing the LCVs because narrow range level is above 15% in all steam generators.

B. Opening the motor driven AFW pump LCVs to establish AFW flow of greater than 435 gpm.

C. Transitioning to FR-H.1, Loss of Secondary Heat Sink.

D. Nothing, the AFW system is responding as designed.

Proposed Answer: D. Nothing, the AFW system is responding as designed.

Explanation:

A. Incorrect. While level above 15% defines an adequate heat sink, the system is responding as designed and no action should be taken.

B. Incorrect. Establishing 435 gpm is only required when steam generator levels are low.

C. Incorrect. There is not a loss of heat sink.

D. Correct. A trip from low level results in little shrink. As level returns to program level, the LCVs will close down to control level. This could result in less than 435 gpm of AFW flow. The motor driven AFW pumps start on a loss of both MFPs.

Technical

References:

E-0.1, AP-15, LD-1 References to be provided to applicants during exam: None Learning Objective: 8402 Describe the operation of the Auxiliary Feed Water System.

Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.41.7 Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Replaced A.

DCPP L091C Exam Rev 1

Examination Outline Cross-Reference Level RO Tier # 1 APE 056 AA2.07 Ability to determine and interpret the following Group # 1 as they apply to the Loss of Offsite Power: Operational status of K/A # APE 056 emergency feedwater pump (motor driven) AA2.07 Rating 4.2 Question 48 Unit 1 is at full power.

A loss of all offsite power occurs. Only Emergency Diesel Generators (EDG) 1-1 and 1-2 are supplying power to their vital 4 kV bus.

Which motor driven AFW pump should be running?

A. AFW pump 1-2 powered from EDG 1-2.

B. AFW pump 1-2 powered from EDG 1-1.

C. AFW pump 1-3 powered from EDG 1-2.

D. AFW pump 1-3 powered from EDG 1-1.

Proposed Answer: B. AFW pump 1-2 powered from EDG 1-1.

Explanation:

A. Incorrect. EDG 1-2 supplies bus G which does not have a motor driven pump as a load.

B. Correct. AFW pump 1-2 is powered off bus H (EDG 11)

C. Incorrect. AFW pump 13 is powered off bus F (EDG 13).

D. Incorrect. AFW pump 13 is powered from bus F (EDG 13).

Technical

References:

OIM J-1-1 References to be provided to applicants during exam: None Learning Objective: 8405 State the power supplies to AFW system components Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.41.8 Components, capacity, and functions of emergency systems.

Changed setup and answer DCPP L091C Exam Rev 1

Examination Outline Cross-Reference Level RO Tier # 1 APE 057 G2.1.30 Loss of Vital AC Electrical Instrument Bus: Group # 1 Ability to locate and operate components, including local K/A # APE 057 controls. G2.1.30 Rating 4.4 Question 49 Unit 1 is at full power.

PK19-19, Vital UPS Failure alarms due to inputs:

o 1484 Instr AC UPS 1-1 Inverter Failure o 1485 Instr AC UPS 1-1 Loss of AC Output What is the expected status of 120 VAC Vital Instrument Panel, PY1-1?

A. De-energized and cannot be re-energized until either either the AC or DC source is restored.

B. De-energized but can be re-energized using the Manual Bypass switch.

C. Energized from its bypass source via the static switch.

D. Energized from its DC source.

Proposed Answer: B. De-energized but can be re-energized using the Manual Bypass switch.

Explanation:

A. Incorrect. The bus can be re-energized by transferring to Manual Bypass.

B. Correct. The PY will be de-energized but can be put on its bypass source using the manual bypass C. Incorrect. The Static Switch is upstream of the AC Output and should have transferred automatically. If it had, then input 1486. UPS on Bypass would be in as well.

D. Correct. The Manual Bypass is would restore the power to the instrument bus and remains aligned until switched back by the operator.

Technical

References:

LJ-10, AR PK19-19 References to be provided to applicants during exam: None Learning Objective: 37807 - Describe controls, indications, and alarms associated with the Instrument AC System.

Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.41.7 Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features Reworded question to avoid having multiple potential answers.

DCPP L091C Exam Rev 1

Examination Outline Cross-Reference Level RO Tier # 1 APE 058 AK3.01 Knowledge of the reasons for the following Group # 1 responses as they apply to the Loss of DC Power: Use of dc K/A # APE058 control power by D/Gs AK3.01 Rating 3.4 Question 50 The operator is transferring DC control power, for an Emergency Diesel Generator, to backup per Attachment 6.3, Turbine Building Actions, of OP AP-8A, Control Room Inaccessibility -

Establishing Hot Standby.

Per OP AP-8A, why is the operator instructed to wait for 10 seconds with the 125 VDC Control Power Transfer Switch in NEUTRAL prior to placing it in BACKUP?

A. The Tach-Pak Diesel circuitry may lockup if deenergized and then reenergized in a short period of time.

B. The diesel may start when control power is restored, waiting will prevent overspeed by allowing time for the fuel racks to return to the no-fuel position.

C. To allow time for the field flash voltage to decay prior to restoring control power.

D. To prevent momentarily cross-tying the normal and backup supplies.

Proposed Answer: A. The Tach-Pak Diesel circuitry may lockup if deenergized and then reenergized in a short period of time.

Explanation:

A. Correct. The Tach Pak per OP AP-8A may experience lockup if power is cycled too rapidly.

B. Incorrect. Control power is switched if the diesel fails to start.

C. Incorrect. This is a problem associated with Appendix R fuses.

D. Incorrect. The switch neutral position will prevent any cross tying issues.

Technical

References:

OP AP-8A, LJ-6B References to be provided to applicants during exam: None Learning Objective: 6408 - Discuss significant precautions and limitations associated with the Diesel Generator System.

Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content: 55.41.7 Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features Editorial change to question DCPP L091C Exam Rev 1

Question 51 Unit 1 is at full power.

A loss of ASW occurs and CCW temperature begins to rise. The crew enters OP AP-11, Malfunction of Component Cooling Water System, section A, Loss of A CCW Pump/High CCW System Temp.

Which of the following is the MINIMUM CCW temperature that would require the crew to manually trip the reactor and stop all the RCPs in accordance with OP AP-11?

A. 96°F B. 112°F C. 128°F D. 144°F Explanation:

A. Incorrect. This is the temperature at which loads are removed from CCWT B. Incorrect. This is less than the limit C. Correct. At 120F, the reactor is tripped and the RCPs are tripped.

D. Incorrect. This is above the limit but also above C, not the minimum temperature.

Technical

References:

OP AP-11 References to be provided to applicants during exam: None Learning Objective: 8105 Explain significant CCW system design features and the importance to nuclear safety Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content: 55.41.10 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Changed answers DCPP L091C Exam Rev 1

Examination Outline Cross-Reference Level RO Tier # 1 EPE 055 AA1.01 Ability to operate and / or monitor the following Group # 1 as they apply to a Station Blackout: In-core thermocouple K/A # APE065 temperatures EA1.01 Rating 3.7 Question 52 Which of the following conditions would require the crew to transition to SACRG-1, Severe Accident Control Room Guideline Initial Response from ECA-0.0, Loss of All Vital AC Power?

A. No AFW flow and all steam generators less than 15% wide range and lowering.

B. Core exit thermocouples greater than 1200°F and increasing.

C. Containment pressure greater than 22 psig and increasing.

D. Spent Fuel Pool level lowering and temperature approaching 212°F.

Proposed Answer: B. Core exit thermocouples greater than 1200°F and increasing Explanation:

A. Incorrect. This is feed and bleed criteria for loss of secondary heat sink.

B. Correct. Core damage is likely occurring and must be addressed.

C. Incorrect. This is a RED path on Containment Integrity with no spray pumps running.

D. Incorrect. This is an indication of a loss of SFP cooling and could result in uncovering the fuel.

Technical

References:

ECA-0.0 step 25 and background References to be provided to applicants during exam: None Learning Objective: 37565 - Analyze automatic features and interlocks associated with the Compressed Air System Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content: 55.41.7 Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Replaced KA and question DCPP L091C Exam Rev 0

Examination Outline Cross-Reference Level RO Tier # 1 E04 EK2.2 Knowledge of the interrelations between the (LOCA Group # 1 Outside Containment) and the following: Facility's heat removal K/A # E04 EK2.2 systems, including primary coolant, emergency coolant, the decay Rating 3.8 heat removal systems, and relations between the proper operation of these systems to the operation of the facility.

Question 53 The crew is performing the actions in ECA-1.2, LOCA Outside Containment.

Per ECA-1.2, which of the following strategies is attempted to isolate the break AND which indication is used to determine if the leak has been isolated?

A. Isolate RHR piping connections; Pressurizer level is monitored, because with the break isolated, RCS inventory will rapidly rise.

B. Isolate RHR piping connections; RCS pressure is monitored, because SI flow will repressurize the RCS with the break isolated.

C. Isolate SI piping connections; Pressurizer level is monitored, because with the break isolated, RCS inventory will rapidly rise.

D. Isolate SI piping connections; RCS pressure is monitored, because SI flow will repressurize the RCS with the break isolated.

Proposed Answer: B. Isolate RHR piping connections; RCS pressure is monitored, because SI flow will repressurize the RCS with the break isolated.

Explanation:

A. Incorrect. Pressurizer level will increase, but after pressure begins to rise and the system is refilled.

B. Correct. If the leak is isolated, pressure will respond and quickly repressurize the system as SI flows into the now intact system.

C. Incorrect. RHR system piping is isolated.

D. Incorrect. RHR system piping is isolated.

Technical

References:

ECA-1.2 and ECA-1.2 background References to be provided to applicants during exam: None Learning Objective: 42461 Explain basis of emergency steps of ECA-1.2 Question Source: Bank # Wolf Creek NRC 2007 X (note changes; attach parent) Modified Bank #

New Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.41.5 Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.

Editorial change to question DCPP L091C Exam Rev 1

Examination Outline Cross-Reference Level RO Tier # 1 E05 G2.4.20 Loss of Secondary Heat Sink: Knowledge of the Group # 1 operational implications of EOP warnings, cautions, and notes. K/A # E05 G2.4.20 Rating 3.8 Question 54 Which of the following describes the basis for this caution in FR-H.1, Response to Loss of Secondary Heat Sink?

CAUTION: Steps 12 through 18 must be performed without delay in order to establish RCS heat removal by RCS bleed and feed.

A. To prevent lifting pressurizer safeties.

B. To prevent or minimize core uncovery.

C. To delay the time until steam generator dryout.

D. To minimize the potential of RCS pressurized thermal shock.

Proposed Answer: B. To prevent or minimize core uncovery Explanation:

A. Incorrect. The reason is to ensure heat removal will be effective.

B. Correct. If quickly established once the effectiveness of the steam generators to remove heat is reduced, bleed and feed will remove RCS decay heat and minimize/prevent core uncovery. If delayed, a much deeper and prolonged core uncovery will occur.

C. Incorrect. RCPs are stopped to delay steam generator dryout. Bleed and feed is established at the point dryout begins.

D. Incorrect. PTS is a concern with many accidents, loss of secondary heat sink is not one of them.

Technical

References:

FR-H.1, FR-H.1 background References to be provided to applicants during exam: None Learning Objective: 7920N Explain basis of emergency procedure steps (FR-Hs)

Question Source: Bank # P-29128 X (note changes; attach Modified Bank #

parent)

New Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content: 55.41.5 Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.

DCPP L091C Exam Rev 0

Examination Outline Cross-Reference Level RO Tier # 1 E11 EA1.3 Ability to operate and / or monitor the following as Group # 1 they apply to the (Loss of Emergency Coolant Recirculation): K/A # E11 EA1.3 Desired operating results during abnormal and emergency Rating 3.7 situations.

Question 55 GIVEN:

• The crew is performing ECA-1.1, Loss of Emergency Coolant Recirculation

• Safety Injection has been reset

• RCS pressure is approximately 250 psig

• Containment spray pumps have been stopped per ECA-1.1 guidance

• Containment pressure is 24 psig

• The crew has established one train of SI flow A loss of offsite power occurs. All emergency diesel generators start and load their respective vital 4 kV buses.

Which of the following describes the expected plant status?

A. No ECCS CCP, SI and RHR pumps running, no Containment Spray pumps running B. Both ECCS CCPs running, no SI or RHR pumps running, all Containment Spray pumps running C. Previously running ECCS charging and SI pumps running, no RHR pumps running, no Containment Spray pumps running D. Both ECCS CCPs running, no SI or RHR pumps running, no Containment Spray pumps running Proposed Answer: D. Both ECCS CCPs running, no SI or RHR pumps running, no Containment Spray pumps running Explanation:

A. Incorrect. Foldout page states the operator must restart ECCS equipment if power is lost after SI is reset. Because SI is reset, none of the SI, RHR or spray pumps will restart, even though pressure is above the spray setpoint. However, the charging pumps, both will start.

B. Incorrect. The spray pumps will remain shutdown.

C. Incorrect. Only the charging pumps restart D. Correct. Charging pumps restart, the SI, RHR and spray pumps remain shutdown.

Technical

References:

ECA-1.1 foldout page, OIM J-6-1 References to be provided to applicants during exam: None Learning Objective: 3552 Given initial conditions, assumptions, and symptoms, determine the correct Emergency Operating Procedure to be used to mitigate an operational event Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No DCPP L091C Exam Rev 0

Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.41.7 Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

DCPP L091C Exam Rev 0

Examination Outline Cross-Reference Level RO E12 EK1.1 Knowledge of the operational implications of the Tier # 1 following concepts as they apply to the (Uncontrolled Group # 1 Depressurization of all Steam Generators): Components, K/A # E12 EK1.1 capacity, and function of emergency systems. Rating 3.4 Question 56 GIVEN:

• The crew is performing ECA-2.1, "Uncontrolled Depressurization of All Steam Generators"

• All steam generator narrow range levels are offscale low

• The RCS has cooled down 125°F in the last 15 minutes Which of the following describes the guidance provided in ECA-2.1 for controlling AFW flow?

A. Isolate all AFW flow to the steam generators to limit the RCS cooldown.

B. Reduce AFW flow to approximately 25 gpm to each steam generator to maintain the steam generator components "wet".

C. Reduce total AFW flow to approximately 435 gpm to limit the RCS cooldown, while maintaining a heat sink.

D. Reduce total AFW flow to approximately 435 gpm to maintain the steam generator components "wet", while maintaining a heat sink.

Proposed Answer: B. Reduce AFW flow to approximately 25 gpm to each steam generator to maintain the steam generator components "wet".

Explanation:

A. Incorrect. A minimum amount of flow is maintained to limit the RCS cooldown while maintaining SG components "wet" to minimize shock if AFW is subsequently restored.

B. Correct. If feed flow to a SG is isolated and the SG is allowed to dry out, subsequent reinitiation of feed flow to the SG could create significant thermal stress conditions on SG components. Maintaining a minimum verifiable feed flow to the SG allows the components to remain in a "wet" condition, thereby minimizing any thermal shock effects if feed flow is increased.

C. Incorrect. 435 gpm is a normal flow rate to maintain a heat sink and limit RCS cooldown.

D. Incorrect. 435 gpm is a normal flow rate, maintaining the components "wet" is the desired result of lowering the AFW flow..

Technical

References:

ECA-2.1, Bases ECA-2.1 References to be provided to applicants during exam: None Learning Objective: 7920J Explain basis of emergency procedure steps (ECA-2.1)

Question Source: Bank # P-29009 X (note changes; attach parent) Modified Bank #

New Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content: 55.41.8 Components, capacity, and functions of emergency systems.

Replaced question (same KA)

DCPP L091C Exam Rev 1

Examination Outline Cross-Reference Level RO Tier # 1 APE 001 AK2.06 Knowledge of the interrelations between the Group # 2 Continuous Rod Withdrawal and the following: Tave./Tref. K/A # APE001 deviation meter AK2.06 Rating 3.0 Question 57 A plant transient causes Tave to be 4°F less than Tref and control rods begin to respond.

Several minutes later, the following conditions exist:

• Turbine load and reactor power are stable

• Tave 1.3°F below Tref

• Control Bank D rods are at 205 steps, in auto and withdrawing at 8 steps per minute Which of the following describes the status of rod control?

A. Rods are responding as designed.

B. Rods should not be moving at this time.

C. Rods should be withdrawing at a faster rate.

D. Rods should be inserting at this time.

Proposed Answer: A. Rods are responding as designed.

Explanation:

A. Correct. Rods begin to move when greater than 1.5°F of Tref but will respond until within 1°F of Tref. At this point, rod speed is in the 8 spm range.

B. Incorrect. If it is believed that rods stop once within 1.5°F, this would be correct or if it is believed rods are above C-11 (223 steps), which would block auto rod withdrawal.

C. Incorrect. there are other possible rod speeds, such as if a greater temperature error exists, then speed could be 8 to 72 spm or if the manual rod speed of 48 spm is used.

D. Incorrect. Rods are moving in the correct direction for the error that exists.

Technical

References:

LA-3A References to be provided to applicants during exam: None Learning Objective: 36987 Describe Rod Control System components Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.41.2 General design features of the core, including core structure, fuel elements, control rods, core instrumentation, and coolant flow.

Added info about position of Bank D rods. It was noted that if above C-11, B would be correct.

DCPP L091C Exam Rev 1

Examination Outline Cross-Reference Level RO Tier # 1 APE 024 AK1.01 Knowledge of the operational implications of Group # 2 the following concepts as they apply to Emergency Boration: K/A # APE024 Relationship between boron addition and change in T-ave AK1.01 Rating 3.4 Question 58 Unit 1 is at 50% power.

Over the last two minutes, Control Bank D has stepped from 180 to 185 steps.

Which of the following could be causing the rod motion?

A. VCT boron concentration lower than RCS boron concentration B. Turbine impulse pressure channel PT-505 slowly failing low C. Leakage through Emergency Borate valve, 8104 D. Turbine impulse pressure channel PT-506 slowly failing high Proposed Answer: C. Leakage through Emergency Borate valve, 8104 Explanation:

A. Incorrect. This would dilute, adding positive reactivity to the RCS and cause rods to insert.

Would be correct if the question is read that rod heights in setup are reversed.

B. Incorrect. PT-505 failing low would cause RCS Tave to be HIGHER than Tref and rods would insert.

C. Correct. Boron adds negative reactivity to the reactor. Tave would lower and rods would step out.

D. Incorrect. PT-506 is not in the circuitry for rod control (PT-505 failing high would cause rods to withdraw).

Technical

References:

TAA18 References to be provided to applicants during exam: None Learning Objective: 10582 DESCRIBE the reactor, RCS and Secondary System responses to each of the following routine evolutions: Changing boron with no changes in turbine load or rod position.

Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.41.1 Fundamentals of reactor theory, including fission process, neutron multiplication, source effects, control rod effects, criticality indications, reactivity coefficients, and poison effects.

Replacement question due to problems with original question. New (not bank) question.

DCPP L091C Exam Rev 1

Examination Outline Cross-Reference Level RO Tier # 1 APE 032 AA1.01 Ability to operate and / or monitor the following Group # 2 as they apply to the Loss of Source Range Nuclear K/A # APE032 Instrumentation: Manual restoration of power AA1.01 Rating 3.1 Question 59 A plant shutdown is in progress.

Intermediate Range channel N36 is undercompensated.

Which of the following describes the effect the N36 undercompensation will have on energizing the Source Range channels?

A. Both Source Range channels will energize when N35 decreases below the P-6 setpoint.

B. N36 will reach the P-6 setpoint early and energize the Source Range channels at a higher level than normal.

C. The Source Range channels will not automatically energize, the operator will have to take the SR Trip RESET/BLOCK switches to BLOCK.

D. The Source Range channels will not automatically energize, the operator will have to take the SR Trip RESET/BLOCK switches to RESET.

Proposed Answer: D. The Source Range channels will not automatically energize, the operator will have to take the SR Trip RESET/BLOCK switches to RESET.

Explanation:

A. Incorrect. N36 will not decrease below the P-6 setpoint, both channels are required to be below P-6 to energize the source ranges.

B. Incorrect. Both channels are required to be below P-6 and in this case, N36 will not lower to less than P-6.

C. Incorrect. The switches are taken to BLOCK to de-energize the source ranges.

D. Correct. Both channels to RESET will energize the source ranges.

Technical

References:

LB-4, OIM References to be provided to applicants during exam: None Learning Objective: 5992 Discuss abnormal conditions associated with the NIS Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.41.2 General design features of the core, including core structure, fuel elements, control rods, core instrumentation, and coolant flow.

Editorial, removed "if any" from question and "no effect" from A DCPP L091C Exam Rev 1

Examination Outline Cross-Reference Level RO Tier # 1 APE 036 AA2.01 Ability to determine and interpret the following Group # 2 as they apply to the Fuel Handling Incidents: ARM system K/A # APE036 indications AA2.01 Rating 3.2 Question 60 RM-58 Fuel Handling Building (FHB) Radiation Control Module indicates the following:

• Red Light is ON for Trip 2

• Amber Light is ON for Trip 1

• Green Operate Light is ON 10 4 Trip 2 Red 10 3 Based on the indications, what automatic actions should have 10 2 Amber Trip 1 occurred? 10 1 10 0 Green Operate 10 -1 A. FHB Evacuation Alarm ONLY TP 4 Slope Adj HV B. Auxiliary Building Ventilation swapped to Iodine removal mode Bias Adj Adj ONLY Meter Trip Ref Trip 2 Set Trip 1 Set C. Both Auxiliary Building and FHB Ventilation swapped to Iodine RM 1 RM 2 Operate Removal Mode TP 3 Check Trip Adj TP 2 TP 1 D. FHB Ventilation swapped to Iodine removal Mode and FHB Evacuation Alarm Proposed Answer: D. FHB Ventilation swapped to Iodine removal Mode and FHB Evacuation Alarm Explanation:

A. Incorrect. This condition would also cause the FHB ventilation system to swap to the Iodine removal mode.

B. Incorrect. Auxiliary Building Ventilation is not swapped to Iodine Removal mode based on RM-58 input signal.

C. Incorrect. Auxiliary Building Ventilation is not swapped to Iodine Removal mode based on RM-58 input signal, only FHB Ventilation is shifted to Iodine Mode.

D. Correct. Red light ON indicates a HI Alarm set point has been exceeded and the FHB evacuation alarm should be sounding and Iodine removal ventilation should be in service.

Technical

References:

AR PK11-10 References to be provided to applicants during exam: None Learning Objective:

Question Source: Bank # DCPP NRC exam 1/2010 X (note changes; attach parent) Modified Bank #

New Question History: Last NRC Exam No DCPP L091C Exam Rev 0

Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.41.11 Purpose and operation of radiation monitoring systems, including alarms and survey equipment.

DCPP L091C Exam Rev 0

Examination Outline Cross-Reference Level RO APE060 AK1.02 Knowledge of the operational implications of the Tier # 1 following concepts as they apply to Accidental Gaseous Radwaste Group # 2 Release: Biological effects on humans of the various types of K/A # APE060 radiation, exposure levels that are acceptable for personnel in a AK1.02 nuclear reactor power plant; the units used for radiation Rating 2.5 intensity measurements and for radiation exposure levels .

Question 61 GIVEN:

• Gas Decay Tank (GDT) 11 has ruptured

• GDT 11 radiation monitor RM-41 is reading 1500 mr/hr

• A Nuclear Operator with a current yearly TEDE exposure of 1000 mr is to go to the GDT to determine the extent of the damage What is the MAXIMUM amount of time the operator could remain in the area and NOT exceed the DCPP Administrative Limit?

A. 20 minutes B. 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and 15 minutes C. 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and 25 minutes D. 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and 50 minutes Proposed Answer: B. 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and 15 minutes Explanation:

A. Incorrect. If the admin guideline of 2000 mr is used, the operator could not stay over 20 minutes.

B. Correct. DCPP admin limit is 4500 mr. The operator could stay a maximum of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and 20 minutes C. Incorrect. If the 4500 mr is used but miscalculates the time D. Incorrect. Neglects the year to date dose.

Technical

References:

RP1.ID6, Personnel Dose Limits and Monitoring Administrative Radiation Exposure Limits, LG-4A References to be provided to applicants during exam: None Learning Objective: STATE the DCPP administrative exposure guidelines Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.41.12 Radiological safety principles and procedures.

Editorial, cap "maximum" and "not" in question DCPP L091C Exam Rev 1

Examination Outline Cross-Reference Level RO Tier # 1 APE 068 G2.4.4 Control Room Evacuation: Ability to recognize Group # 2 abnormal indications for system operating parameters that are K/A # APE068 entry-level conditions for emergency and abnormal operating G2.4.4 procedures. Rating 4.5 Question 62 Other than a fire in the Control Room, a fire in which of the following areas would be an entry condition into OP AP-8A, Control Room Inaccessibility - Establishing Hot Standby?

A. Vital Battery Room B. Cable Spreading Room C. Vital 4 kV bus room D. Main Feed pump area of the Turbine Building Proposed Answer: B. Cable Spreading Room.

Explanation:

A. Incorrect. A fire in the CSR is also a potential entry into OP AP-8A B. Correct. The entry conditions for OP AP-8A lists the CSR as an area that would be an entry condition.

C. Incorrect. The Vital 4 kV rooms would not threaten the operation of the control room.

D. Incorrect. The turbine building is not an area that would require control room evacuation or loss of operation of safety related equipment.

Technical

References:

OP AP-8A References to be provided to applicants during exam: None Learning Objective: 3478 Given initial conditions, assumptions, and symptoms, determine the correct abnormal operating procedure to be used to mitigate an operational event Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content: 55.41.10Administrative, normal, abnormal, and emergency operating procedures for the facility.

Removed "if any" and replaced A DCPP L091C Exam Rev 1

Examination Outline Cross-Reference Level RO Tier # 1 E02 EK3.4 Knowledge of the reasons for the following responses Group # 2 as they apply to the (SI Termination): RO or SRO function K/A # E02 EK3.4 within the control room team as appropriate to the assigned Rating 3.5 position, in such a way that procedures are adhered to and the limitations in the facilities license and amendments are not violated.

Question 63 The crew is performing step 9, Verify ECCS Flow Not Required, of E-1.1, SI Termination.

What parameter should the Shift Foreman have the operator check first and why is that parameter checked first per E-1.1?

A. Subcooling because it is the most direct indication that there are no voids in the upper head of the reactor vessel.

B. Pressurizer level because if it is on scale, normal pressure control will be possible.

C. Subcooling because it is the most direct indication that there is adequate core cooling.

D. Pressurizer level because if on scale, then there are no voids in the RCS.

Proposed Answer: C. Subcooling because it is the most direct indication that there is adequate core cooling.

Explanation:

A. Incorrect. While voids in the RCS are not desirable, this is not the reason subcooling is checked first. Also, RVLIS is the most direct indication of voids.

B. Incorrect. Pressurizer level is always checked after subcooling.

C. Correct. Subcooling is the most direct check of adequate core heat removal.

D. Incorrect. Pressurizer level is only reliable if there is subcooling and as such only checked after subcooling.

Technical

References:

E-1.1, Westinghouse Executive Volume - SI Termination References to be provided to applicants during exam: None Learning Objective: 7920S Explain basis of emergency procedure steps (E-1.1, E-1.2)

Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.41.5 Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.

Changed A to be clearly incorrect.

DCPP L091C Exam Rev 1

Examination Outline Cross-Reference Level RO Tier # 1 E03 EA2.2 Ability to determine and interpret the following as Group # 2 they apply to the (LOCA Cooldown and Depressurization): K/A # E03 EA2.2 Adherence to appropriate procedures and operation within the Rating 3.5 limitations in the facility's license and amendments.

Question 64 GIVEN:

• The crew is performing ECCS flow reduction in accordance with E-1.2, Post LOCA Cooldown and Depressurization

• RHR pumps are shutdown

• All ECCS charging and SI pumps are running In accordance with E-1.2, which of the following criteria, if possible, will be used by the crew to reduce ECCS flow?

A. Shutdown both SI pumps first, establish letdown, then a charging pump B. Shutdown both ECCS charging pumps first, then an SI pump C. Shutdown an ECCS charging and SI pump on the same train D. Shutdown an ECCS charging and SI pump on alternate trains Proposed Answer: D. Shutdown an ECCS charging and SI pump on alternate trains.

Explanation:

A incorrect. A charging pump is stopped first.

B incorrect. Only one charging pump is stopped.

C incorrect. Alternate trains are used, if possible.

D correct. Note states that if possible, pumps are stopped on alternate trains.

Technical

References:

E-1.2 References to be provided to applicants during exam: None Learning Objective: 6745 State the general sequence of ECCS reduction sequence Question Source: Bank # B-1164 X (note changes; attach parent) Modified Bank #

New Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.41.5 Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.

Editorial, add iaw procedure to question.

DCPP L091C Exam Rev 1

Examination Outline Cross-Reference Level RO E06 EK2.1 Knowledge of the interrelations between the Tier # 1 (Degraded Core Cooling) and the following: Components, and Group # 2 functions of control and safety systems, including K/A # E06 EK2.1 instrumentation, signals, interlocks, failure modes, and automatic Rating 3.6 and manual features.

Question 65 The crew is performing a cooldown at less than 100°F/hour in accordance with FR-C.2, Degraded Core Cooling.

Which of the following could occur if the crew fails to block Low Main Steam Line Pressure Safety Injection in accordance with FR-C.2?

A. A red path on RCS Integrity Critical Safety Function B. A red path on Core Cooling Critical Safety Function C. Main Steam Isolation D. Reinitiation of Safety Injection Proposed Answer: C. Main Steam Isolation Explanation:

A. Incorrect. A red path P.1 during C.2 could occur when accumulators inject.

B. Incorrect. A red path on C.1 is monitored later in C.2 C. Correct. MSI could occur if not blocked when RCS pressure is less than P-11 D. Incorrect. Auto SI is blocked and will not actuate.

Technical

References:

E-1.2 References to be provided to applicants during exam: None Learning Objective: 6892 Explain the consequences of not maintaining RCS pressure/temperature limits during the EOP set.

Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content: 55.41.7 Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Editorial, add IAW C.2 to question DCPP L091C Exam Rev 1

Examination Outline Cross-Reference Level RO Tier # 3 G2.1.26 Knowledge of industrial safety procedures (such as Group # 1 rotating equipment, electrical, high temperature, high pressure, K/A # G2.1.26 caustic, chlorine, oxygen and hydrogen). Rating 3.4 Question 66 An operator calls and reports the spill of hydrazine in the vicinity of the Auxiliary Feedwater pump chemical addition skid.

Per CP M-9A, Hazardous Materials Incident Initial Emergency Response/Mitigation Procedure, assistance should be immediately requested from:

A. DCPP Safety B. Radiation Protection C. DCPP Fire Department D. Chemistry and Environmental Operations Proposed Answer: C. DCPP Fire Department Explanation:

A. Incorrect.

B. Incorrect.

C. Correct. DCPP Hazardous Materials Emergency Response Team (DCPP Fire Department),

is dispatched.

D. Incorrect.

Technical

References:

CP M-9A References to be provided to applicants during exam: None Learning Objective: 39579 Discuss the appropriate response to various chemical spills Question Source: Bank # DCPP NRC Exam 2/2009 X (note changes; attach parent) Modified Bank #

New Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content: 55.41.10 Administrative, normal, abnormal, and emergency operating procedures for the facility.

DCPP L091C Exam Rev 0

Examination Outline Cross-Reference Level RO Tier # 3 G2.1.40 Knowledge of refueling administrative requirements. Group # 1 K/A # G2.1.40 Rating 2.8 Question 67 In accordance with OP B-8DS2, Core Loading, which of the following is a responsibility of the RO in the Control Room during core loading?

A. Coordinating Core Alterations B. Authorizing an unplanned deviation C. Determining the cause of a Containment radiation monitor high alarm D. Maintaining the record of fuel movement in accordance with the Fuel Movement Tracking Sheet of PEP R-8DS2, "Core Loading Sequence" Proposed Answer: D. Maintaining the record of fuel movement in accordance with the Fuel Movement Tracking Sheet of PEP R-8DS2, "Core Loading Sequence" Explanation:

A. Incorrect. This is the responsibility of the Refueling SRO B. Incorrect. Responsibility of Reactor Engineering C. Incorrect. Responsibility of Refueling SRO D. Correct. Responsibility of RO in the Control Room Technical

References:

OP B-8DS2 References to be provided to applicants during exam: None Learning Objective:

Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content: 55.41.10 Administrative, normal, abnormal, and emergency operating procedures for the facility.

DCPP L091C Exam Rev 0

Examination Outline Cross-Reference Level RO Tier # 3 G2.1.45 Ability to identify and interpret diverse indications to Group # 1 validate the response of another indication. K/A # G2.1.45 Rating 4.3 Question 68 Unit 1 is at full power.

Letdown isolates due to a Pressurizer level instrument failure. The Control Operator states that the controlling level channel has failed.

Which of the following would validate the accuracy of the operator's diagnosis that the controlling channel, and not the backup channel, has failed?

A. Backup heaters turn off B. Letdown orifice valves close C. PK05-22, Pzr Level Hi/Low Control", in alarm D. Charging flow control valve FCV-128 opens to increase charging flow Proposed Answer: D. Charging flow control valve FCV-128 opens to increase charging flow Explanation:

A. Incorrect. Both channels will turn off heaters B. Incorrect. Both channels close the orifice valves C. Incorrect. Both channels will cause the alarm D. Correct. Charging response is the diverse indication that will support the operators diagnosis Technical

References:

OIM A-4 References to be provided to applicants during exam: None Learning Objective: 36923 Analyze automatic features and interlocks associated with the Pzr, Pzr Pressure and Level Control System Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.41.7 Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

DCPP L091C Exam Rev 0

Examination Outline Cross-Reference Level RO Tier # 3 G2.2.6 Knowledge of the process for making changes to Group # 2 procedures. K/A # G2.2.6 Rating 3.0 Question 69 While performing a procedure, it is noted that a minor typographical change is required. The change will not change the intent of the procedure and nuclear safety is NOT affected.

In accordance with AD1.ID2, Procedure Process Control, which of the following is the proper procedure revision process?

A. Normal revision B. On the spot change C. Editorial correction D. Expedited procedure revision Proposed Answer: C. Editorial correction Explanation:

A. Incorrect. Revision for typo or minor corrections is not required B. Incorrect. Change specific pages of an effective procedure when time or the situation does not permit use of a revision.

C. Correct. A process to correct minor errors, clarify wording, and update items such as titles and organizational names.

D. Incorrect. procedure revision that bypasses the normal word processing activities. The XPR is used for urgent and substantive changes that are not appropriate for an On-The-Spot Change.

Technical

References:

References to be provided to applicants during exam: None Learning Objective: 9640 Explain definitions relevant to procedures Question Source: Bank # P-33375 X (note changes; attach parent) Modified Bank #

New Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content: 55.41.10 Administrative, normal, abnormal, and emergency operating procedures for the facility Editorial, added procedure name to question DCPP L091C Exam Rev 1

Examination Outline Cross-Reference Level RO Tier # 3 G2.2.35 Ability to determine Technical Specification Mode of Group # 2 Operation. K/A # G2.2.35 Rating Question 70 Which of the following would meet the Technical Specification definition of a MODE change?

A. Reactor power is raised from 1 to 5%

B. RCS temperature increases from 190°F to 205°F C. The first reactor vessel closure head bolt is tensioned D. The last reactor vessel closure head bolt is de-tensioned Proposed Answer: B. RCS temperature increases from 190°F to 205°F Explanation:

A. Incorrect. In MODE 2 until power is greater than 5%

B. Correct. Heat up above 200F is transition to MODE 4 C. Incorrect. Entry into MODE 5 is the last head bolt tensioned D. Incorrect. Entry into MODE 6 is the first head bolt tensioned Technical

References:

Technical Specifications section 1.0 References to be provided to applicants during exam: None Learning Objective: 9696 Define Technical Specification items found in the Definition Sections Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content: 55.41.10 Administrative, normal, abnormal, and emergency operating procedures for the facility DCPP L091C Exam Rev 0

Examination Outline Cross-Reference Level RO Tier # 3 G2.3.7 Ability to comply with radiation work permit Group # 3 requirements during normal or abnormal conditions. K/A # G2.3.7 Rating 3.5 Question 71 In accordance with RP D-220, Control of Access to High, Locked High, and Very High Radiation Areas, which of the following would meet the minimum requirement for entry into a High Radiation Area?

Individual has been briefed, is entered on a valid Radiation Work Permit and as a MINIMUM:

A. has a radiation monitoring device that continuously integrates the radiation dose rates in the area and alarms when the device's dose alarm set point is reached, with an appropriate alarm set point.

B. has a monitoring device which continuously displays area radiation dose rate and is continuously under the surveillance of a Radiation Protection Technician equipped with a self-reading dosimeter.

C. is accompanied by a Radiation Protection Technician with a neutron radiation monitoring instrument.

D. has a monitoring device which continuously displays area radiation dose rate and is accompanied by a Radiation Protection Technician with a neutron radiation monitoring instrument.

Proposed Answer: A. has a radiation monitoring device that continuously integrates the radiation dose rates in the area and alarms when the device's dose alarm set point is reached, with an appropriate alarm set point.

Explanation:

A. Correct. Per RCP D-220, only necessary to have has a radiation monitoring device that continuously integrates the radiation dose rates in the area and alarms when the device's dose alarm set point is reached, with an appropriate alarm set point..

B. Incorrect. Not necessary to be accompanied by RP C. Incorrect. Not necessary to have neutron monitoring D. Incorrect. Not necessary to have neutron monitoring or be accompanied by RP Technical

References:

RCP D-220 References to be provided to applicants during exam: None Learning Objective:

Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content: 55.41.12 Radiological safety principles and procedures.

Editorial - added procedure number and name to question DCPP L091C Exam Rev 1

Examination Outline Cross-Reference Level RO Tier # 3 G2.3.14 Knowledge of radiation or contamination hazards that Group # 3 may arise during normal, abnormal, or emergency conditions or K/A # G2.3.14 activities. Rating 3.4 Question 72 GIVEN:

• A plant shutdown is in progress because RCS activity levels are greater than allowed by Technical Specifications, when a small break LOCA occurs

• The crew has transitioned to E-1.2, Post LOCA Cooldown and Depressurization

• The crew is now preparing to establish RCP seal return flow

• CCW valves to the RCP have remained open In accordance with E-1.2, prior to opening 8100 and 8112, RCP Seal Water Return Stop Valves, an evaluation should be performed to assess the consequences of which of the following?

A. Inter-system LOCA B. Flashing in the seal water heat exchanger C. Thermal shock to the RCP seals D. Increased radiation levels in the auxiliary building Proposed Answer: D. Increased radiation levels in the auxiliary building Explanation:

Only D is Correct. Caution at step 28 states: If excess activity levels in the RCS are suspected, then an evaluation of the consequences of re-establishing seal return flow should be made prior to placing RCP seal return flow in service.

Technical

References:

E-1.2 References to be provided to applicants during exam: None Learning Objective: 7920S Explain basis of emergency procedure steps (E-1.1, E-1.2)

Question Source: Bank # B-1164 X (note changes; attach parent) Modified Bank #

New Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content: 55.41.12 Radiological safety principles and procedures.

Editorial - added IAW E-1.2 to question DCPP L091C Exam Rev 1

Examination Outline Cross-Reference Level RO Tier # 3 G2.4.19 Knowledge of EOP layout, symbols, and icons Group # 4 K/A # G2.4.19 Rating 3.4 Question 73 When proceeding through Emergency Operating Procedures, a step number preceded by a diamond () , denotes that the step:

A. is on the foldout page.

B. is a continuous action.

C. is an immediate action.

D. impacts Time Critical Operator Actions.

Proposed Answer: D. impacts Time Critical Operator Actions.

Explanation:

A. Incorrect. No symbol for foldout page items B. Incorrect. Step would be enclosed by a box C. Incorrect. Step number would be enclosed in a box.

D. Correct. Diamonds signify steps whose performance impacts TCOAs Technical

References:

AD1.DC12, E-0 References to be provided to applicants during exam: None Learning Objective:

Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content: 55.41.10 Administrative, normal, abnormal, and emergency operating procedures for the facility DCPP L091C Exam Rev 0

Examination Outline Cross-Reference Level RO Tier # 3 G2.4.28 Knowledge of procedures relating to a security event Group # 4 (non-safeguards information). K/A # G2.4.28 Rating 3.2 Question 74 The crew is going to perform OP J-6C:II, Diesel Fuel Oil System - Alignment Verification for Plant Startup.

According to OP J-6C:II, who must be contacted prior to opening the pump vault door?

A. Mechanical Maintenance B. Security C. The Control Room D. Electrical Maintenance Proposed Answer: B. Security Explanation:

A incorrect. Security must be contacted.

B is correct. Prior to opening the door to the vault, security is contacted.

C incorrect. While the operator may call the CR, it is not mandated by the procedure D incorrect. Security must be contacted.

Technical

References:

OP J-6C:II References to be provided to applicants during exam: None Learning Objective:

Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content: 55.41.10 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Added IAW procedure, which makes C incorrect.

DCPP L091C Exam Rev 1

Examination Outline Cross-Reference Level RO Tier # 3 G2.4.46 Ability to verify that the alarms are consistent with the Group # 4 plant conditions K/A # G2.4.46 Rating 4.2 Question 75 GIVEN:

• 68% reactor power.

• One Shutdown Bank A Rod has dropped into the core.

• The crew is recovering the dropped rod.

PK03-17, ROD CONT URGENT FAILURE is received when rod withdrawal begins.

Which of the following describes the Rod Control System Urgent Failure alarm and the plant response?

The alarm is:

A. unexpected. Rod withdrawal will not occur until the alarm is reset at the Logic Cabinet.

B. expected. The alarm will have to be reset to allow rod recovery to continue.

C. unexpected. Rod withdrawal will not occur until the alarm is reset at the Power Cabinet.

D. expected. Rod withdrawal is unaffected and recovery may continue.

Proposed Answer: D. expected. Rod withdrawal is unaffected and recovery may continue.

Explanation:

A. Incorrect. Would be correct if on SD Bank C or D B. Incorrect. The alarm is expected due to no group movement in one group C. Incorrect. Alarm is expected.

D. Correct. Rod motion is unaffected.

Technical

References:

PK03-17, OP AP-12C References to be provided to applicants during exam: None Learning Objective: 36990 Describe controls, indications and alarms associated the Rod Control System Question Source: Bank # Wolf Creek 2007 X (note changes; attach parent) Modified Bank #

New Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.41.2 General design features of the core, including core structure, fuel elements, control rods, core instrumentation, and coolant flow.

DCPP L091C Exam Rev 0

Examination Outline Cross-Reference Level SRO Tier # 1 EPE011 EA2.03 Ability to determine or interpret the following as Group # 1 they apply to a Large Break LOCA: Consequences of managing K/A # 011 EA2.03 LOCA with loss of CCW Rating 4.2 Question SRO 1 (76)

GIVEN:

• A LOCA has occurred

• The crew is performing the actions of E-1, Loss of Reactor or Secondary Coolant

• No CCW pumps are running

• The Work Control Lead is reviewing OP AP-11, Malfunction of Component Cooling Water

• All ECCS pumps and RCPs are secured due to the loss of cooling

• Core exit thermocouples are 620°F and rising Based on a recommendation from the Work Control Lead, the Shift Foreman considers directing the operator to start Charging Pump 1-3.

What action should be taken by the Shift Foreman regarding starting of the 1-3 Charging pump?

A. Do not direct the action because it is not a step in E-1.

B. Invoke 50.54x, notify the NRC and then direct the operator to start the pump.

C. Pull a step forward from E-1 forward and then direct the operator to start the 1-3 CCP.

D. Direct the operator to start the pump, this is allowed by OP1.DC10, Conduct of Operations.

Proposed Answer: D. Direct the operator to start the pump, this is allowed by OP1.DC10, Conduct of Operations.

Explanation:

A. Incorrect. Action outside of the procedure may be taken to protect plant equipment and/or protect the health and safety of the public.

B. Incorrect. There is procedural guidance, for instance in OP AP-11 to start the CCP if no CCW pumps are running. Additionally, the NRC is contacted after the action is taken, not before.

C. Incorrect. There is no step in E-1 to start the 13 CCP. The pump is stopped in E-0, appendix E.

D. Correct. Action can be taken to mitigate the accident, which protects the health and safety of the public.

Technical

References:

OP AP-11, OP1.DC10 References to be provided to applicants during exam: None Learning Objective:

Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental X DCPP L091C Exam Rev 1

Comprehensive/Analysis 10CFR Part 55 Content: 55.43.3 Facility licensee procedures required to obtain authority for design and operating changes in the facility..

Reworded question to say "considers directing" DCPP L091C Exam Rev 1

Examination Outline Cross-Reference Level SRO Tier # 1 APE015/017 G2.4.49 RCP Malfunctions: Ability to perform Group # 1 without reference to procedures those actions that require K/A # APE015/017 immediate operation of system components and controls. G2.4.49 Rating 4.4 Question SRO 02 (77)

GIVEN:

• The crew is performing step 19, Restore Temporary Core Cooling of FR-C.1, Response to Inadequate Core Cooling

• Core Exit Thermocouples are 1350°F and rising

• RCP seal cooling was lost to all RCPs but has been re-established

• Narrow Range Steam Generator Levels:

o 11 = 35%

o 12 = 50%

o 13 = 5%

o 14 = 0%

Per the guidance of FR-C.1, what action should the Shift Foreman take regarding starting the RCPs?

A. Wait for an evaluation of the RCP seals to be performed prior to starting any RCP.

B. Direct the operator to start either the 1-3 or the 1-4 RCP.

C. Direct the operator to start the 1-2 RCP.

D. Direct the operator to start the 1-1 RCP.

Proposed Answer: D. Direct the operator to start the 1-1 RCP.

Explanation: the RCP malfunction is the loss of seal cooling. The SFM must realize that in C.1, it is desired but not required to have normal conditions established to start an RCP.

A. Incorrect. The evaluation is not required, an RCP must be started, if a heat sink is available in that loop in an effort to restore temporary core cooling.

B. Incorrect. Neither loop has sufficient inventory to start an RCP.

C. Incorrect. Loop 2, although the highest level, should be started last.

D. Correct. Loop 1 has a heat sink and would be started before starting the RCP in Loop 2.

Technical

References:

FR-C.1 References to be provided to applicants during exam: None Learning Objective: 5712 Identify existing RCP start criteria in emergency procedures Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.43.5 Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations Editorial change, added "Per C.1" to question DCPP L091C Exam Rev 1

Examination Outline Cross-Reference Level SRO Tier # 2 EPE038 EA2.11 Ability to determine or interpret the following as Group # 1 they apply to a SGTR: Local radiation reading on main steam K/A # EPE038 lines EA2.11 Rating 3.9 Question SRO 03 (78)

GIVEN:

• RM-15R indicates a leak rate of approximately 200 gpd

• A plant shutdown in accordance with OP AP-3, Tube Leak is in progress The plant trips due to loss of 12 kV bus D.

What action should be taken by the Shift Foreman concerning the tube leak?

A. Direct the operator to initiate Safety Injection and then transition to E-3, Steam Generator Tube Rupture, from E-0, Reactor Trip or Safety Injection.

B. At step 4 of E-0, Reactor Trip or Safety Injection, transition to OP AP-3 to complete the recovery and then transition to E-0.1, Reactor Trip Response.

C. At step 4 of E-0, Reactor Trip or Safety Injection, transition to E-0.1, Reactor Trip Response and when complete, then complete the actions OP AP-3.

D. At step 4 of E-0, Reactor Trip or Safety Injection, transition to E-0.1, Reactor Trip Response; OP AP-3 does not need to be completed.

Proposed Answer: C. At step 4 of E-0, Reactor Trip or Safety Injection, transition to E-0.1, Reactor Trip Response and when complete, then complete OP AP-3.

Explanation:

A. Incorrect. OP AP-3 will handle the tube leak, initiation of SI to get to E-3 is not required.

B. Incorrect. EOPs take precedence over the AP C. Correct. Per AP-3, the procedure must be completed once the EOPs are completed.

D. Incorrect. AP-3 must be implemented and completed.

Technical

References:

OP AP-3, LPA3 References to be provided to applicants during exam: None Learning Objective: 3794 Given initial conditions, assumptions, and symptoms, predict the operational implications for any size S/G tube leak.

Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.43.5 Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Editorial changes to B (then transition to AP-3) and C (remove redundant "and then")

DCPP L091C Exam Rev 1

Examination Outline Cross-Reference Level SRO Tier # 1 APE040 AA2.02 Ability to determine and interpret the following Group # 1 as they apply to the Steam Line Rupture: Conditions requiring a K/A # APE040 reactor trip AA2.02 Rating 4.7 Question SRO 04 (79)

Unit 1 is at full power.

PK09-03, SG 1-3, Press, Lvl, Flow alarms. Reactor power begins to increase and the crew confirms a safety is lifting and has not reseated on Steam Generator 1-3.

In accordance with PK09-03, what should be the initial action taken by the Shift Foreman?

A. Direct the operator to insert control rods to lower power below 100% and go to OP AP-25, Rapid Load Reduction or Shutdown.

B. Direct the operator to trip the reactor and go to E-0, Reactor Trip or Safety Injection.

C. Direct the operator to insert control rods to lower power below 100% and go to OP AP-6, Emergency Boration.

D. Direct the operator to lower turbine load to reduce reactor power to less than 87% to meet the Technical Specification required action for an inoperable safety valve.

Proposed Answer: B. Direct the operator to trip the reactor and go to E-0, Reactor Trip or Safety Injection.

Explanation:

A. Incorrect. Inserting control rods will not lower secondary load, which will remain at greater than 100%. IF a trip was not required, the action would be to initiate a ramp (to lower turbine load) to accommodate the steam flow from the safety and go to AP-25.

B. Correct. A trip is required due to the inability to get the safety to close.

C. Incorrect. An overpower is a logical candidate for an emergency boration.

D. Incorrect. Reducing load would be required if the unit was to remain at power. The unit would not be kept at power with the safety continuing to lift.

Technical

References:

AR PK09-03, OP AP-6, OP AP-25, LCO 3.7.

References to be provided to applicants during exam: None Learning Objective: 7927 Given initial conditions and assumptions, determine if a reactor trip or safety injection actuation is required Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.43.5 Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Replaced D, (initiate SI) because the action is not immediate and out of a PK, it was felt that it maybe reasonable to do so. Added IAW PK to question.

DCPP L091C Exam Rev 1

Examination Outline Cross-Reference Level SRO Tier # 1 EPE 055 G2.4.4 Station Blackout: Ability to recognize abnormal Group # 1 indications for system operating parameters that are entry-level K/A # EPE055 conditions for emergency and abnormal operating procedures. G2.4.4 Rating 4.7 Question SRO 05 (80)

GIVEN:

• Unit 2 has completed a refueling outage

• RCS temperature is 195°F A loss of offsite power occurs. None of the Emergency Diesels start.

Which of the following should be performed by the Shift Foreman and in accordance with OP1.DC10, Conduct of Operations, should the Shift Foreman "Go to" or "Refer to" the procedure?

A. Go to OP AP SD-1, Loss of AC Power.

B. Refer to OP AP SD-1, Loss of AC Power.

C. Go to EOP ECA-0.0, Loss of All AC Power.

D. Refer to EOP ECA-0.0, Loss of All AC Power.

Proposed Answer: A. GO TO OP AP SD-1, Loss of AC Power.

Explanation:

A. Correct. GO TO - Immediately leave the procedure in effect and go to the designated step. SD-1 applies in MODES 5 and 6.

B. Incorrect. REFER TO - The designated procedure may be used for guidance at the Operator's discretion. There is no discretion, the SFM will enter and perform SD-1.

C. Incorrect. ECA-0.0 applies in MODES 1 thru 4 D. Incorrect. ECA-0.0 applies in MODES 1 thru 4 Technical

References:

WOG EOP users guide (Modes of applicability)

OP1.DC10, attachment 7.1, Emergency Operating Procedures, Synopsis of Rules of Usage OP SD-1, Scope step 1.1 References to be provided to applicants during exam: None Learning Objective: 3552 - Given initial conditions, assumptions, and symptoms, determine the correct Emergency Operating Procedure to be used to mitigate an operational event Question Source: Bank #

(note changes; attach parent) Modified Bank # 78 DCPP NRC exam 1/10 X New Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.43.5 Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Addded to question "IAW OP1.DC10 and "Go to" and "Refer to" to question.

DCPP L091C Exam Rev 0

Examination Outline Cross-Reference Level SRO Tier # 1 APE 062 G2.4.20 Loss of Nuclear Service Water: Knowledge of Group # 1 operational implications of EOP warnings, cautions, K/A # APE062 and notes. G2.4.20 Rating 4.3 Question SRO 06 (81)

According to CP M-5, Response to Tsunami Warning, Attachment 10.1, Aux Salt Water Pump Operation, which of the following is an aspect of the attachment that should be considered for the Tsunami assumed in the FSAR?

A. If water level is observed to be above the main deck of the intake structure, both reactors should be tripped.

B. The Tsunami should be expected to flood the Aux Saltwater ventilation shaft openings.

C. The affected reactor and ASW pumps should be tripped if PK01-02, "AUX SALT WTR PPS ROOM," is in alarm when the Tsunami Warning is issued.

D. During the Tsunami drawdown, the Aux Saltwater pumps should be stopped and not restarted until after the Tsunami has crested.

Proposed Answer: A. If water level is observed to be above the main deck of the intake structure, both reactors should be tripped.

Explanation:

A. Correct. If CCW temperatures exceed 120°F or there is gross flooding, actions may apply (reactors tripped) to both units to restore ASW.

B. Incorrect. The shaft openings are above the postulated tsunami.

C. Incorrect. If the alarm is in, security is instructed to close the door, if possible.

D. Incorrect. The pumps are stopped if cavitating.

Technical

References:

CP M-5, LPE7 References to be provided to applicants during exam: None Learning Objective: 3974 Explain the conditions that require the reactor to be tripped during a Tsunami Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content: 55.43.1 Conditions and limitations in the facility license.

Replaced A to be more specific.

DCPP L091C Exam Rev 1

Examination Outline Cross-Reference Level SRO Tier # 1 APE 028 AA2.04 Ability to determine and interpret the following Group # 2 as they apply to the Pressurizer Level Control Malfunctions: K/A # APE028 Ammeters and running indicators for CVCS charging pumps AA2.04 Rating 3.1 Question SRO 07 (82)

Unit 1 is at 100% power.

The operator reports the following:

• Pressurizer level is lowering slowly

• Pressurizer backup heaters are OFF

• Charging flow indicates 0 gpm

• CCP 1-3 amps are approximately 10 amps with a slight oscillation

• FCV-128, CCP Flow Control Valve, demand indicates 100%

• Seal injection is 0 gpm to each RCP

• Letdown is in service at 75 gpm

• VCT level is rising Which of the following describes the likely failure that has occurred causing Pressurizer level to lower and what action should be taken by the Shift Foreman?

A. The charging pump shaft has sheared, go to OP AP-17, Loss of Charging, section A, Loss of All Charging.

B. Loss of instrument air to FCV-128 has caused it to fail closed, go to OP AP-17, Loss of Charging, section B, Charging System Equipment Malfunctions.

C. A large leak upstream of FCV-128 has occurred, go to OP AP-17, Loss of Charging, section C, Charging Line Leak at Power.

D. The controlling Pressurizer level channel has failed high, go to OP AP-5, Malfunction of Eagle 21 Protection or Control Channel.

Proposed Answer: A. The charging pump shaft has sheared, go to OP AP-17, Loss of Charging, section A, Loss of All Charging.

Explanation:

A. Correct. The charging pump is not operating (low amps, no flow, rising VCT level).

Section A of OP AP-17 addresses the loss of charging.

B. Incorrect. FCV-128 fails open on a loss of instrument air.

C. Incorrect. VCT level would be lowering.

D. Incorrect. If the controlling channel failed high, the back heaters would be energized due to level deviation.

Technical

References:

OIM page A-4-2b, B-1-1, OP AP-17A, B, C References to been provided to applicants during exam: None Learning Objective: 3478 Given initial conditions, assumptions, and symptoms, determine the correct abnormal operating procedure to be used to mitigate an operational event.

Question Source: Bank #

DCPP L091C Exam Rev 1

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.43.5 Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Removed "rcs pressure stable" from setup. Thought to be confusing given that pressurizer level is lowering and heaters are off.

DCPP L091C Exam Rev 1

Examination Outline Cross-Reference Level SRO Tier # 1 APE 051 AA2.01 Ability to determine and interpret the following Group # 2 as they apply to the Loss of Condenser Vacuum: Cause for low K/A # APE051 vacuum condition AA2.01 Rating 2.7 Question SRO 08 (83)

GIVEN:

• A turbine load increase is in progress with all systems operating normally in automatic control

• Turbine Load is 400 MWe The operator reports the following:

• Condenser Pressure Recorder PR-11A and B both show condenser pressure is slowly rising.

• Condenser Pressure PI-44 reads 4.0Hg Abs

• Condenser differential pressure is 7 psid and stable on all quadrants

• PK10-11 CONDENSER PRESS/LEVEL is in alarm Which of the following actions should be taken by the Shift Foreman?

A. Direct the operator to trip the turbine and go to AP-29, Main Turbine Malfunction, due to high condenser pressure.

B. Direct the operator to trip the turbine and go to AP-29, Main Turbine Malfunction, due to high quadrant DP.

C. Go to OP AP-7, Degraded Condenser, section A, Loss of Condenser Vacuum, and reduce load as necessary to restore condenser pressure to within operating limits.

D. Go to OP AP-7, Degraded Condenser, section B, Condenser Fouling, and reduce load to remove a Circulating Water pump from operation to lower condenser differential pressure.

Proposed Answer: C. Go to OP AP-7, Degraded Condenser, section A, Loss of Condenser Vacuum, and reduce load as necessary to restore condenser pressure to within operating limits.

Explanation:

A. Incorrect. The nominal setpoint for tripping on low vacuum is 7.2 inches Hg.

B. Incorrect. High quadrant DP trip setpoint is any DP greater than 10 psid and increasing rapidly or, all halves greater than 10 or any greater than 13 psid.

C. Correct. The problem is condenser vacuum, the action is to reduce load in an effort to stabilize vacuum.

D. Incorrect. DP is stable Technical

References:

AR PK10-11, OP AP-7, section A and B References to be provided to applicants during exam: None Learning Objective: 3477G Given an abnormal condition, summarize the major actions of OP AP-7 to mitigate an event in progress.

Question Source: Bank # 83 DCPP NRC Exam 1/2010 X (note changes; attach parent) Modified Bank #

DCPP L091C Exam Rev 0

New Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.43.5 Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

DCPP L091C Exam Rev 0

Examination Outline Cross-Reference Level SRO Tier # 1 WE01 Rediagnosis: Knowledge of EOP mitigation strategies Group # 2 K/A # WE01 G2.4.6 Rating 4.7 Question SRO 09 (84)

The crew has transitioned to E-1, Loss of Reactor or Secondary Coolant from E-0, Reactor Trip or Safety Injection based on high Containment pressure.

At step 2 of E-1, a check is made for a faulted steam generator.

What is the basis for the Shift Foreman checking for a faulted steam generator at this point in E-1?

A. To ensure there is a heat sink available.

B. To check for an uncontrolled cooldown.

C. To alert the crew to a possible misdiagnosis or subsequent steam generator failure which must be isolated.

D. The diagnosis step for a LOCA based on Containment parameters occurs before the check for a faulted steam generator in E-0.

Proposed Answer: C. To alert the crew to a possible misdiagnosis or subsequent steam generator failure which must be isolated.

Explanation:

A. Incorrect. Heat sink is checked later when checking steam generator level. If the LOCA was large, steam generators are not needed for heat removal.

B. Incorrect. Basis is to check for a faulted steam generator, there is no action taken to control temperature.

C. Correct. Per the background document, An uncontrolled SG pressure decrease or a completely depressurized (i.e., near containment or atmospheric pressure) SG indicates a failure of the secondary pressure boundary. If it cannot be verified that all faulted SG(s) steamlines and feedlines are isolated, the operator is instructed to leave E-1 and transfer to E-2, FAULTED STEAM GENERATOR ISOLATION, to perform the isolation actions.

Therefore, this step alerts the operator to a possible misdiagnosis or subsequent failure.

D. Incorrect. E-2 is the first accident checked in E-0 Technical

References:

E-0, E-1 Background and E-1 References to be provided to applicants during exam: None Learning Objective: 7920E Explain basis of emergency procedure steps (E-2) including:

Bases for TCOAs with operator actions of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or less.

Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental X DCPP L091C Exam Rev 0

Comprehensive/Analysis 10CFR Part 55 Content: 55.43.5 Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

DCPP L091C Exam Rev 0

Examination Outline Cross-Reference Level SRO Tier # 1 E09 G2.4.1 Natural Circulation: Knowledge of the parameters Group # 2 and logic used to assess the status of safety functions, such as K/A # E09 G2.4.21 reactivity control, core cooling and heat removal, reactor Rating 4.6 coolant system integrity, containment conditions, radioactivity release control, etc.

Question SRO 10 (85)

GIVEN:

• The crew is cooling down the RCS in accordance with E-0.2, Natural Circulation Cooldown

• All Critical Safety Functions, except for Heat Sink, are GREEN

• Heat Sink Critical Safety Function is YELLOW due to steam generator narrow range levels less than 15%

In accordance with F-0, Critical Safety Function Status Trees, what is the frequency the WCSFM is required to be monitoring the Critical Safety Functions while in E-0.2?

A. Continuously until the crew exits the EOPs.

B. Every 10 to 20 minutes as long as plant conditions remain stable.

C. Every 10 to 20 minutes until the Heat Sink Critical Safety Function is GREEN, then monitoring can be suspended.

D. Monitoring can be suspended because E-0.2 procedure is a non-accident (no SI required)

EOP.

Proposed Answer: B. Every 10 to 20 minutes as long as plant conditions remain stable.

Explanation:

A. Incorrect. Continuous monitoring is required when plant conditions are changing rapidly.

In E-0.2, plant conditions are not unstable, and are under the control of the operator.

B. Correct. Per F-0, IF extreme or severe challenges do NOT exist, AND plant conditions are NOT changing rapidly, THEN critical safety function status trees shall be monitored every 10 to 20 minutes.

C. Incorrect. All CSFs Green is not a requirement for suspension of monitoring, it is required while in the EOPs.

D. Incorrect. While there is no SI present, as there is for rest of the EOP network, while in the EOP, monitoring is required.

Technical

References:

F-0, LPE-FR References to be provided to applicants during exam: None Learning Objective: 38107 Apply the Rules of Usage in EOPs for the CSFSTs and FRGs, including:

• when to monitor and/or implement the CSFSTs and FRGs Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental X DCPP L091C Exam Rev 1

Comprehensive/Analysis 10CFR Part 55 Content: 55.43.5 Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Editorial - added IAW F-0 to question.

DCPP L091C Exam Rev 1

Examination Outline Cross-Reference Level SRO Tier # 2 APE 003 G2.4.8 RCP: Knowledge of how abnormal operating Group # 1 procedures are used in conjunction with EOPs. K/A # APE003 G2.4.8 Rating 4.5 Question SRO 11 (86)

GIVEN:

• The crew has isolated CCW to RCP 1-1 thermal barrier by entering Containment and closing CCW 234 in accordance with OP AP-28, Reactor Coolant Pump Malfunction, Section E, Loss of CCW to a RCP or RCP High Temperature

• A rapid plant shutdown to remove RCP 1-1 from service is in progress

• Reactor power is 28%

The seal injection filter clogs and seal injection lowers to 0 gpm on all RCPs.

In accordance with OP AP-28, which of the following actions should be taken by the Shift Foreman?

A. Direct the operator to trip the reactor, trip the 1-1 RCP and go to E-0, Reactor Trip or Safety Injection, while continuing with OP AP-28.

B. Direct the operator to trip the reactor, trip all the RCPs and go to E-0, Reactor Trip or Safety Injection, while continuing with OP AP-28.

C. Direct the Aux Building watch to immediately swap seal injection filters, if unsuccessful in 5 minutes, direct the operator to trip the 1-1 RCP and go to OP AP-25, Rapid Load Reduction or Shutdown, to perform a plant shutdown.

D. Direct the Aux Building watch to immediately swap seal injection filters, if unsuccessful in 5 minutes, direct the operator to trip the reactor, trip all the RCPs and go to E-0, Reactor Trip or Safety Injection.

Proposed Answer: A. Direct the operator to trip the reactor, trip the 1-1 RCP and go to E-0, Reactor Trip or Safety Injection, while continuing with OP AP-28.

Explanation:

A. Correct. Per OP AP-28, caution in section D and per step 1 of section F, an immediate trip and trip of the AFFECTED RCP is required if seal injection and CCW thermal barrier cooling is lost.

B. Incorrect. Loss of CCW to all RCPs would require a trip of all RCPs, but cooling flow is only lost to one RCP. A trip of all RCPs is not appropriate and would remove forced flow (the preferred source) as the decay heat removal mechanism.

C. Incorrect. Regardless of the power level, if an RCP must be stopped, the reactor is tripped first.

D. Incorrect. Only 1 RCP must be tripped and it must be immediately. 5 minutes is the time to restore CCW cooling to RCP lube oil coolers.

Technical

References:

OP AP-28 References to be provided to applicants during exam: None DCPP L091C Exam Rev 1

Learning Objective: 7927 Given initial conditions and assumptions, determine if a reactor trip or safety injection actuation is required Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.43.5 Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Editorial - Added "IAW OP AP-28" to question.

DCPP L091C Exam Rev 1

Examination Outline Cross-Reference Level SRO Tier # 2 005 A2.04 Ability to (a) predict the impacts of the following Group # 1 malfunctions or operations on the RHRS, and (b) based on those K/A # 005A2.04 predictions, use procedures to correct, control, or mitigate the Rating 2.9 consequences of those malfunctions or operations: RHR valve malfunctions Question SRO 12 (87)

Unit 1 is in MODE 4.

The operator reports:

• PK02-16, RHR System, is in alarm due to low RHR pump discharge flow

• PK05-25, PRT Press/Lvl/Temp, is in alarm due to PRT level rising rapidly

• RCS pressurizer level is lowering rapidly Which of the following actions should be taken by the Shift Foreman?

A. Go to OP AP-16, Malfunction of the RHR System, for a break of the RHR system piping in the RHR pump room.

B. Go to OP AP-16, Malfunction of the RHR System, for a failed RHR relief valve.

C. Go to OP AP SD-2, Loss of RCS Inventory, for a break of the RHR piping in the RHR pump room.

D. Go to OP AP SD-2, Loss of RCS Inventory, for a failed RHR relief valve.

Proposed Answer: B. Go to OP AP-16, Malfunction of the RHR System, for a failed RHR relief valve.

Explanation:

A. Incorrect. If the break were on the RHR system, the RHR sump pumps would be running and PRT level would not be rising.

B. Correct. A failed relief valve would lower discharge flow and the water would be directed to the PRT.

C. Incorrect. SD-2 is used in MODES 5 and 6 and the indications are not correct for an RHR system break.

D. Incorrect. SD-2 is used in MODES 5 and 6.

Technical

References:

AR PK02-16, AR PK05-25, AP SD-2, AP-16 References to be provided to applicants during exam: None Learning Objective: 3478 Given initial conditions, assumptions, and symptoms, determine the correct abnormal operating procedure to be used to mitigate an operational event Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.43.5 Assessment of facility conditions and selection DCPP L091C Exam Rev 0

of appropriate procedures during normal, abnormal, and emergency situations.

DCPP L091C Exam Rev 0

Examination Outline Cross-Reference Level SRO Tier # 2 039 A2.04 Ability to (a) predict the impacts of the following Group # 1 malfunctions or operations on the MRSS; and (b) based on K/A # 039 A2.04 predictions, use procedures to correct, control, or mitigate the Rating 3.7 consequences of those malfunctions or operations:

Malfunctioning steam dump Question SRO 13 (88)

GIVEN:

• Unit 2 has tripped and the crew has just transitioned to E-0.1, Reactor Trip Response

• RCS temperature is 542°F and lowering

• RCS pressure is 1950 psig and lowering

• Pressurizer level is 12% and lowering

• The BOPCO reports that two condenser steam dump valves are open In accordance with E-0.1, which of the following actions should be taken by the Shift Foreman?

A. Direct the crew to initiate safety injection and return to E-0, Reactor Trip or Safety Injection, step 1.

B. Direct the crew to initiate safety injection and return to E-0, Reactor Trip or Safety Injection, step 4.

C. Direct the crew to close the MSIVs and adjust the 10% steam dumps to control at 1040 psig (8.67 turns).

D. Direct the crew to close the MSIVs and adjust the 10% steam dumps to control at less than or equal to 1005 psig (8.38 turns).

Proposed Answer: D. Direct the crew to close the MSIVs and adjust the 10% steam dumps to control at less than or equal to 1005 psig (8.38 turns).

Explanation:

A. Incorrect. Subcooling, pressure and pressurizer level are all above the points requiring SI actuation.

B. Incorrect. If SI was required, the proper transition would be to return to step 1 of E-0, not the point of transitioning out of E-0.

C. Incorrect. 1040 psig is used in E-3 for the ruptured steam generator.

D. Correct. The unit is below P-11 and the dumps should be closed. The MSIVs (and bypass valves) are closed to stop the cooldown.

Technical

References:

E-0.1, steam tables References to be provided to applicants during exam: None Learning Objective: 3552 Given initial conditions, assumptions, and symptoms, determine the correct Emergency Operating Procedure to be used to mitigate an operational event Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No DCPP L091C Exam Rev 1

Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.43.5 Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Lowered pressurizer level to be closer to manual SI actuation setpoint. Added Foldout page to question reference and "IAW E-0.1" to question DCPP L091C Exam Rev 1

Examination Outline Cross-Reference Level SRO Tier # 2 061 A2.07 Ability to (a) predict the impacts of the following Group # 1 malfunctions or operations on the AFW system; and (b) based on K/A # 061 A2.07 those predictions, use procedures to correct, control, or mitigate Rating 3.5 the consequences of those malfunctions or operations: Air or MOV failure Question SRO 14 (89)

GIVEN:

• Unit 1 trips from full power

• The crew is performing E-0, Reactor Trip or Safety Injection

• 4 kV Bus G is de-energized

• None of MDAFW pumps can be started

• Narrow range steam generator levels are 12% in all Steam Generators

• Safety Injection has not actuated Which of the following actions should be taken by the Shift Foreman as a transition is made from E-0?

A. Go to E-0.1, Reactor Trip Response, the turbine driven AFW pump is supplying all four steam generators.

B. Go to FR-H.1, Response to Loss of Secondary Heat Sink, there is a RED path on Secondary Heat Sink because the TDAFW level control valves, LCV-106, 107, 108 and 109 are closed due to loss of Bus G.

C. Go to E-0.1, Reactor Trip Response, and direct the operator to locally open the TDAFW level control valves, LCV-106, 107, 108 and 109 to establish AFW flow from the TDAFW pump.

D. Go to FR-H.1, Response to Loss of Secondary Heat Sink, there is a RED path on Secondary Heat Sink because the TDAFW Steam Supply valve, FCV-95 did not open due to the loss of Bus G.

Proposed Answer: A. Go to E-0.1, Reactor Trip Response, the turbine driven AFW pump is supplying all four steam generators.

Explanation:

A. Correct. The TDAFW pump is running, it starts on 2/4 steam generators less than 15%.The TDAFW pump LCVs are powered from Bus G, however, are left full open and will remain open. Full flow from the pump will be supplying all four steam generators.

B. Incorrect. The LCVs remain open despite the loss of power to them, therefore, there is no challenge to Heat Sink CSF.

C. Incorrect. The valves will be fully open due to the loss of power and will have to locally throttled when heat sink is greater than 16%.

D. Incorrect. FCV-95 is DC powered.

Technical

References:

LD-1, F-0 References to be provided to applicants during exam: None Learning Objective: 8405 State the power supplies to Auxiliary Feed Water System DCPP L091C Exam Rev 1

components.

Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.43.5 Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Editorial - replaced "from low power" to "from full power". This ensures that the low levels in the steam generator are an expected response.

DCPP L091C Exam Rev 1

Examination Outline Cross-Reference Level SRO 076 G2.2.25 Service Water (ASW): Knowledge of the bases in Tier # 2 Technical Specifications for limiting conditions for operations Group # 1 and safety limits. K/A # 076 G2.2.25 Rating 4.2 Question SRO 15 (90)

What would be the impact on the OPERABILITY of a train of the Aux Saltwater System (ASW), if a pump vault drain check valve or both vacuum relief valves in that train of ASW were to fail?

A. Neither of these are components covered by Technical Specifications.

B. The failure of both vacuum relief valves would cause the train to be inoperable, the pump vault drain check valve does not impact the OPERABILITY of the train.

C. The failure of the pump vault drain check valve would cause the train to be inoperable, the vacuum relief valves do not impact the OPERABILITY of the train.

D. The failure of either the pump vault drain check valve or both of the vacuum relief valves would cause the train to be inoperable.

Proposed Answer: D. The failure of either the pump vault drain check valve or both of the vacuum relief valves would cause the train to be inoperable.

Explanation:

A. Incorrect. Per the bases, both are required for the train to be OPERABLE.

B. Incorrect. The pump vault drain check valve also must be OPERABLE.

C. Incorrect. At least one of the two vacuum relief valves must be OPERABLE.

D. Correct. Per the bases for Tech Spec 3.7.8, An ASW train is considered OPERABLE during MODES 1, 2, 3, and 4 when:

a. The pump is OPERABLE; and

b. The associated piping, valves, heat exchanger, and instrumentation and controls required to perform the safety related function are OPERABLE.

This requires that at least one vacuum relief valve be OPERABLE. Each ASW train has a vacuum relief system consisting of two vacuum relief valves (check valves) which function to prevent water hammer in the system piping during an ASW pump trip and restart transient.

The associated pump vault drain check valve is OPERABLE. The ASW pump vault check valves prevent flooding of the ASW pump vaults during design flood events Technical

References:

Tech Spec Bases 3.7.8 References to be provided to applicants during exam: None Learning Objective: 9694G Apply TS 3.7 Technical Specification bases Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content: 55.43.2 Facility operating limitations in the technical specifications and their bases.

Editorial - removed "if any" from question and "none" from A.

DCPP L091C Exam Rev 1

Examination Outline Cross-Reference Level SRO Tier # 2 015 G2.4.49 NIS: Ability to perform without reference to Group # 2 procedures those actions that require immediate operation of K/A # 015 G2.4.49 system components and controls. Rating 4.4 Question SRO 16 (91)

GIVEN:

• The crew is performing ECA-2.1, Uncontrolled Depressurization of All Steam Generators

• The operator has reduced AFW to approximately 25 gpm to each steam generator

• RCS pressure is 1500 psig

• Wide range steam generator levels are approximately 20 % in all steam generators The STA reports the following:

• Heat Sink Critical Status tree is RED due to low AFW flow and low steam generator levels

• Subcriticality Status Tree is MAGENTA due to a positive startup rate on both Intermediate Range nuclear instrument channels What action should be taken by the Shift Foreman?

NOTE:

FR-S.1 - Response to Nuclear Power Generation/ATWS FR-H.1 - Response to Loss of Secondary Heat Sink A. Continue to direct the actions for ECA-2.1; the Critical Safety Function status trees are monitored for "information only" until pressure begins to increase in at least one steam generator.

B. Go directly to FR-S.1 to direct action to make the reactor subcritical; a challenge to the Heat Sink Critical Safety Function is expected due to intentional operator action and does not have to be addressed.

C. Go to FR-H.1 and perform the first step to verify procedure performance is not required, then transition to FR-S.1 to direct action to make the reactor subcritical.

D. Go to FR-H.1, trip the RCPs and initiate bleed and feed, then transition to FR-S.1.

Proposed Answer: C. Go to FR-H.1 and perform the first step to verify procedure performance is not required, then transition to FR-S.1 to direct action to make the reactor subcritical.

Explanation:

A. Incorrect. Unlike other EOPs, such as E-1.3 or ECA-0.0, valid challenges to the CSFs are addressed during ECA-2.1 B. Incorrect. Although only the initial step of H.1 will be performed, the RED path must be addressed.

C. Correct. A check that the RED path is due to operator action is performed, then the subcriticality challenge due to the positive SUR is addressed.

D. Incorrect. If the H.1 procedure was performed, this would be the correct action to take.

Technical

References:

F-0, FR-H.1 DCPP L091C Exam Rev 1

References to be provided to applicants during exam: None Learning Objective: 38107 Apply the Rules of Usage in EOPs for the CSFSTs and FRGs, including:

• the six status trees

• the priority of use of the status trees

• the priority of use of the color of each CSF

• when to monitor and/or implement the CSFSTs and FRGs Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.43.5 Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Editorial - removed "if any" from question, reworded A and B.

DCPP L091C Exam Rev 1

Examination Outline Cross-Reference Level SRO Tier # 2 002 A2.02 Ability to (a) predict the impacts of the following Group # 2 malfunctions or operations on the RCS; and (b) based on those K/A # 002 A2.02 predictions, use procedures to correct, control, or mitigate the Rating 4.4 consequences of those malfunctions or operations: Loss of coolant pressure Question SRO 17 (92)

Unit 1 is at 100% power. There is a 150 gpd tube leak in the 1-1 Steam Generator.

When the following indications are observed, the crew initiates Safety Injection:

• Pressurizer level is 41% and decreasing

• Pressurizer Pressure is 2100 psig and decreasing

• All Pressurizer Backup Heaters are ON

• Charging flow is 150 gpm

• Containment Pressure is 1.0 psig and increasing

• Containment sump level is increasing

• Counts on Steamline radiation monitor, RM-71 are stable Which of the following procedure flow paths would be expected from E-0, Reactor Trip or Safety Injection?

A. E-1, Loss of Reactor or Secondary Coolant, ONLY B. E-1, Loss of Reactor or Secondary Coolant and then to E-3, Steam Generator Tube Rupture C. E-3, Steam Generator Tube Rupture and then to ECA-3.1, SGTR With Loss of Reactor Coolant - Subcooled Recovery Desired D. E-3, Steam Generator Tube Rupture and then to ECA-3.2, SGTR With Loss of Reactor Coolant - Saturated Recovery Desired Proposed Answer: A. To E-1, Loss of Reactor or Secondary Coolant, ONLY Explanation:

A. Correct. The tube leakage in the 11 Steam Generator has not increased. There is indications of a RCS leak, E-1 will be entered.

B. Incorrect. No transition to E-3 will be made from E-1. There is no upward trend or spike as required.

C. Incorrect. No transition to E-3 will be made from E-0. There is no upward trend or spike as required. This would be correct if there was also a tube rupture, which there is not.

D. Incorrect. ECA-3.2 is entered from ECA-3.1 Technical

References:

E-0, E-1 References to be provided to applicants during exam: None Learning Objective: 3552 Given initial conditions, assumptions, and symptoms, determine the correct Emergency Operating Procedure to be used to mitigate an operational event Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X DCPP L091C Exam Rev 1

Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.43.5 Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Editorial - all caps ONLY in A DCPP L091C Exam Rev 1

Examination Outline Cross-Reference Level SRO Tier # 2 056 A2.04 Ability to (a) predict the impacts of the following Group # 2 malfunctions or operations on the Condensate System; and (b) K/A # 056 A2.04 based on those predictions, use procedures to correct, control, or Rating 2.8 mitigate the consequences of those malfunctions or operations:

Loss of condensate pumps Question SRO 18 (93)

Unit 1 is at full power. One Condensate Booster pump set is removed from service.

The following events occur:

• PK10-06 CNDS + CNDS BSTR PPS alarms due to an overcurrent trip of one of the running Condensate Booster pumps

• PK10-13 HEATER 2 DRAIN TK LVL HI-LO due to low level

• MFW pump suction pressure is 170 psig and lowering

• PK09-13, Main Feedwater Pump 11 is in alarm due to turbine bearing vibration increasing rapidly to 3.5 mils Which of the following actions should be taken by the Shift Foreman?

The Shift Foreman should go to and direct the actions of:

A. OP AP-15, Loss of Feedwater Flow, Section A, One MFP Trips With Both MFPs Operating, (perform the immediate actions).

B. OP AP-15, Loss of Feedwater Flow, Section C, Heater 2 Drain Pump Trip, (verify a program ramp is in progress).

C. PK09-13, reduce unit load to 550 MW at a ramp rate of 5 MW/Min (and continue to attempt to restore Feedwater flow to normal).

D. OP AP-15, Loss of Feedwater Flow, Section D, Condensate/Booster Pump Set Trip, (trip the reactor and go to E-0, Reactor Trip or Safety Injection).

Proposed Answer: D. Go to and direct the actions of OP AP-15, Loss of Feedwater Flow, Section D, Condensate/Booster Pump Set Trip, (trip the reactor and go to E-0, Reactor Trip or Safety Injection).

Explanation:

A. Incorrect. There is no loss of the main Feedwater pump, the immediate concern is the low suction pressure and loss of the booster pumps.

B. Incorrect. More important to address the loss of the booster pump set.

C. Incorrect. the reason for the trip is the loss of suction pressure, not the loss of the heater drain tank pump or high vibration on the MFP.

D. Correct. The action to take is to address the condensate/booster pump trip and the need to trip the reactor.

Technical

References:

OP AP-15 section D, PK09-13 References to be provided to applicants during exam: None Learning Objective: 537600 - Respond to a loss of feedwater flow DCPP L091C Exam Rev 1

Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.43.5 Reworded C to match format of A, B and D. Moved redundant words to question.

DCPP L091C Exam Rev 1

Examination Outline Cross-Reference Level SRO Tier # 3 G2.1.6 Ability to manage the control room crew during plant Group # 1 transients - prioritize crew actions K/A # G2.1.6 Rating 4.8 Question SRO 19 (94)

The crew is performing the actions of E-0.1, Reactor Trip Response.

An abnormal indication develops involving the Auxiliary Saltwater pumps, meeting the entry conditions for OP AP-10, Loss of Auxiliary Salt Water.

According to OP1.DC10, Conduct of Operations, the Shift Foreman (SFM) should:

A. complete E-0.1 and then enter OP AP-10.

B. suspend E-0.1 and direct the performance of OP AP-10, then continue with E-0.1.

C. continue in E-0.1 and assign a single board operator to implement OP AP-10.

D. concurrently implement E-0.1 and OP AP-10, with the SFM reading and directing both procedures.

Proposed Answer: C. continue in E-0.1 and assign a single board operator to implement OP AP-10.

Explanation:

A. Incorrect. The AP for the loss of saltwater can be addressed but it is of lesser importance and is assigned to an operator per OP1.DC10.

B. Incorrect. The EOP must be completed.

C. Correct. The lesser procedure is addressed by an operator per OP1.DC10.

D. Incorrect. The SFM does not direct both procedures.

Technical

References:

OP1.DC10 step 5.15.3.f.9.g) (page 43)

References to be provided to applicants during exam: None Learning Objective:

Question Source: Bank # NRC-45608 X (note changes; attach parent) Modified Bank #

New Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content: 55.43.5 Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Editorial - added "OP1.DC10" to question, in C. changed "continue reading" to "continue in" DCPP L091C Exam Rev 1

Examination Outline Cross-Reference Level SRO Tier # 3 G2.2.7 Knowledge of the process for conducting special or Group # 2 infrequent tests. K/A # G2.2.7 Rating 3.6 Question SRO 20 (95)

Per OP1.ID4, Conduct of Infrequently Performed Tests or Evolutions, approval to perform an Infrequently Performed Tests or Evolutions (IPTE) is the responsibility of the:

A. Test Director B. Work Shift Control Foreman C. Shift Foreman D. Shift Manager Proposed Answer: D. Shift Manager Explanation:

A. Incorrect. The Test Director has many responsibilities during an IPTE, but cannot approve its performance.

B. Incorrect. Most work goes thru the WCSFM but approval of an IPTE is the responsibility of the SM.

C. Incorrect. The SFM controls work on their respective unit, but approval of IPTE performance is the responsibility of the SM D. Only D is correct. the SM approves the IPTE to be performed.

Technical

References:

OP1.ID4 step 5.7.1 References to be provided to applicants during exam: None Learning Objective: 33440 - Identify the responsibilities associated with various participants in the conduct and oversight of infrequently performed tests and evolutions.

Question Source: Bank # R-63228 X (note changes; attach parent) Modified Bank #

New Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content: 55.43.3 Facility licensee procedures required to obtain authority for design and operating changes in the facility Editorial - added "Per OP1.ID4" to question.

DCPP L091C Exam Rev 1

Examination Outline Cross-Reference Level SRO Tier # 3 G2.2.15 Ability to determine the expected plant configuration Group # 2 using design and configuration control documentation, such as K/A # G2.2.15 drawings, line-ups, tagouts, etc. Rating 4.3 Question SRO 21 (96)

Per OP1.DC20, Sealed Components, whose authorization is REQUIRED to break a Category 1 seal when the associated sealed component checklist is required to be current?

A. The Work Control Lead B. The Work Control Shift Foreman C. The Shift Manager D. The Operations Manager Proposed Answer: B. The Work Control Shift Foreman Explanation:

A. Incorrect. The WCL briefs and coordinates the work, but per OP1.DC20, the appropriate SFM authorization is required.

B. Correct. Per step 5.2.5, seals on Category 1 components may be broken only when authorization by the appropriate Shift Foreman is received. The appropriate SFM is the WCSFM who approves all the work.

C. Incorrect. SFM approval required. Because Category 1 components are components which affect Tech Specs, its conceivable someone above the SFM must approve the seals removal.

D. Incorrect. SFM approval required. Because Category 1 components are components which affect Tech Specs, its conceivable someone above the SFM must approve the seals removal.

Technical

References:

OP1.DC20 References to be provided to applicants during exam: None Learning Objective:

Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content: 55.43.3 Facility licensee procedures required to obtain authority for design and operating changes in the facility.

Editorial - all caps for "required" in question DCPP L091C Exam Rev 1

Examination Outline Cross-Reference Level SRO Tier # 3 G2.3.6 Ability to approve release permits Group # 3 K/A # G2.3.6 Rating 3.8 Question SRO 22 (97)

Which of the following lists the person responsible for preparing and the person responsible for approving a gaseous radwaste discharge permit in accordance with Form 69-21595, Gas Decay Tank Discharge Authorization?

Prepares Discharge Approves Discharge Permit Permit A. Chemistry Shift Foreman B. Chemistry Shift Manager C. Rad Protection Shift Foreman D. Rad Protection Shift Manager Proposed Answer: A. Chemistry/Shift Foreman Explanation:

A. Correct. The permit is prepared by chemistry and approved by the Shift Foreman.

B. Incorrect. The shift manager has overall control of the plant, but the SFM approves work or discharges on their unit.

C. Incorrect. While offsite dose is part of the permit, the calculation is done by chemistry.

D. Incorrect. Prepared by chemistry, approved by the SFM.

Technical

References:

69-21595, OP G-2:V References to be provided to applicants during exam: None Learning Objective: 274964 Review ANY radioactive release related discharge permit.

Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content: 55.43.4 Radiation hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions.

Editorial - added form name and title to question.

DCPP L091C Exam Rev 1

Examination Outline Cross-Reference Level SRO Tier # 3 G2.3.11 Ability to control radiation releases Group # 3 K/A # G2.3.11 Rating 4.3 Question SRO 23 (98)

GIVEN:

• A Gas Decay Tank rupture with an off-site release has occurred

• The EOF and TSC are not activated

• The WCSFM is performing an initial release rate calculation in accordance with EP R-2, Release of Airborne Radioactive Materials Initial Assessment Per EP-R-2, how can the Work Control Shift Foreman obtain reliable data to calculate the release rate calculation?

A. Rad monitor values from SPDS B. Data from the Field Monitoring Team(s)

C. Rad monitor values from the PPC D. Instruct Radiation Protection to conduct surveys of the area around the gas decay tank and at the site boundary Proposed Answer: C. Rad monitor values from the PPC Explanation:

A. Incorrect. EP R-2 states that SPDS is not to be used because the readings may be based on different units of measurements than the procedure uses.

B. Incorrect. FMTs are sent out by the EOF, which is not activated.

C. Correct. Either at the panel or at the PPC location would provide the necessary data.

D. Incorrect. RP is not contacted, or asked to perform surveys.

Technical

References:

EP G-2, EP R-2 and EP EF-3 References to be provided to applicants during exam: None Learning Objective: 42161 Explain when an EP R-2 assessment is required to be performed.

Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content: 55.43.4 Radiation hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions.

Changed D DCPP L091C Exam Rev 1

Examination Outline Cross-Reference Level SRO Tier # 3 G2.4.41 Knowledge of the emergency action level thresholds and Group # 4 classifications. K/A # G2.4.41 Rating 4.6 Question SRO 24 (99)

GIVEN:

• Unit 1 is performing a core offload

• Unit 2 is at full power There is report of the sound of an explosion inside the Protected Area. The Security Watch Commander reports that a Hostile Action has caused failure of the Unit 2 Spent Fuel Cooling systems.

What Emergency Action Level should be declared by the Shift Manager?

A. Unusual event B. Alert C. Site Area Emergency D. General emergency Proposed Answer: C. Site Area Emergency Explanation:

A. Incorrect. SAE is the proper classification B. Incorrect. Correct if the event occurred in the OCA.

C. Correct. Unit 2 is at full power, and the event occurred inside the PA.

D. Incorrect. Could be the proper classification if it had occurred on Unit 1. Note, that the GE is not applicable for unit 2 per the basis of the classification: This EAL encompasses conditions under which a hostile action has resulted in a loss of physical control of vital areas (containing vital equipment or controls of vital equipment) required to maintain safety functions and control of that equipment cannot be transferred to and operated from another location. Typically, these safety functions are reactivity control, RCS inventory, and secondary heat removal. If control of the plant equipment necessary to maintain safety functions can be transferred to another location, then the above initiating condition is not met.

Technical

References:

EP G-1 (H4)

References to be provided to applicants during exam: EAL Chart (G1) - modified to remove Containment Spray information from Containment Barrier EAL (helps with another question)

Learning Objective: 42285 Given indications of an event, use EP G-1 to classify the event with 100% accuracy within 15 minutes Question Source: Bank # R-76664 X (note changes; attach parent) Modified Bank #

New Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental DCPP L091C Exam Rev 1

Comprehensive/Analysis X 10CFR Part 55 Content: 55.43.5 Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Editorial - added "inside Protected Area" to setup.

DCPP L091C Exam Rev 1

Examination Outline Cross-Reference Level SRO Tier # 3 G2.4.45 Ability to prioritize and interpret the significance of each Group # 4 annunciator or alarm. K/A # G2.4.45 Rating 4.3 Question SRO 25 (100)

GIVEN:

• Unit 1 is performing a plant heatup

• RCS temperature is approximately 485°F

• Steam Generator pressures are approximately 580 psig

• RCS pressure is being raised and is currently approximately 1850 psig PK06-06, RCS/SG PRESSURE MISMATCH alarms and clears several times in a minute and is now in alarm.

In accordance with OP1.DC10, Conduct of Operations, what action, if any, should be taken by the Shift Foreman?

A. No action required, this is an expected alarm during the heatup until pressure is above P-11.

B. Direct the operator to remove the "nuisance" alarm from scan.

C. This is an unexpected alarm, the Shift Foreman should enter the alarm response procedure and direct action to raise RCS pressure.

D. This is an unexpected alarm, the Shift Foreman should enter the alarm response procedure and direct action to raise RCS temperature.

Proposed Answer: D. This is an unexpected alarm, the Shift Foreman should enter the alarm response procedure and direct action to raise RCS temperature.

Explanation:

A. Incorrect. This alarm is not expected. If it was expected, this would be a correct action per OP1.DC10.

B. Incorrect. This alarm is not an alarm that is spuriously actuating and must be addressed.

C. Incorrect. Per OP1.DC10, the SFM will respond to unexpected alarms, but the RCS pressure is not the problem, its low steam generator pressure.

D. Correct. Steam generator pressure is less than 600 psig and pressure is above 1850 psig.

The alarms warns of a potential situation of high RCS pressure, above P-11 and low steam generator pressure, which result in low steam pressure SI.

Per OP1.DC10, the action for unexpected alarms are:

• The CO notifies the SFM.

• The SFM concurs with the alarm description.

• The SFM addresses the appropriate annunciator response procedure and updates the crew upon procedure entry.

Note: this alarm is significant, at least one plant has had SI actuation due to these conditions.

Technical

References:

OP1.DC10, PK06-06, steam tables References to be provided to applicants during exam: None Learning Objective:

DCPP L091C Exam Rev 1

Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.43.5 Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Editorial - added "IAW OP1.DC10" to question and changed "will address" to "should reference" in C and D DCPP L091C Exam Rev 1