ML082830327

SO-2008-09-Draft Written Exam

ML082830327
Person / Time
Site:	San Onofre
Issue date:	09/26/2008
From:	Ryan Lantz Operations Branch IV
To:	Rosenblum R Southern California Edison Co
References
50-361/08-301, 50-362/08-301
Download: ML082830327 (394)

Text

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 003 A2.01 Importance Rating 3.5 Reactor Coolant Pump System: Ability to (a) predict the impacts of the following malfunctions or operations on the RCPs; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Problems with RCP seals, especially rates of seal leak off Proposed Question: Common 1 The following annunciators are received in the Control Room:

56C24 - RCP P001 SEAL PRESS HI/LO.

56B57 - RCP BLEEDOFF FLOW HI/LO.

The Reactor Operator reports the following for Reactor Coolant Pump P-001:

• Vapor seal cavity pressure = 64 psia.

• Upper seal cavity pressure = 2015 psia.

• Middle seal cavity pressure = 2238 psia.

Which ONE (1) of the following describes the event in progress and the action required?

A. Middle and Upper seals have failed. Immediately trip the Reactor and initiate SO23-12-1, Standard Post Trip Actions. When Reactivity Control is verified, trip Reactor Coolant Pump P-001.

B. Middle and Lower seals have failed. Initiate a controlled Plant Shutdown per SO23-5-1.7, Power Operations and stop Reactor Coolant Pump P-001 after the Reactor is tripped and CEAs have been inserted for 5 seconds.

C. Middle and Upper seals have failed. Initiate a controlled Plant Shutdown per SO23-5-1.7, Power Operations and stop Reactor Coolant Pump P-001 after the Reactor is tripped and CEAs have been inserted for 5 seconds.

D. Middle and Lower seals have failed. Immediately trip the Reactor and initiate SO23-12-1, Standard Post Trip Actions. When Reactivity Control is verified, trip Reactor Coolant Pump P-001.

Proposed Answer: B Page 1 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible because 2 seals have failed, however, the wrong diagnosis was made. The procedural action is required for three seals failing.

B. Correct. This is the correct seal failure diagnosis and action required per the RNO column of the AOI.

C. Incorrect. Plausible because the procedural response is correct, however, the wrong set of seals is diagnosed.

D. Incorrect. Plausible because this seal failure analysis is correct, however, this action is required only if three seals fail.

Technical Reference(s) SO23-13-6, Step 2 and Attachment 1 Attached w/ Revision # See Comments / Reference Proposed references to be provided during examination: None Learning Objective: Using SO23-13-6, Reactor Coolant Pump Seal Failure, DESCRIBE:

55452 - The basis for each Step, Caution, or Note.

- The expected plant response for each Step.

Question Source: Bank # 126593 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 Page 2 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SO23-13-6, Step 2 Revision # 5 Comments /

Reference:

From SO23-13-6, Attachment 1 Revision # 5 Page 3 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 003 K5.03 Importance Rating 3.1 Reactor Coolant Pump System: Knowledge of the operational implications of the following concepts as the apply to the RCPs: Effects of RCP shutdown on Tave, including the reason for the unreliability of Tave in the shutdown loop Proposed Question: Common 2 When stopping the last Reactor Coolant Pump with Shutdown Cooling in service, which ONE (1) of the following instruments is transitioned to when monitoring Reactor Coolant System cooldown rates?

A. Representative Core Exit Thermocouple because they provide the most accurate cooldown rate when transitioning from Reactor Coolant Pumps to Shutdown Cooling.

B. T-115 / T-125, Reactor Coolant System Loop Tcold because pump coast down will continue to provide an accurate RCS temperature as cooldown progresses.

C. T-351X, Shutdown Cooling Combined Outlet Temperature because this will avoid exceeding cooldown rates when transitioning from Reactor Coolant Pumps to Shutdown Cooling.

D. T-8148 / T-8149, Shutdown Cooling Heat Exchanger Inlet Temperature because adequate mixing has taken place and the cooldown rate can be monitored where turbulent flow exists.

Proposed Answer: C Explanation:

A. Incorrect. Plausible because the CETs may be the only valid Reactor Core temperature indication, however, this occurs when the Unit is in Midloop conditions.

B. Incorrect. Plausible because this is the last operating loop Tcold that is used prior to transitioning to SDC.

C. Correct. Per SO23-5-1.8, Shutdown Operations (MODE 5 & 6) L&S, this is the indication to use during the transition in order to provide an accurate cooldown rate.

D. Incorrect. Plausible because flow is secured through the Heat Exchangers at this time and it could be thought that once flow is reintroduced an accurate indication would exist for cooldown.

Technical Reference(s) SO23-5-1.8, L&S 9.7 & 9.8 Attached w/ Revision # See SO23-5-1.8, L&S 9.10 Comments / Reference SO23-3-2-6, L&S 7.5 Proposed references to be provided during examination: None Page 4 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Learning Objective: DESCRIBE the instrumentation used to monitor the operation of the Shutdown 52765 Cooling System, including the name, function, sensing points, normal values for the parameter being measured, and location of each instrument.

Question Source: Bank # 145431 Modified Bank # (Note changes or attach parent)

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

Reference:

From SO23-5-1.8, L&S 9.7 & 9.8 Revision # 18 Comments /

Reference:

From SO23-5-1.8, L&S 9.10 Revision # 18 Page 5 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SO23-3-2-6, L&S 7.5 Revision # 24 Page 6 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 004 K3.07 Importance Rating 3.8 Chemical and Volume Control System: Knowledge of the effect that a loss or malfunction of the CVCS will have on the following: PZR level and pressure Proposed Question: Common 3 Given the following conditions in MODE 1 at 50% power at MOC:

• TV-0223, Letdown Heat Exchanger Temperature Control Valve, fails closed.

• All control systems are in AUTOMATIC.

• TV-0224B, Ion Exchanger Bypass Valve remains in ION EXCHANGE (does not reposition to BYPASS).

• Pressurizer level is on program.

• Assume NO operator action is taken.

Which ONE (1) of the following describes the effect on Pressurizer level and pressure when TV-0223, Letdown Heat Exchanger Temperature Control Valve fails CLOSED?

Pressurizer level _____________ and Pressurizer pressure ____________.

A. rises; rises.

B. lowers; remains the same.

C. lowers; rises.

D. rises; remains the same.

Proposed Answer: B Explanation:

A. Incorrect. Plausible if thought that boron is removed from the RCS. In this condition level will rise due to temperature increase and pressure will also rise.

B. Correct. Level lowers due to boron sloughing off into the RCS which lowers temperature and lowers PZR level, pressure remains essentially constant.

C. Incorrect. Plausible because level lowers due to boron sloughing off into the RCS which lowers temperature, however, pressure will remain the same.

D. Incorrect. Plausible if thought that boron is removed from the RCS. In this condition level will rise due to temperature increase. Pressure remaining the same is plausible due to the action of the Spray Valves.

Page 7 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Technical Reference(s) SO23-3-2.1, L&S 3.2 Attached w/ Revision # See SD-SO23-390, page 42 Comments / Reference Proposed references to be provided during examination: None Learning Objective: DESCRIBE the operation of the following Letdown and Charging Subsystem 56037 components, including function, location, and specific features including type, capacity, and power supplies where applicable:

- Letdown Temperature Control Valve, TV-0223.

Question Source: Bank #

Modified Bank # 152349 (Note changes or attach parent)

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

Reference:

From SO23-3-2.1, L&S 3.2 Revision # 27 Page 8 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SD-SO23-390, Page 42 Revision # 17 Page 9 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From Exam Bank #152349 Revision 02/04/08 Given the following conditions:

• Unit 3 is in MODE 1 at 50% power at MOC

• 3TV-0223, Letdown Heat Exchanger Temperature Control Valve, fails closed

• All control systems are in AUTOMATIC

• Assume NO operator action

• 3TV-0224B, Ion Exchanger Bypass Valve remains in ION EXCHANGE (does not transfer to BYPASS)

Which one (1) of the following describes the effect on the Pressurizer Level Control System, 30 minutes after 3TV-0223, Letdown Heat Exchanger Temperature Control Valve fails CLOSED?

Actual Pressurizer level _____________ and Pressurizer level setpoint ____________.

A. lowers; lowers B. remains the same; remains the same C. lowers; remains the same D. remains the same; lowers Page 10 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 005A4.02 Importance Rating 3.4 Residual Heat Removal System: Ability to manually operate and/or monitor in the control room: Heat exchanger bypass flow control Proposed Question: Common 4 Which of the following identifies how system flow is controlled while shifting from the Shutdown Cooling (SDC) Heat Exchanger Bypass Flow Valve, HV-8160, to the Shutdown Cooling (SDC) Heat Exchanger Bypass Standby Flow Valve, HV-0396?

A. CLOSE HV-8160, SDC Heat Exchanger Bypass Flow Valve and OPEN HV-0396, SDC Heat Exchanger Standby Flow Valve.

B. Fully OPEN the in-service SDC Heat Exchanger Outlet Valve then CLOSE HV-8160, SDC Heat Exchanger Bypass Flow Valve.

OPEN HV-0396, SDC Heat Exchanger Bypass Standby Flow Valve then THROTTLE SDC Heat Exchanger Outlet Valve to its original position.

C. Verify power to HV-0396, SDC Heat Exchanger Standby Flow Valve then CLOSE the in-service SDC Heat Exchanger Outlet Valve.

CLOSE HV-8160, SDC Heat Exchanger Bypass Flow Valve and then OPEN HV-0396, SDC Heat Exchanger Standby Flow Valve.

OPEN Shutdown Cooling Heat Exchanger Outlet Valve.

D. Verify power to HV-0396, SDC Heat Exchanger Standby Flow Valve then alternately throttle OPEN HV-0396 while throttling HV-8160, SDC Heat Exchanger Bypass Flow Valve CLOSED.

Proposed Answer: D Explanation:

A. Incorrect. Plausible because this method would prevent pump runout as identified in Attachment 6, however, it does not accommodate steady shutdown cooling pump flow.

B. Incorrect. Plausible because this method appears to compensate for the closing of the in-service bypass valve, however, this would result in an uneven cooldown rate.

C. Incorrect. Plausible because the valve is powered up as required, however, this method would result in a cessation of shutdown cooling flow.

D. Correct. Per SO23-3-2.6, this is the correct method to transfer from one valve to the other without causing pump runout.

Technical Reference(s) SO23-3-2.6, Attachment 6 Attached w/ Revision # See Comments / Reference Page 11 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Proposed references to be provided during examination: None Learning Objective: DESCRIBE the flowpaths and major components of the Shutdown Cooling 52692 System used for normal, abnormal, and emergency system operations.

Question Source: Bank # 77876 Modified Bank # (Note changes or attach parent)

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

Reference:

From SO23-3-2.6, Attachment 6 Revision # 25 Page 12 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SO23-3-2.6, Attachment 6 Revision # 25 Page 13 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 006 K4.09 Importance Rating 3.9 Emergency Core Cooling System: Knowledge of the ECCS design features and/or interlocks which provide for the following: Valve positioning on safety injection signal Proposed Question: Common 5 Which ONE (1) of the following describes the signal(s) that position the Safety Injection Minimum Flow Isolation Valves (HV-9306, HV-9307, HV-9347, and HV-9348)?

The Safety Injection Minimum Flow Isolation Valves automatically...

A. OPEN on a SIAS signal.

B. CLOSE on a RAS signal with greater than 5 feet in the Containment Emergency Sump.

C. OPEN on a SIAS signal and CLOSE on a RAS signal.

D. CLOSE on a RAS signal with greater than 4 feet in the Containment Normal Sump.

Proposed Answer: B Explanation:

A. Incorrect. Plausible because the valves are required to be open during a safety injection to provide recirculation flow, therefore, it is reasonable to assume that a SIAS would open these valves.

B. Correct. These valves automatically close on a Recirculation Actuation Signal with 5 feet in the Containment Emergency Sump (which corresponds to a level of 16'6" inside Containment).

C. Incorrect. Plausible because the valves are required to be open during a safety injection to provide recirculation flow, therefore, it is reasonable to assume that a SIAS would open these valves.

Conversely, closing these valves on a RAS would prevent over fill of the RWST.

D. Incorrect. Plausible because 4 feet in the Containment Sump will generate a Containment Sump high level alarm. This corresponds to a level of a 14'6" inside Containment.

Technical Reference(s) SD-SO23-740, Figure 1 Attached w/ Revision # See SD-SO23-740, Page 38 Comments / Reference SO23-15-56.A.56 Proposed references to be provided during examination: None Learning Objective: DESCRIBE the operation of the following Safety Injection and Containment 55522 Spray Systems component, including function, location, and specific features including type, capacity, and power supplies as applicable.

- HPSI Pump Mini-flow and Recirculation lines.

Page 14 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Question Source: Bank #

Modified Bank # (Note changes or attach parent)

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

Reference:

From SD-SO23-740, Figure 1 Revision # 17 Page 15 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SD-SO23-740, Page 38 Revision # 17 Page 16 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SO23-15-56.A.56 Revision # 6 Note located at bottom of page:

Page 17 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 006 G 2.1.27 Importance Rating 3.9 Emergency Core Cooling System: Conduct of Operations: Knowledge of system purpose and/or function Proposed Question: Common 6 Which ONE (1) of the following is the reason for maintaining the Refueling Water Storage Tank 362,800 gallons above the Emergency Core Cooling System suction connection during MODES 1 through 4?

A. Shutdown the Reactor and refill the RCS in the event of an Excess Steam Demand Event.

B. Provide a 20 minute supply for ECCS injection and Containment Spray to ensure Containment pressure will decrease to less than 30 psig.

C. Provide a sufficient level in the Containment Emergency Sump to permit recirculation following a LOCA.

D. Shutdown the Reactor following mixing of the RWST and RCS water volumes with the most reactive CEA not inserted.

Proposed Answer: C Explanation:

A. Incorrect. Plausible because this event will contract the Reactor Coolant System and shutting down the Reactor is a concern, however, this is not the reason for the volume of water maintained.

B. Incorrect. Plausible because providing a 20 minute supply for ECCS injection and Containment Spray is correct, however, the reason is to ensure a 20 minute injection time before swap over to the Containment Emergency Sump.

C. Correct. The reason the RWST is maintained at this volume is to ensure that there is sufficient water in the Containment Emergency Sump to provide adequate net positive suction head once recirculation begins.

D. Incorrect. Plausible because this answer is correct regarding the boron concentration, however, the question asks about the RWST volume.

Technical Reference(s) SD-SO23-740, Pages 24 & 25 Attached w/ Revision # See Unit 2 Tech Spec LCO 3.5.4 Bases Comments / Reference Proposed references to be provided during examination: None Learning Objective: STATE the functions and design bases of the Safety Injection and 81665 Containment Spray System and its components.

Page 18 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Question Source: Bank # 73868 Modified Bank # (Note changes or attach parent)

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

Reference:

From SD-SO23-740, Page 24 Revision # 17 Comments /

Reference:

From SD-SO23-740, Page 25 Revision # 17 Page 19 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From Unit 2 Technical Specification LCO 3.5.4 Bases Amendment #127 Page 20 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 007 K1.03 Importance Rating 3.0 Pressurizer Relief / Quench Tank System: Knowledge of the physical connections and/or cause-effect relationships between the PRTS and the following systems: RCS Proposed Question: Common 7 Given the following conditions:

• T-011, Quench Tank level has slowly risen 1% over the past hour.

• The crew has isolated the Primary Makeup Water supply to the Quench Tank.

Which ONE (1) of the following systems could be leaking into the Pressurizer Quench Tank without attendant alarm annunciation (ignoring Pressurizer Quench Tank alarms)?

A. Pressurizer Spray Valve leakoff.

B. Reactor Coolant Pump Control Bleedoff Relief Valve.

C. Reactor Vessel Flange Seal leakoff.

D. Reactor Head Vent Valves.

Proposed Answer: D Explanation:

A. Incorrect. Plausible because Pressurizer Spray Valve leak off is not annunciated, however, this fluid is sent to the Reactor Coolant Drain Tank vice the Quench Tank.

B. Incorrect. Plausible because the Reactor Coolant Pump Control Bleedoff Relief Valve is aligned to the Quench Tank, however, that line is equipped with a Control Board alarm at 180°F.

C. Incorrect. Plausible because it could be thought that the Reactor Vessel Flange Seal leak off is aligned to the Quench Tank; however, it is aligned to the Reactor Coolant Drain Tank and has an alarm that annunciates in the Control Room.

D. Correct. The Reactor Head Vent Valves are aligned to the Quench Tank. Should these valves leak, they would cause an increase in Quench Tank level without an attendant alarm.

Technical Reference(s) SD-SO23-360, Figure I-1 Attached w/ Revision # See SD-SO23-650, Figure 1 Comments / Reference SO23-15-56.B.59 SO23-15-50.A2.35 Proposed references to be provided during examination: None Page 21 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Learning Objective: IDENTIFY Reactor Coolant System flowpaths, components, and locations 70386 / 56698 including being able to draw and label system diagrams.

DESCRIBE the operation of alarms associated with the Reactor Coolant System, including setpoints, possible causes, and effects on system or overall plant operation.

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 3, 7 55.43 Comments /

Reference:

From SD-SO23-360, Figure I-1 Revision # 17 Page 22 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SD-SO23-360, Figure I-1 Revision # 17 Comments /

Reference:

From SO23-15-56.B.59 Revision # 6 Comments /

Reference:

From SO23-15-50.A2.35 Revision # 11 Page 23 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SD-SO23-650, Figure 1 Revision # 12 Page 24 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 008 K2.02 Importance Rating 3.0 Component Cooling Water System: Knowledge of bus power supplies to the following: CCW pump, including emergency backup Proposed Question: Common 8 Given the following conditions:

• Component Cooling Water Pumps P-024 and P-026 are operating.

• Component Cooling Water Pump P-025 is aligned to Train A in Standby.

• A Safety Injection Actuation Signal (SIAS) and a Loss of Voltage Signal (LOVS) on both 1E 4 kV buses for greater than five seconds has been received.

Which ONE (1) of the following correctly describes the Component Cooling Water Pump responses when the buses are reenergized?

A. P-025 and P-026 start, P-024 does not start.

B. P-024 and P-026 start, P-025 does not start.

C. All CCW Pumps start.

D. P-024 and P-025 start, P-026 does not start.

Proposed Answer: A Explanation:

A. Correct. With P-025 pump in Standby on Train A, P-025 and P-026 pumps will both start with the conditions listed. This is because the third-of-a-kind pump when aligned to either Train, has starting priority in this condition.

B. Incorrect. Plausible because it could be thought that the pumps that were initially aligned would start, however, the circuitry starts the third-of-a-kind pump with the conditions listed.

C. Incorrect. Plausible because since the standby pump is aligned it could be thought that all Component Cooling Water Pumps would start on a Safety Injection Actuation Signal.

D. Incorrect. Plausible if thought that only two pumps would start and since the standby pump is aligned, that would be the selected pump.

Technical Reference(s) SO23-2-17, Attachment 9, L&S 2.12 Attached w/ Revision # See SD-SO23-400, Page 14 Comments / Reference Proposed references to be provided during examination: None Page 25 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Learning Objective: DESCRIBE the operation of the following Component Cooling Water System 56251 components, including function, location, and specific features which include type, capacity, power supplies, and normal operating parameters where applicable:

- CCW Pumps - P024, P025, & P026.

Question Source: Bank # 75550 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam 2000 NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Comments /

Reference:

From SO23-2-17, Attachment 9, L&S 2.12 Revision # 26 Comments /

Reference:

From SD-SO23-400, Page 14 Revision # 18 Page 26 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 010 A1.04 Importance Rating 3.6 Pressurizer Pressure Control System: Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the PZR PCS controls including: Effect of temperature change during solid operation Proposed Question: Common 9 Given the following conditions:

• The Pressurizer is solid and all Pressurizer Heaters are energized for drawing a bubble.

• PIC-100, Pressurizer Pressure Controller is in MANUAL with output set at 0%.

• Both Letdown Backpressure Control Valves are in MANUAL and 10% open.

• Pressurizer temperature is 350°F and rising at 1°F/minute.

• Pressurizer pressure is 350 psia.

Which ONE (1) of the following describes the plant response as Pressurizer temperature rises?

As Pressurizer temperature rises, Pressurizer pressure rises until...

A. PSV-9206, Letdown Backpressure Relief Valve opens at its setpoint of 406 psig and then holds RCS pressure constant.

B. a bubble is formed, at which time the Pressurizer pressure stabilizes.

C. the Proportional Pressurizer Heaters deenergize at 2275 psia.

D. all Pressurizer Heaters will be deenergized at 2340 psia.

Proposed Answer: D Page 27 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible because this valve will open, however, the increase in RCS pressure will be greater than the relief valve capacity. Additionally, this is the relief setpoint of the LTOP Valves not the backpressure relief valve (650 psig).

B. Incorrect. Plausible because conditions are set for a pressurizer bubble being formed, however, with the Letdown Back Pressure Control Valves in MANUAL, as Pressurizer temperature rises Pressurizer pressure will also rise and the valves do not open sufficiently in response to this pressure rise.

C. Incorrect. Plausible because with the conditions listed, pressure will continue to rise, however, only the Backup Heaters will deenergize at this pressure. The Proportional Heaters will not deenergize until 2340 psia.

D. Correct. With the conditions listed, pressure will continue to rise until all heaters deenergize at 2340 psia. This is because the Letdown Back Pressure Control Valves will not open as RCS pressure rises and bubble formation will not occur until PZR temp is 432°F.

Technical Reference(s) SO23-3-1.4, Attachment 4, Step 2.5 Attached w/ Revision # See SO23-3-1.4, Attachment 16, L&S 12.4 Comments / Reference SO23-3-1.10, Attachment 9, L&S 2.3 SO23-3-1.10, Attachment 8 Proposed references to be provided during examination: None Learning Objective: DESCRIBE the operation of Pressurizer Pressure Control System 56417 components, instrumentation, controls and alarms including function, location, interlocks, capacity and power supplies where applicable.

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

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

Reference:

From SO23-SO23-3-1.10, Attachment 9, L&S 2.3 Revision # 18 Comments /

Reference:

From SO23-3-1.4, Attachment 16, L&S 12.4 Revision # 30 Page 28 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SO23-3-1.10, Attachment 8 Revision # 18 Page 29 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SO23-3-1.4, Attachment 4, Step 2.5 Revision # 30 Page 30 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 010 A2.02 Importance Rating 3.9 Pressurizer Pressure Control System: Ability to (a) predict the impacts of the following malfunctions or operations on the PZR PCS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Spray valve failures Proposed Question: Common 10 Given the following Unit 2 conditions at 30% power:

• Power ascension is in progress via dilution and forcing of Pressurizer Spray.

• PV-0100A and PV-0100B, Pressurizer Spray Valves are each 20% open.

• The Main Turbine is Ramp Loading at 3 MWe / minute.

30 minutes into the power ascension the following conditions are noted:

• Pressurizer pressure is lowering about 1 psia / second.

• Annunciator 50A14 - PZR PRESS HI / LO is in alarm.

• PV-0100A, Pressurizer Spray Valve is 80% open.

• PV-0100B, Pressurizer Spray Valve is closed.

Which ONE (1) of the following correctly describes the event that has occurred, and the required actions?

A. PV-0100B, Pressurizer Spray Valve has failed closed.

Refer to SO23-13-27, Pressurizer Pressure and Level Control Malfunctions.

Place HIC-0100B, Spray Valve Controller in MANUAL and open PV-0100B.

B. PV-0100A, Pressurizer Spray Valve has failed open.

Refer to SO23-13-27, Pressurizer Pressure and Level Control Malfunctions.

Place HIC-0100A, Spray Valve Controller in MANUAL and close PV-0100A.

C. PIC-0100, Pressurizer Pressure Controller has failed high.

When Pressurizer pressure reaches 2025 psia and decreasing, trip the Reactor and enter SO23-12-1, Standard Post Trip Actions.

D. PG-3, Main Turbine ramp rate has increased.

Secure the dilution per SO23-5-1.7, Power Operations.

Energize additional Pressurizer heaters and depress Turbine Control PG3 HOLD pushbutton.

Proposed Answer: B Page 31 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible because the procedure and associated actions are correct if the valve had in fact failed closed; however, if PV-0100B fails closed then PV-0100A would only be open approximately 40%. Pressure would also remain controlled.

B. Correct. Given the conditions listed and the fact that Pressurizer pressure is decreasing along with the position of PV-0100A (it would only be at 40% open if PV-0100B had in fact failed closed), this valve has failed open.

C. Incorrect. Plausible because with the controller failing high Pressurizer pressure would be decreasing, however, one would expect both Spray Valves to be open in this condition. A Reactor trip would be necessary if Pressurizer pressure were not controlled.

D. Incorrect. Plausible because if ramp rate had increased then Pressurizer pressure will decrease, however, given the conditions listed a Spray Valve has failed open.

Technical Reference(s) SO23-13-27, Step 3 Attached w/ Revision # See SO23-15-50.A1.14 Comments / Reference Proposed references to be provided during examination: None Learning Objective: DESCRIBE the cause/effect relationships associated with the Pressurizer 56419 Pressure and Level Control System and an increasing or decreasing Pressurizer pressure and/or level.

Question Source: Bank # 127472 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 5, 7 55.43 Page 32 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SO23-13-27, Step 3 Revision # 3 Page 33 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SO23-15-50.A1.14 Revision # 7 Page 34 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 012 K6.09 Importance Rating 3.6 Reactor Protection System: Knowledge of the effect of a loss or malfunction of the following will have on the RPS: CEAC Proposed Question: Common 11 The Unit is operating at 100% power with the following conditions:

• Control Element Assembly Calculator #2 is failed.

• Core Protection Calculator Point ID 062 (CEANOP) has been set to "2" in all Core Protection Calculator Channels.

If Control Element Assembly Calculator #1 fails and no additional operator action is taken, which ONE (1) of the following identifies how the Core Protection Calculators will respond?

A. All 4 Core Protection Calculator Channels will trip after a 90 minute time delay.

B. No channel will trip since the Core Protection Calculators will use the last available valid set of Control Element Assembly position information in the determination of any penalty factors.

C. All 4 Core Protection Calculator Channels will trip after a 30 second time delay.

D. No channel will trip since having the CEANOP flag set at a value of "2" already indicates both Control Element Assembly Calculators are INOPERABLE.

Proposed Answer: A Explanation:

A. Correct. Given the conditions listed and no operator action, a trip will occur after a 90 minute time delay.

B. Incorrect. Plausible because a snapshot of a CEA position is stored within the Core Protection Calculator, however, this is a misinterpretation of how the CPC works since a penalty factor is generated.

C. Incorrect. Plausible because some penalty factors are induced after 20 seconds, however, not for the conditions listed in the stem.

D. Incorrect. Plausible because it could be thought that this condition is correct, however, the CEANOP flag does not function this way.

Technical Reference(s) SO23-3-2.13, Attachment 8, L&S 1.10 Attached w/ Revision # See SO23-3-2.13, Precaution 4.6 Comments / Reference SO23-3-2.13, Attachment 8, L&S 2.8 Page 35 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Proposed references to be provided during examination: None Learning Objective: Given a change to one or more CPC input parameters, PREDICT the CPC 52413 response to the changing conditions.

Question Source: Bank # 129838 Modified Bank # (Note changes or attach parent)

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

Reference:

From SO23-3-2.13, Attachment 8, L&S 1.10 Revision # 14-1 Comments /

Reference:

From SO23-3-2.13, Precaution 4.6 Revision # 14-1 Comments /

Reference:

From SO23-3-2.13, Attachment 8, L&S 2.8.1 Revision # 14-1 Page 36 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 013 K4.02 Importance Rating 3.7 Engineered Safety Features Actuation System: Knowledge of the ESFAS design features and/or interlocks which provide for the following:

Safety injection block Proposed Question: Common 12 Which ONE (1) of the following Engineered Safety Feature Actuation System (ESFAS) inputs or signals is bypassed to permit plant cooldown to Shutdown Cooling entry conditions?

A. Narrow Range Steam Generator Level - High B. Reactor Coolant Pump Speed - Low C. Wide Range Pressurizer Pressure - Low D. Wide Range Steam Generator Level - Low Proposed Answer: C Explanation:

A. Incorrect. Plausible because filling of the Steam Generator to cool it down is normally accomplished during a plant cooldown.

B. Incorrect. Plausible because the outputs from the Core Protection Calculator are bypassed (low DNBR, high LPD), however, the RCP speed input is not.

C. Correct. Pressurizer Pressure low can be bypassed, procedurally at ~375 psia.

D. Incorrect. Plausible because low Steam Generator pressure can be reset, however, not SG level.

Technical Reference(s) SO23-5-1.5, Step 6.7.19 Attached w/ Revision # See SD-SO23-710, Figure 1 Comments / Reference Proposed references to be provided during examination: None Learning Objective: DESCRIBE the operation of the Plant Protection System components and 56627 instrumentation, including function, location, design basis, interlocks, setpoints, special features and power supplies, where applicable.

Page 37 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Question Source: Bank # 74264 Modified Bank # (Note changes or attach parent)

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

Reference:

From SO23-5-1.5, Step 6.7.19 Revision # 28 Comments /

Reference:

From SD-SO23-710, Figure 1 Revision # 7 Page 38 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SD-SO23-710, Figure 1 Revision # 7 Page 39 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SD-SO23-710, Figure 1 Revision # 7 Page 40 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 022 G 2.4.45 Importance Rating 4.1 Containment Cooling System: Emergency Procedures / Plan: Ability to prioritize and interpret the significance of each annunciator or alarm Proposed Question: Common 13 Given the following conditions:

• The crew has manually initiated a Safety Injection Actuation Signal due to decreasing Pressurizer pressure and level.

• Pressurizer pressure has stabilized at and not decreased below 1800 psia.

• The following annunciators are in alarm:

• 57A01 - SIAS TRAIN A ACTUATION

• 57B01 - SIAS TRAIN B ACTUATION

• The following annunciators are not in alarm:

• 57A07 - CCAS TRAIN A ACTUATION

• 57B07 - CCAS TRAIN B ACTUATION Which ONE (1) of the following identifies how the Containment Emergency Cooling Units respond during this event?

The Containment Emergency Cooling Units will...

A. automatically start if a manual Containment Cooling Actuation Signal is initiated.

B. only start by manually aligning Component Cooling Water and manually starting fans.

C. automatically start when the manual SIAS is actuated on both Trains.

D. NOT automatically start if a Containment Cooling Actuation Signal is later received.

Proposed Answer: A Page 41 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Explanation:

A. Correct. As stated in the reference material, the instrumentation that auto initiates an SIAS also generates a CCAS. When a manual SIAS is initiated and neither PZR nor Containment pressures reach their respective auto setpoints (1740 psia and 3.4 psig) then the CCAS will not occur. The CCAS components will auto start if CCAS is manually actuated.

B. Incorrect. Plausible because it could be thought that without the automatic actuation signal present the operator would be required to manually align components.

C. Incorrect. Plausible because it could be thought that the Containment ECUs are already operating given that an SIAS was manually actuated.

D. Incorrect. Plausible because it could be thought that without an initial auto actuation signal present that any subsequent signals would not initiate a CCAS.

Technical Reference(s) SD-SO23-720, Pages 15 & 16 Attached w/ Revision # See SD-SO23-720, Figures 2A & 2B Comments / Reference Proposed references to be provided during examination: None Learning Objective: EXPLAIN the interfaces between the Containment Air Handling System and 81637 / 81638 other plant systems.

DESCRIBE the configuration and operational characteristics of Containment Air Handling System components.

Question Source: Bank # 75577 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam 2000 NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7, 8 55.43 Comments /

Reference:

From SD-SO23-720, Page 15 Revision # 8 Page 42 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SD-SO23-720, Page 16 Revision # 8 Page 43 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SD-SO23-720, Figure 2B Revision # 8 Page 44 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SD-SO23-720, Figure 2A Revision # 8 Page 45 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 022 A3.01 Importance Rating 4.1 Containment Cooling System: Ability to monitor automatic operation of the CCS, including: Initiation of safeguards mode of operation Proposed Question: Common 14 Which ONE (1) of the following parameters will automatically initiate a Containment Cooling Actuation Signal (CCAS)?

A. Containment Wide Range Pressure or Pressurizer Narrow Range Pressure.

B. Containment Narrow Range Pressure or Pressurizer Wide Range Pressure.

C. Containment Wide Range Pressure or Pressurizer Wide Range Pressure.

D. Containment Narrow Range Pressure or Pressurizer Narrow Range Pressure.

Proposed Answer: B Explanation:

A. Incorrect. Plausible because Containment Wide Range Pressure is used as an input for a Containment Spray actuation.

B. Correct. These are the initiating devices for a Containment Cooling signal.

C. Incorrect. Plausible because Containment Wide Range Pressure is used as an input for a Containment Spray actuation and Pressurizer Wide Range Pressure is correct.

D. Incorrect. Plausible because Containment Wide Range Pressure is used, however, Pressurizer Narrow Range Pressure is not.

Technical Reference(s) SD-SO23-720, Figures 2B & 2C Attached w/ Revision # See SD-SO23-720, Pages 14 & 17 Comments / Reference Proposed references to be provided during examination: None Learning Objective: DESCRIBE the inputs to the Plant Protection System, the purpose of each, 56628 their trip setpoints and actuation logic.

Question Source: Bank # 73634 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam 2000 NRC Exam Page 46 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 Page 47 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SD-SO23-720 Figure 2B Revision # 8 Page 48 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SD-SO23-720 Figure 2C Revision # 8 Page 49 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SD-SO23-720, Page 14 Revision # 8 Page 50 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SD-SO23-720, Page 17 Revision # 8 Page 51 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 026 K3.01 Importance Rating 3.9 Containment Spray System: Knowledge of the effect that a loss or malfunction of the CSS will have on the following: CCS Proposed Question: Common 15 Given the following conditions during a Large Break Loss of Coolant Accident:

• SIAS, CCAS, CIAS, and CSAS have all actuated.

• Both Containment Spray Pump Discharge Valves have failed to open.

Which ONE (1) of the following identifies the impact on Containment Cooling?

Containment pressure and temperature design bases will not be exceeded as long as a minimum of...

A. two (2) Containment Emergency Cooling Units (ECUs) operating with both ECUs aligned to Component Cooling Water.

B. four (4) Containment Emergency Cooling Units (ECUs) operating with at least two (2) ECUs aligned to Component Cooling Water and two (2) Containment Dome Air Circulator Fans in operation.

C. two (2) Containment Emergency Cooling Units (ECUs) operating with both ECUs aligned to Component Cooling Water and three (3) Containment Dome Air Circulator Fans in operation.

D. four (4) Containment Emergency Cooling Units (ECUs) operating and four (4)

ECUs aligned to Component Cooling Water.

Proposed Answer: D Page 52 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible because with both trains of Containment Spray (CS) unavailable it could be thought that only one train of Containment Emergency Cooling would be required, however, this would only provide 50% of the required cooling.

B. Incorrect. Plausible because with both trains of CS unavailable it could be thought that both trains of Containment ECUs and one (1) train of Containment Dome Air Circulator Fans would be required, however, an ECU must have CCW aligned to be considered OPERABLE.

C. Incorrect. Plausible because with both trains of CS unavailable it could be thought that one train of Containment ECUs and Containment Dome Air Circulator Fans would be required in addition to the one (1) Containment Dome Air Circulator Fan that is always running during normal operation.

D. Correct. Per the Design Basis, two trains of Containment Spray or two trains of Containment Emergency Cooling will maintain Containment pressure and temperature design bases.

Technical Reference(s) Unit 2 Tech Spec 3.6.6.1 Bases Attached w/ Revision # See Comments / Reference Proposed references to be provided during examination: None Learning Objective: STATE the functions and design bases of the Containment Air Handling 81634 System and its components.

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

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

Reference:

From Unit 2 Tech Spec 3.6.6.1 Bases Amendment # 127 Page 53 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From Unit 2 Tech Spec 3.6.6.1 Bases Amendment # 127 Comments /

Reference:

From Unit 2 Tech Spec 3.6.6.1 Bases Amendment # 128 Comments /

Reference:

From Unit 2 Tech Spec 3.6.6.1 Bases Amendment # 128 Page 54 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From Unit 2 Tech Spec 3.6.6.1 Bases Amendment # 128 Page 55 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 039 K4.08 Importance Rating 3.3 Main and Reheat Steam System: Knowledge of MRSS design features and/or interlocks which provide for the following: Interlocks on MSIV and bypass valves Proposed Question: Common 16 Following a Containment Isolation Actuation Signal (CIAS), which ONE (1) of the following describes the MINIMUM action required to open a Main Steam Isolation Valve?

A. Reset the CIAS and depress 1 of 2 valve OPEN pushbuttons.

B. Reset the CIAS and depress 2 of 2 valve OPEN pushbuttons.

C. Place the valve in OVERRIDE and depress 1 of 2 OPEN pushbuttons.

D. Place the valve in OVERRIDE and depress 2 of 2 OPEN pushbuttons.

Proposed Answer: B Explanation:

A. Incorrect. Plausible because this action would work if the valve were to be closed (two solenoids deenergized), however, opening requires all four (4) solenoids to be energized.

B. Correct. The CIAS must be reset and all four (4) solenoids to be energized by depressing 2 of 2 valve OPEN pushbuttons.

C. Incorrect. Plausible because this action would work if the valve were to be closed (two solenoids deenergized), however, opening requires all four (4) solenoids to be energized and the OVERRIDE pushbutton is only available for the MSIV Bypass Valve (see Reference).

D. Incorrect. Plausible because all four (4) solenoids to be energized by depressing 2 of 2 valve and OPEN pushbuttons, however, the OVERRIDE pushbutton is only available for the MSIV Bypass Valve (see Reference).

Technical Reference(s) SD-SO23-160, Pages 29 & 30 Attached w/ Revision # See SO23-3-2.22, Attachments 7 & 11 Comments / Reference Proposed references to be provided during examination: None Learning Objective: As the RO, VERIFY alignment of Main Steam System Valves following Main 53902 Steam Isolation System Actuation per SO23-3-2.22.

Question Source: Bank # 127497 Modified Bank # (Note changes or attach parent)

New Page 56 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7, 10 55.43 Comments /

Reference:

From SD-SO23-160, Page 29 Revision # 20 Comments /

Reference:

From SD-SO23-160, Page 30 Revision # 20 Comments /

Reference:

From SO23-3-2.22, Attachment 7 Revision # 15-5 Note at bottom of page:

Page 57 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SO23-3-2.22, Attachment 7 Revision # 15-5 Note at bottom of page:

Page 58 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SO23-3-2.22, Attachment 11 Revision # 15-5 Note at bottom of page:

Page 59 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 039 A4.07 Importance Rating 2.8 Main and Reheat Steam System: Ability to manually operate and/or monitor in the control room: Steam dump valves Proposed Question: Common 17 Given the following conditions following a Unit 2 trip from 100% power:

• Containment Pressure has risen to 11 psig.

• RCS pressure is at 1350 psia and slowly lowering.

• Steam Generator E-088 level is at 45% wide range and lowering.

• Steam Generator E-088 pressure has lowered to 300 psia.

• Steam Generator E-089 level is at 27% narrow range and rising.

• Steam Generator E-089 pressure is at 980 psia and stable.

Given the above conditions, which ONE (1) of the following actions must be performed to open Steam Generator E-089 HV-8421, Atmospheric Dump Valve?

Establish an HV-8421, Atmospheric Dump Valve setpoint approximately 200 psia above Steam Generator E-088 pressure and depress A. OVERRIDE then AUTO.

B. CLOSE then OPEN/MODULATE then AUTO.

C. OVERRIDE then OPEN/MODULATE then AUTO.

D. AUTO then OPEN/ MODULATE.

Proposed Answer: C Explanation:

A. Incorrect. Plausible because OVERRIDE must be depressed, however this action will not open the valve. The AUTO / MANUAL pushbutton is a controller function (as is the demand signal).

B. Incorrect. Plausible because without the MSIS the valve would open.

C. Correct. This is the action necessary to open the ADV with an MSIS signal present.

D. Incorrect. Plausible because without the MSIS the valve would open. This is how the ADV is placed in service following a trip when the Steam Bypass Control System is not available.

Page 60 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Technical Reference(s) SD-SO23-160, Page 21 Attached w/ Revision # See SO23-3-2.22, Attachment 11 Comments / Reference Lesson Plan 2XPR02, Slide #31 Proposed references to be provided during examination: None Learning Objective: As the RO, OPERATE the Atmospheric Steam Dump Valves from the Control 55067 Room per SO23-3-2.18.1.

Question Source: Bank # 127370 Modified Bank # (Note changes or attach parent)

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

Reference:

From SD-SO23-160, Page 21 Revision # 20 Page 61 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SD-SO23-160, Page 24 Revision # 20 Comments /

Reference:

From SO23-3-2.22, Attachment 11 Revision # 15-5 Note at bottom of page:

Page 62 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From Lesson Plan 2XPR02, Slide #31 Revision # 4 Page 63 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 059 A4.12 Importance Rating 3.4 Main Feedwater System: Ability to manually operate and/or monitor in the control room: Initiation of automatic feedwater isolation Proposed Question: Common 18 Which ONE (1) of the following is the INITIAL response of the Digital Feedwater Control System to a High Level Override signal from Steam Generator E-089 ONLY during operation at 100% power?

Main Feedwater Pump Speed will...

A. NOT be affected.

Regulating Valve will go fully closed.

Bypass Valve will go fully closed.

B. go to minimum.

Regulating Valve will go fully closed.

Bypass Valve will close to 50%.

C. go to minimum.

Regulating Valve will go fully closed.

Bypass Valve will go fully closed.

D. NOT be affected.

Regulating Valve will go fully closed.

Bypass Valve will close to 50%.

Proposed Answer: A Explanation:

A. Correct. Given the conditions in the stem, the High Level Override (HLO) does not affect Main Feed Pump speed (it is a high select signal) but will close both the Main Feed Regulating and Main Feed Bypass valves.

B. Incorrect. Plausible because these actions occur with a Reactor Trip Override. It could be thought that a Reactor trip occurred, however, the HLO occurs at 85% level and the RTO occurs coincident with a Reactor trip at 89% level.

C. Incorrect. Plausible because the Main Feed Regulating and Bypass Valve positions are correct, however, Main Feed Pump speed going to minimum is associated with a Reactor Trip Override.

D. Incorrect. Plausible because Main Feed Pump Speed will not be affected, however, the other actions are associated with a Reactor Trip Override. It could be thought that a Reactor trip occurred, however, the HLO occurs at 85% level and the RTO occurs coincident with a Reactor trip at 89% level.

Page 64 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Technical Reference(s) SO23-9-6, Precaution 4.7 Attached w/ Revision # See SD-SO23-250, Pages 75 & 76 Comments / Reference SO23-15-52.A.16 Proposed references to be provided during examination: None Learning Objective: INTERPRET instrumentation and controls utilized in the Main Feedwater Pump 64707 / 64708 and Turbine System.

ANALYZE normal and abnormal operations of the Main Feedwater Pump and Turbine System.

Question Source: Bank # 110369 Modified Bank # (Note changes or attach parent)

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

Reference:

From SO23-9-6, Precaution 4.7 Revision # 18 Page 65 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SD-SO23-250, Page 75 Revision # 15 Page 66 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SD-SO23-250, Page 76 Revision # 15 Page 67 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SO23-15-52.A.16 Revision # 9 Page 68 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 059 G 2.4.31 Importance Rating 4.2 Main Feedwater System: Emergency Procedures / Plan: Knowledge of annunciator alarms, indications, or response procedures Proposed Question: Common 19 Given the following conditions:

• Unit 2 has tripped with Annunciator 53A03 - MFWP TURBINE K006 TRIP in alarm.

Which ONE (1) of the following parameters would cause the Main Feedwater Pump trip?

A. MFW Pump P-062 suction pressure @ 180 psig.

B. MFWP Turbine K-006 vibration @ 3 mils.

C. MFW Pump P-062 thrust bearing wear @ 12 mils.

D. MFWP Turbine K-006 thrust bearing wear @ 20 mils.

Proposed Answer: A Explanation:

A. Correct. The MFWP has tripped on low NPSH at 220 psig.

B. Incorrect. Plausible because K-006 vibration PRETRIP alarms at 3 mils, however, the trip is at 5 mils.

C. Incorrect. Plausible because P-062 vibration does alarm at 12 mils, however, there is no associated trip.

D. Incorrect. Plausible because K-006 thrust bearing wear PRETRIP alarms at 20 mils, however, the trip is at 25 mils.

Technical Reference(s) SO23-15-53.A.03, 08, 13, 15, & 25 Attached w/ Revision # See Comments / Reference Proposed references to be provided during examination: None Learning Objective: DESCRIBE the instrumentation used to monitor the operations of the 53419 Feedwater Pump and Turbine System, including the name, function, sensing points, and location of each instrument.

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Page 69 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7, 10 55.43 Comments /

Reference:

From SO23-15-53.A.03 Revision # 13 Page 70 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SO23-15-53.A.08 Revision # 13 Comments /

Reference:

From SO23-15-53.A.13 Revision # 13 Page 71 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SO23-15-53.A.15 Revision # 13 Comments /

Reference:

From SO23-15-53.A.25 Revision # 13 Page 72 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 061 A3.02 Importance Rating 4.0 Auxiliary/Emergency Feedwater System: Ability to monitor automatic operation of the AFW, including: RCS cooldown during AFW operations Proposed Question: Common 20 Which ONE (1) of the following describes the limitations on the Auxiliary Feedwater Pump Discharge Valves (HV-4705, 4706, 4712 and 4713) while feeding with the Auxiliary Feedwater system during a plant cooldown?

The Auxiliary Feedwater Pump Discharge Valves should not be throttled...

A. less than 26% open at any time.

B. less than 10% open at any time. If an Emergency Feedwater Actuation Signal is present then the valves should not be throttled less than 26% open.

C. less than 10% open at any time. If a low Steam Generator pressure condition is present then the valves should not be throttled less than 26% open.

D. to greater than 10% open with a low Steam Generator pressure condition.

Proposed Answer: C Explanation:

A. Incorrect. Plausible because the 26% open limit is correct, however, it is only when a low Steam Generator pressure condition exists.

B. Incorrect. Plausible because the valve positioning information is correct, however, it is during a low Steam Generator pressure condition vice an Emergency Feedwater Actuation Signal.

C. Correct. Per the Limitations and Specifics identified in the Auxiliary Feedwater procedure, these are the valve position limits.

D. Incorrect. Plausible because a low Steam Generator pressure condition is the concern, however, the valve throttle limit is incorrect.

Technical Reference(s) SO23-2-4, L&S 2.1 Attached w/ Revision # See Comments / Reference Proposed references to be provided during examination: None Learning Objective: Given an operational condition of the Auxiliary Feedwater System addressed 55812 by a procedural precaution, limitation, or administrative requirement, STATE the limiting condition and the basis for that limiting condition.

Page 73 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Question Source: Bank # 75612 Modified Bank # (Note changes or attach parent)

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

Reference:

From SO23-2-4, L&S 2.1 Revision # 23 Page 74 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 064 K1.05 Importance Rating 4.1 Emergency Diesel Generator System: Knowledge of the effect that a loss or malfunction of the AC distribution system will have on the following: EDG Proposed Question: Common 21 Given the following conditions on Unit 2:

Buses 2A04, 2A06, 3A04, and 3A06 have low voltage alarms.

All four (4) 1E Bus Voltages are approximately 4090 volts.

Low voltage alarms have been in for 2 minutes.

No SIAS actuation is present on either Unit.

All other equipment is either OPERABLE or in AUTO.

Which ONE (1) of the following identifies how the 1E Degraded Voltage Protection Circuit responds on Unit 2?

A. Emergency Diesel Generators remain off; Unit 3 energizes Buses 2A04 and 2A06 via Bus Tie Breakers.

B. Emergency Diesel Generators start; Unit 3 energizes Buses 2A04 and 2A06 via Bus Tie Breakers.

C. Emergency Diesel Generators remain off; Unit 2 continues to supply Buses 2A04 and 2A06.

D. Emergency Diesel Generators start; 2G002 energizes Bus 2A04 and 2G003 energizes Bus 2A06.

Proposed Answer: D Explanation:

Incorrect. Plausible because this is the Preferred Alternate Power Source, however, after 110 seconds the EDGs will start.

Incorrect. Plausible because this is the Preferred Alternate Power Source, however, with low voltage on Unit 3 the EDGs will close onto the buses.

Incorrect. Plausible because this condition existed up until the point 110 seconds was reached. Once two minutes had elapsed the SDVS circuit picked up which started and loaded the EDGs.

Correct. This is the 1E Degraded Voltage Protection Circuit response to a Sustained Degraded Voltage Signal (SDVS). If the Unit 3 voltages were normal the Unit 2 buses would have transferred to Unit 3 via the Reserve Auxiliary Transformers (as it is the Preferred Alternate Power Source).

Page 75 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Technical Reference(s) SO23-6-2, L&S 4.0 Attached w/ Revision # See SO2-15-63.B.05 Comments / Reference SD-SO23-120, Page 107 Proposed references to be provided during examination: None Learning Objective: DESCRIBE the integrated operation of SDVS (Sustained Degraded Voltage 52899 Signal) and DGVSS (Degraded Grid Voltage with SIAS Signal) including sensor inputs, actuation logic, and actuated devices.

Question Source: Bank # 112931 Modified Bank # (Note changes or attach parent)

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

Reference:

From SO23-6-2, L&S 4.0 Revision # 13 Page 76 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SO2-15-63.B.05 Revision # 12 Page 77 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SD-SO23-120, Page 107 Revision # 19 Page 78 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 063 A3.01 Importance Rating 2.7 DC Electrical Distribution System: Ability to monitor automatic operation of the DC electrical system, including: Meters, annunciators, dials, recorders, and indicating lights Proposed Question: Common 22 Given the following conditions of Unit 2 DC Bus voltage:

• DC Bus 2D1 is at 132 volts and -2 amps.

• DC Bus 2D2 is at 130 volts and -1 amp.

• DC Bus 2D3 is at 131 volts and -3 amps.

• DC Bus 2D4 is at 132 volts and -2 amps.

After a plant transient the following is observed:

• DC Bus 2D1 is at 131 volts and -2 amps.

• DC Bus 2D2 is at 119 volts and +186 amps.

• DC Bus 2D3 is at 130 volts and -1 amp.

• DC Bus 2D4 is at 123 volts and +127 amps.

Which ONE (1) of the following events has caused the change in DC Bus voltage?

A loss of...

A. Motor Control Center BQ has occurred.

B. Bus 2A06 has occurred.

C. Motor Control Center BS has occurred D. Bus 2A04 has occurred.

Proposed Answer: B Page 79 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible because MCC BQ powers the recently installed Swing Battery Charger, however, this Battery Charger can only be aligned to D1 or D3.

B. Correct. The initial conditions in the stem signify the typical DC bus voltages and amperages. The negative indications associated with Bus amperage are reflective of the trickle charge placed on the batteries. A loss of 4160 VAC Bus 2A06 causes loss of the D2 and D4 Battery Chargers. The resultant draw on the buses is reflected in the final amperage readings. Readings provided were validated in the Simulator pre- and post-transient.

C. Incorrect. Plausible because MCC BS powers the recently installed Swing Battery Charger for Buses D2 and D4, however, this Battery Charger can only be aligned to one bus at a time.

D. Incorrect. Plausible if thought that these buses were powered from Battery Chargers associated with this bus.

Technical Reference(s) SD-SO23-130, Figure 1 Attached w/ Revision # See Comments / Reference Proposed references to be provided during examination: None Learning Objective: ANALYZE normal and abnormal operations of the 1E 125 VDC and 120 VAC 80607 Electrical System.

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 5, 7, 8 55.43 Page 80 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SD-SO23-130, Figure 1 Revision # 15 Page 81 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SD-SO23-130, Figure 1 Revision # 15 Page 82 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 064 K1.05 Importance Rating 3.4 Emergency Diesel Generator System: Knowledge of the physical connections and/or cause-effect relationships between the EDG system and the following systems: Starting air system Proposed Question: Common 23 Unit 2 is operating normally at 100% power and the Unit 2 Emergency Diesel Generator Air Compressor, C-012A, control switch is in the AUTO position.

Which ONE (1) of the following parameters will directly cause an automatic protective trip of the running Compressor?

A. High oil level.

B. High vibration.

C. High oil pressure.

D. High air discharge temperature.

Proposed Answer: D Explanation:

A. Incorrect. Plausible because there is a compressor trip associated with low oil level, however, not high oil level.

B. Incorrect. Plausible because there is a high vibration alarm on the Local Control Panel, however, it is associated with the Emergency Diesel Generator.

C. Incorrect. Plausible because there is a high oil pressure alarm on the Local Control Panel, however, this alarm is associated with the fuel oil strainer.

D. Correct. A high air discharge temperature will cause a trip of the air compressor.

Technical Reference(s) SO23-5-2.35.1, Alarm 1-6-1 Attached w/ Revision # See SO23-5-2.35.1, Alarm 2-2-1 Comments / Reference SO23-5-2.35.1, Alarm 2-1-5 Proposed references to be provided during examination: None Learning Objective: DESCRIBE the operations of the controls for the following Emergency Diesel 53463 Generator Mechanical Systems including name, function, interlocks and location for each:

- Emergency Diesel Generator Air Start System Page 83 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Question Source: Bank # 141718 Modified Bank # (Note changes or attach parent)

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

Reference:

From SO23-5-2.35.1 (Alarm 1-6-1) Revision # 9 Page 84 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SO23-5-2.35.1 (Alarm 1-6-1) Revision # 9 Comments /

Reference:

From SO23-5-2.35.1 (Alarm 1-6-1) Revision # 9 Page 85 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SO23-5-2.35.1 (Alarm 2-1-5) Revision # 9 Comments /

Reference:

From SO23-5-2.35.1 (Alarm 1-1-5) Revision # 9 Page 86 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 073 A1.01 Importance Rating 3.2 Process Radiation Monitoring System: Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the PRM system controls including: Radiation levels Proposed Question: Common 24 Which ONE (1) of the following is the effect of a high radiation as sensed by RE-7865, Containment Purge and Plant Vent Stack Wide Range Radiation Monitor, when it is aligned to the Plant Vent Stack?

A radiation signal that exceeds the setpoint to RE-7865 would send a...

A. CLOSE signal to the Outside Containment Purge Isolation Valves.

B. CLOSE signal to the Containment Purge and Mini-Purge Supply and Exhaust valves.

C. CLOSE signal to FV-7202, Waste Gas Discharge Flow Control Valve.

D. TRIP signal to the Continuous Exhaust Fans A-310, A-311 and A-312.

Proposed Answer: C Explanation:

A. Incorrect. Plausible because with this rad monitor aligned to the Containment Purge Stack it will isolate these valves on a high radiation signal; however, it is aligned to the Plant Vent Stack.

B. Incorrect. Plausible because this rad monitor will isolate the outside Containment Purge Valves, however, it does not input into the Containment Purge Isolation Signal circuitry.

C. Correct. When RE-7865 is aligned to the Primary Vent Stack a high radiation signal will isolate FV-7202, Waste Gas Discharge Flow Control Valve.

D. Incorrect. Plausible because this action is related to the closure of FV-7202, Waste Gas Discharge valve, however, it is the loss of the Continuous Exhaust fans that closes FV-7202.

Technical Reference(s) SO23-8-15, L&S 1.3 & 4.5 Attached w/ Revision # See SD-SO23-690, Page 9 Comments / Reference Proposed references to be provided during examination: None Learning Objective: EXPLAIN the interfaces between the Radiation Monitoring System and other 103328 plant systems.

Page 87 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Question Source: Bank # 73822 Modified Bank # (Note changes or attach parent)

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

Reference:

From SO23-8-15, L&S 4.5 Revision # 17 Comments /

Reference:

From SO23-8-15, L&S 1.3 Revision # 17 Comments /

Reference:

From SD-SO23-690, Page 9 Revision # 16 Page 88 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 076 K2.08 Importance Rating 3.1 Service Water System: Knowledge of bus power supplies to the following: ESF-actuated MOVs Proposed Question: Common 25 Which ONE (1) of the following describes the power supply for the Component Cooling Water Heat Exchanger Saltwater Cooling Normal Outlet Valves (HV-6495 and HV-6497)?

A. 7 Turbine Building MCC BM.

B. 50 Control Building MCCs BS and BQ.

C. 7 Turbine Building MCC BK.

D. 50 Control Building MCCs BY and BZ.

Proposed Answer: D Explanation:

A. Incorrect. Plausible because the mnemonic for MCC BM is by the mountains which is the closest MCC to valves HV-6495 and HV-6497.

B. Incorrect. Plausible because both of these MCCs located on the 50 Control Building, however, they are Non-1E powered.

C. Incorrect. Plausible because the mnemonic for MCC BK is by the intake which is the next closest MCC to valves HV-6495 and HV-6497 (after MCC BM).

D. Correct. HV-6495 and HV-6497 are powered from a 1E MCCs BY and BZ.

Technical Reference(s) SD-SO23-410, Page 18 Attached w/ Revision # See Comments / Reference Proposed references to be provided during examination: None Learning Objective: EXPLAIN the interfaces between the Salt Water Cooling System and other 60304 plant systems.

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Page 89 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Comprehension or Analysis 10 CFR Part 55 Content: 55.41 4, 7 55.43 Comments /

Reference:

From SD-SO23-410, Page 18 Revision # 7 Page 90 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 076 K4.06 Importance Rating 2.8 Service Water System: Knowledge of SWS design features and/or interlocks which provide for the following: Service water train separation Proposed Question: Common 26 Which ONE (1) of the following design features provides for Saltwater Cooling System train separation?

A. One (1) Unit 2 Train A and one (1) Unit 2 Train B Saltwater Cooling Pump is located in the Unit 2 SWC Pump Room.

One (1) Unit 2 Train A and one (1) Unit 2 Train B Saltwater Cooling Pump is located in the Unit 3 SWC Pump Room.

B. Two (2) Unit 2 Train A Saltwater Cooling Pumps are located in the Unit 2 SWC Pump Room.

Two (2) Unit 2 Train B Saltwater Cooling Pumps are located in the Unit 3 SWC Pump Room.

C. Two (2) Unit 3 Train A Saltwater Cooling Pumps are located in the Unit 3 SWC Pump Room.

Two (2) Unit 3 Train B Saltwater Cooling Pumps are located in the Unit 2 SWC Pump Room.

D. One (1) Unit 3 Train A and two (2) Unit 2 Train A Saltwater Cooling Pumps are located in the Unit 2 SWC Pump Room.

Two (2) Unit 3 Train B and one (1) Unit 2 Train B Saltwater Cooling Pumps are located in the Unit 3 SWC Pump Room.

Proposed Answer: A Explanation:

A. Correct. Refer to the attached drawing from SD-SO23-410, Figure 1.

B. Incorrect. Plausible because there are two (2) Train A Pumps in the Unit 2 SWC Pump Room, however, they provide Train A for each Unit. The 2nd statement is also plausible because there are two (2) Train B Pumps in Unit 3 SWC Pump Room; however, they provide Train B for each Unit.

C. Incorrect. Plausible because there are two (2) Train A Pumps in the Unit 3 SWC Pump Room, however, they provide Train A for each Unit. The 2nd statement is also plausible because there are two (2) Train B Pumps in Unit 2 SWC Pump Room; however, they provide Train B for each Unit.

D. Incorrect. Plausible because there are four (4) SWC Pumps in each Units Pump Room. This configuration accounts for each of the four SWC Pumps while implying that one (1) Unit 3 Train A Pump is in the Unit 3 SWC Pump Room and that one (1) Unit 2 Train B Pump is in the Unit 2 SWC Pump Room.

Page 91 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Technical Reference(s) SD-SO23-410, Figure 1 Attached w/ Revision # See Comments / Reference Proposed references to be provided during examination: None Learning Objective: IDENTIFY Salt Water Cooling System flowpaths, components, and locations 60303 including being able to draw and label system diagrams.

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7, 8 55.43 Page 92 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SD-SO23-410, Figure 1 Revision # 7 Page 93 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 078 K1.03 Importance Rating 3.3 Instrument Air System: Knowledge of the physical connections and/or cause-effect relationships between the IAS and the following systems:

Containment air Proposed Question: Common 27 Which ONE (1) of the following isolates and restores Instrument Air to Containment during high usage conditions?

HV-5343, Reactor Instrument Air Excess Flow Check Valve closes on high flow of greater than or equal to...

A. 200 cfm and AUTO resets when flow drops to 0 cfm.

B. 25 cfm and AUTO resets when flow drops to 0 cfm.

C. 200 cfm and can be reset manually when flow returns to normal.

D. 25 cfm and can be reset manually when flow returns to normal.

Proposed Answer: C Explanation:

A. Incorrect. Plausible because the Excess Flow Check Valve isolates at 200 cfm, however, it must be manually reset by the operator.

B. Incorrect. Plausible because the backup nitrogen system is equipped with a check valve and alarm at 25 cfm, however, this would only occur when a backup nitrogen valve opened.

C. Correct. The Excess Flow Check Valve isolates at 200 cfm and can be reset manually at CR-57 by depressing an OPEN pushbutton.

D. Incorrect. Plausible because the backup nitrogen system is equipped with a check valve and alarm at 25 cfm, however, this check valve internals are removed.

Technical Reference(s) SD-SO23-570, Pages 6, 7, 16, & 18 Attached w/ Revision # See SD-SO23-570, Figure I-1 Comments / Reference Proposed references to be provided during examination: None Learning Objective: DESCRIBE the operation of the controls associated with the Instrument Air 53980 System, including the name, function, interlocks, and location of each.

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Page 94 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7, 9 55.43 Comments /

Reference:

From SD-SO23-570, Pages 6 & 7 Revision # 16 Page 95 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SD-SO23-570, Page 16 Revision # 16 Page 96 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SD-SO23-570, Page 18 Revision # 16 Page 97 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SD-SO23-570, Figure I-1 Revision # 16 Comments /

Reference:

From SD-SO23-570, Figure I-1 Revision # 16 Page 98 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SD-SO23-570, Figure I-1 Revision # 16 Page 99 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 103 A2.05 Importance Rating 2.9 Containment System: Ability to (a) predict the impacts of the following malfunctions or operations on the containment system; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Emergency containment entry Proposed Question: Common 28 Given the following conditions:

• Containment pressure is 3.2 psig.

• Containment average temperature is 125°F.

• Containment nitrogen concentration is 82%.

• Containment oxygen concentration is 17%.

Which ONE (1) of the following:

1.) Identifies the impact on performing an Emergency Containment entry?

2.) What action should be taken to mitigate the situation?

A. 1.) An Emergency Containment entry can be performed.

2.) Reduce Containment temperature by placing Containment Emergency Cooling Units in service per SO23-1-4.1, Containment Emergency Cooling.

B. 1.) An Emergency Containment entry cannot be performed.

2.) Reduce Containment pressure to less than 3 psig per SO23-1-4.2, Containment Purge and Recirculation Filtration System, Attachment 6, Operation of the Containment Mini-Purge System.

C. 1.) An Emergency Containment entry can be performed.

2.) Don a Self Contained Breathing Apparatus per SO23-3-2-34, Containment Access Control with oxygen level less than or equal to 19.5%.

D. 1.) An Emergency Containment entry cannot be performed.

2.) Reduce Containment nitrogen concentration to 80% by purging one volume from Containment per SO23-3-2-34, Containment Access Control.

Proposed Answer: B Page 100 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible because average Containment temperature is greater than 105°F which requires implementing Containment Emergency Cooling; however, Containment pressure is too high for entry at this time. Tech Spec LCO 3.6.5 requires temperature to be 120°F in MODES 1 to 4.

B. Correct. With Containment pressure greater than 3 psig, Containment entry is not allowed. Placing the Containment Mini-Purge in operation is required because a direct of venting of Containment is only allowed if pressure is less than 1 psig.

C. Incorrect. Plausible because donning self-contained breathing apparatus is required at this oxygen concentration, however, Containment pressure is too high for an emergency entry to be performed.

D. Incorrect. Plausible because the emergency entry cannot be performed and this could be the action for high nitrogen, however, there is no specific concentration value for nitrogen (only oxygen) and Containment pressure is still too high.

Technical Reference(s) SO23-3-2.34, Precaution 4.2 Attached w/ Revision # See SO23-3-2.34, Section 6.8 Comments / Reference SO23-3-2.34, L&S 1.5 SO23-1-4.1, Section 6.1 SO23-1-4.2, Attachments 6 & 11 Proposed references to be provided during examination: None Learning Objective: ANALYZE normal and abnormal operations of the Containment Air Handling 81640 System.

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

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

Reference:

From SO23-3-2.34, Precaution 4.2 Revision # 21 Page 101 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SO23-3-2.34, Section 6.8 Revision # 21 Comments /

Reference:

From SO23-3-2.34, L&S 1.5 Revision # 21 Page 102 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SO23-1-4.1, Section 6.1 Revision # 13 Page 103 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SO23-1-4.2, Attachment 6 Revision # 28 Comments /

Reference:

From SO23-1-4.2, Attachment 11 Revision # 28 Page 104 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 2 K/A # 001 A4.08 Importance Rating 3.7 Control Rod Drive System: Ability to manually operate and/or monitor in the control room: Mode select for CRDS; operation of rod control M/G sets and control panel Proposed Question: Common 29 Which ONE (1) of the following is the procedural switch alignment required to withdraw a single Control Element Assembly (CEA)?

A. Group Select Switch selected to any Group.

Mode Select Switch in any Mode except OFF.

Individual CEA Select Switch is aligned to the appropriate CEA.

B. Group Select Switch is selected to appropriate Group.

Mode Select Switch is in Manual Individual Mode.

Individual CEA Select Switch may be aligned to any CEA.

C. Group Select Switch is selected to the appropriate Group.

Mode Select Switch is in Manual Individual Mode.

Individual CEA Select Switch is aligned to the appropriate CEA.

D. Group Select Switch selected to any Group.

Mode Select Switch in any Mode except OFF.

Individual CEA Select Switch is aligned to the appropriate CEA.

Proposed Answer: C Explanation:

A. Incorrect. Plausible because the Individual CEA Select Switch is aligned to the appropriate CEA, however, the Group Select and Mode Select Switches must be properly aligned.

B. Incorrect. Plausible because all requirements are met, however, the Individual CEA Select Switch must be selected by procedure.

C. Correct. All three requirements must be met to withdraw a single CEA.

D. Incorrect. Plausible because it could be thought that as long as the Individual CEA Select Switch is aligned to the appropriate CEA and the Mode Select Switch is not in OFF, the CEA can be moved.

Technical Reference(s) SO23-3-2.19, Section 6.3 Attached w/ Revision # See Comments / Reference Proposed references to be provided during examination: None Page 105 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Learning Objective: DESCRIBE the configuration and operational characteristics of Control 81786 Element Drive Mechanism Control System (CEDMCS) components.

Question Source: Bank # 75157 Modified Bank # (Note changes or attach parent)

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

Reference:

From SO23-3-2.19, Section 6.3 Revision # 20 Page 106 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 2 K/A # 002 A2.04 Importance Rating 4.3 Reactor Coolant System: Ability to (a) predict the impacts of the following malfunctions or operations on the RCS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Loss of heat sinks Proposed Question: Common 30 Given the following conditions during MODE 5:

• Reactor Coolant System level as read on Reactor Water Level Indicator / Digital Level Monitoring System LI-1520N has lowered from 40 inches (40) to 26 inches (26).

• The running Shutdown Cooling Pump has tripped on overcurrent.

• Annunciator 50A42 - DLMS HI / LO SDCS / MIDLOOP TROUBLE has alarmed.

Which ONE (1) of the following:

1.) Identifies the impact on the operational condition of the Reactor Coolant System?

2.) What action should be taken to mitigate the situation?

A. 1.) Reactor Coolant System transitions from Midloop Operations to a Reduced Inventory Condition.

2.) Monitor Reactor Coolant temperature using the Core Exit Thermocouples until the system is refilled and the standby SDC Pump is placed in service.

B. 1.) Reactor Coolant System remains in Midloop Operations.

2.) Monitor Reactor Coolant temperature using TR-0351A (T351X), SDC Combined Outlet temperature until the system is refilled and the standby SDC Pump is placed in service.

C. 1.) Reactor Coolant System transitions from Midloop Operations to a Reduced Inventory Condition.

2.) Monitor Reactor Coolant temperature using TR-0351A (T351X), SDC Combined Outlet temperature until the system is refilled and the standby SDC Pump is placed in service.

D. 1.) Reactor Coolant System remains in Midloop Operations.

2.) Monitor Reactor Coolant temperature using the Core Exit Thermocouples until the system is refilled and the standby SDC Pump is placed in service.

Proposed Answer: D Page 107 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible because the Core Exit Thermocouples provide an accurate indication of core temperature in this condition, however, the unit is in a Reduced Inventory Condition before it enters a Midloop condition which is a standard misconception based on the terminology used.

B. Incorrect. Plausible because the Reactor Coolant System is in the Midloop condition, however, with a loss of Shutdown Cooling flow only the Core Exit Thermocouples provide an accurate indication of core temperature.

C. Incorrect. Plausible because this is the temperature indication used when there are no Reactor Coolant Pumps operating, however, with a loss of Shutdown Cooling flow only the Core Exit Thermocouples provide an accurate indication of core temperature. Additionally, the unit is in a Reduced Inventory Condition before it enters a Midloop condition which is a standard misconception based on the terminology used.

D. Correct. The unit is already in a Reduced Inventory Condition as soon as water level is below the Reactor Vessel Flange; therefore, having level lower from 40 to 26 inches maintains the Reactor Coolant System in the Midloop condition. Per the guidance in the Loss of Shutdown Cooling procedure, only the Core Exit Thermocouples provide an accurate indication of core temperature.

Technical Reference(s) SO23-5-1.8, L&S 3.11 & 3.12 Attached w/ Revision # See SO23-3-2.6, Attachment 10 Comments / Reference SO23-13-15, Entry Conditions Proposed references to be provided during examination: SO23-3-2.6, Attachment 10 Learning Objective: Given plant and equipment conditions during Reduced Inventory Conditions 54686 and applicable procedures, EVALUATE plant status, DETERMINE the required actions and EXPLAIN the basis for these actions.

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 5, 10 55.43 Page 108 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SO23-5-1.8, L&S 3.11 & 3.12 Revision # 18 Comments /

Reference:

From SO23-3-2.6, Attachment 10 Revision # 24 Page 109 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SO23-13-15, Entry Conditions Revision # 18 Page 110 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 2 K/A # 014 K1.02 Importance Rating 3.0 Rod Position Indication System: Knowledge of the physical connections and/or cause-effect relationships between the RPIS and the following systems: NIS Proposed Question: Common 31 Which ONE (1) of the following describes the relationship between the Rod Position Indication System and the Nuclear Instrument System?

A. The Control Element Assembly Reed Switch Position Indication System works in concert with the Excore Nuclear Instruments to generate a Core Protection Calculator signal used for Departure from Nuclear Boiling and Local Power Density signals.

B. The Control Element Assembly Pulse Counting Position Indication System works in concert with the Excore Nuclear Instruments to generate a Control Element Assembly Withdrawal Prohibit (CWP) signal.

C. The Control Element Assembly Reed Switch Position Indication System works in concert with the Incore Nuclear Instruments to generate a Core Protection Calculator signal used for Departure from Nuclear Boiling and Local Power Density signals.

D. The Control Element Assembly Pulse Counting Position Indication System works in concert with the Incore Nuclear Instruments to generate a Control Element Assembly Withdrawal Prohibit (CWP) signal.

Proposed Answer: A Explanation:

A. Correct. The Reed Switch Position Indication System and the Excore Nuclear instruments provide the signals listed. The Core Protection Calculator also receives input from Pressurizer pressure, Reactor Coolant System temperature and the Reactor Coolant Pump speed to generate core protection signals.

B. Incorrect. Plausible because the Excore Instruments are part of the instrumentation that generates the CEA Withdrawal Prohibit signal, however, it is the Reed Switch not the Pulse Counting Indication System that is used.

C. Incorrect. Plausible because all parts of this statement are true with the exception of the Incore instruments. The Incore Instruments are used as part of the Core Operating Limits Supervisory System (COLSS).

D. Incorrect. Plausible because the Incore Instruments are used as part of the Core Operating Limits Supervisory System (COLSS), however, the CWP is generated via the Excore Instruments and the Reed Switch Position Indication System.

Page 111 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Technical Reference(s) SD-SO23-710, Pages 23, 24, 45, 47, 53, 57 Attached w/ Revision # See SD-SO23-710, Figure 1 Comments / Reference Proposed references to be provided during examination: None Learning Objective: DESCRIBE the purpose, location and operation of the various operator 54673 controls associated with the CPCs and CEACs.

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

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

Reference:

From SD-SO23-710, Page 23 Revision # 7 Page 112 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SD-SO23-710, Page 24 Revision # 7 Page 113 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SD-SO23-710, Page 45 Revision # 7 Page 114 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SD-SO23-710, Page 47 Revision # 7 Comments /

Reference:

From SD-SO23-710, Page 53 Revision # 7 Page 115 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SD-SO23-710, Page 57 Revision # 7 Comments /

Reference:

From SD-SO23-710, Figure 1 Revision # 7 Page 116 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 2 K/A # 027 G 2.1.23 Importance Rating 4.3 Containment Iodine Removal System: Conduct of Operations: Ability to perform specific system and integrated plant procedures during all modes of plant operation Proposed Question: Common 32 Given the following conditions:

• Health Physics has requested a Containment Air Cleanup.

• The Control Room Supervisor directs you to perform SO23-1-4.2, Containment Purge and Recirculation Filtration System, Attachment 7, Containment Recirculation Filtration System Operation.

Which ONE (1) of the following describes the operation of the Unit 2 Containment Recirculation Filtration Unit A-353?

The Unit 2 Containment Recirculation Filtration Unit A-353 can...

A. be manually started from 2L-155, Heating, Ventilation, and Air Conditioning System Control.

B. automatically start or be manually started from 2/3 CR-60, Emergency Heating, Ventilation, and Air Conditioning Panel.

C. automatically start or be manually started from 2L-155, Heating, Ventilation, and Air Conditioning System Control.

D. be manually started from 2/3 CR-60, Emergency Heating, Ventilation, and Air Conditioning Panel.

Proposed Answer: A Explanation:

A. Correct. A-353 is located on the Containment 45 level and operated from 2L-155, HVAC.

B. Incorrect. Plausible because CR-60 contains a temperature recorder that is used to monitor operation of A-353 heaters, however, the unit is started at 2L-155. Also, its associated heater unit can operate automatically.

C. Incorrect. Plausible because the location is correct and it could be thought that an ESFAS signal will start the Filtration Unit. Also, its associated heater unit can operate automatically.

D. Incorrect. Plausible because the unit can only be manually started, however, this location is only used for temperature monitoring, not for A-353 start.

Page 117 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Technical Reference(s) SO23-1-4.2, Attachment 7 Attached w/ Revision # See SD-SO23-770, Pages 52 & 53 Comments / Reference Proposed references to be provided during examination: None Learning Objective: DESCRIBE the configuration and operational characteristics of Containment 81638 Air Handling System components.

Question Source: Bank # 151671 Modified Bank # (Note changes or attach parent)

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

Reference:

From SO23- 1-4.2, Attachment 7 Revision # 28 Page 118 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SD-SO23-770, Page 52 Revision # 8 Page 119 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SD-SO23-770, Page 53 Revision # 8 Page 120 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 2 K/A # 041 K6.03 Importance Rating 2.7 Steam Dump/Turbine Bypass Control System: Knowledge of the effect of a loss or malfunction of the following will have on the SDS:

Controller and positioners, including ICS, SG, CRDS Proposed Question: Common 33 Given the following conditions with Unit 2 at 100% power at End-of-Cycle:

• Pressurizer pressure is 2250 psia.

• Tcold is 539°F.

• The selected Main Steam pressure instrument input into the Steam Bypass Control System fails HIGH.

• The Reactor subsequently tripped.

Assuming NO action by the crew, which ONE (1) of the following is the response of the Atmospheric Dump Valves and Steam Bypass Control Valves over the next five (5) minutes?

Atmospheric Dump Valves...

A. OPEN and Steam Bypass Control Valves OPEN.

B. remain CLOSED and Steam Bypass Control Valves OPEN.

C. OPEN and Steam Bypass Control Valves CLOSE.

D. remain CLOSED and Steam Bypass Control Valves remain CLOSED.

Proposed Answer: B Page 121 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible because it could be thought that the Atmospheric Dump Valves (ADVs) used the higher pressure setpoint, however, it is the lower of two inputs.

B. Correct. When a steam pressure instrument fails high, it activates the permissive signal for the Steam Bypass Control Valves. The ADVs do not respond because although they also have two pressure inputs, they use pressure transmitters from the Steam Generator as opposed to the Main Steam System. Additionally, the ADVs use the lower of the two pressure signals to prevent a failure such as this from opening the ADVs and they are not aligned.

C. Incorrect. Plausible because this is the reverse of what actually would occur. One could arrive at this conclusion if thought that the ADVs selected the higher of two pressure inputs and the SBCS selected the lower.

D. Incorrect. Plausible because the ADVs will remain closed. Confusion over the Permissive Signal versus the Modulation Signal could yield this answer for SBCS.

Technical Reference(s) Lesson Plan 2XIR05, Slides 28, 29, & 72 Attached w/ Revision # See SD-SO23-160, Page 21 & 22 Comments / Reference Proposed references to be provided during examination: None Learning Objective: DESCRIBE the operation of the Steam Bypass Control System controls.

54350 / 102466 INTERPRET instrumentation and controls utilized in the Main Steam System.

Question Source: Bank # 126731 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam 2003 NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Page 122 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From Lesson Plan 2XIR05, Slide #28 Revision # 3 Comments /

Reference:

From Lesson Plan 2XIR05, Slide #29 Revision # 3 Page 123 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From Lesson Plan 2XIR05, Slide #72 Revision # 3 Comments /

Reference:

From SD-SO23-160, Page 21 & 22 Revision # 20 Page 124 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 2 K/A # 029 A1.03 Importance Rating 3.0 Containment Purge System: Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating Containment Purge System controls including: Containment pressure, temperature, and humidity Proposed Question: Common 34 Given the following conditions while in MODE 5:

• Containment pressure is greater than atmospheric pressure.

• The Containment Equipment Hatch is closed.

• The Containment Main Purge System is to be placed in service.

Which ONE (1) of the following identifies the effect that Containment pressure has on the Containment Main Purge System?

When Containment is greater than atmospheric pressure, equalize pressure to prevent...

A. damage to the Main Purge Supply and Exhaust Valve Actuators.

B. excessive differential pressure across and damage to the Containment Normal Purge Supply and Exhaust Unit particulate filters.

C. damage to the Mini-Purge Supply and Exhaust Valve Actuators.

D. failure of the Main Purge Supply and Exhaust ductwork due to exposure to higher than design pressure.

Proposed Answer: A Explanation:

A. Correct. Per the Limitations and Specifics, equalize pressure prior to opening the Main Purge Supply and Exhaust Valves to prevent damage to the valve actuators.

B. Incorrect. Plausible because if valves were opened out of sequence this could occur, however, it is the valve actuators that must be protected.

C. Incorrect. Plausible if thought that a high differential pressure would damage these valves, however, these valves are designed to open with a high P.

D. Incorrect. Plausible if thought that this could occur, however, it is the valve actuators that must be protected.

Page 125 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Technical Reference(s) SO23-1-4.2, L&S 1.5 Attached w/ Revision # See SO23-1-4.2, Attachment 5 Comments / Reference SD-SO23-770, Figure 1 Proposed references to be provided during examination: None Learning Objective: DESCRIBE the configuration and operational characteristics of Containment 81446 System components.

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

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

Reference:

From SO23-1-4.2, L&S 1.5 Revision # 28 Comments /

Reference:

From SD-SO23-770, Figure 1 Revision # 8 Page 126 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SO23-1-4.2, Attachment 5 Revision # 28 Page 127 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 2 K/A # 033 G 2.4.49 Importance Rating 4.6 Spent Fuel Cooling System: Emergency Procedures / Plan: Ability to perform without reference to procedures those actions that require immediate operation of system components and controls Proposed Question: Common 35 Unit 2 is in MODE 1 when the following events occur:

• Spent Fuel Pool Cooling Pump P-010 trips.

• SO23-13-23, Loss of Spent Fuel Pool Cooling, is initiated.

• Spent Fuel Pool Cooling Pump P-009 is started.

• Spent Fuel Pool temperature is slowly rising.

• Annunciator 61C07 - SPENT FUEL POOL PUMP P009 PRESS LO, is in alarm.

• The Radwaste Operator reports P-009 discharge pressure indicates 18 psig, and no system leakage is apparent.

Which ONE (1) of the following actions is required?

A. Attempt to vent the Spent Fuel Pool Cooling Pump casing.

B. Throttle OPEN on the Spent Fuel Pool Cooling Pump Discharge Valve.

C. Throttle CLOSED on the Spent Fuel Pool Cooling Pump Suction Valve.

D. Throttle CLOSED on the Spent Fuel Pool Cooling Pump Discharge Valve.

Proposed Answer: D Explanation:

A. Incorrect. Plausible because this action could assist if the pump were cavitating, however, a Note states this action is unnecessary due to system design and a Caution alerts personnel to the potential danger of an uncontrolled SFP level loss.

B. Incorrect. Plausible because the Spent Fuel Pool Cooling Pump Discharge Valve is throttled as a matter of system operation, however, in this condition the valve needs to be throttled closed.

C. Incorrect. Plausible because venting of the suction piping may be necessary if cavitation is detected (as outlined in the RNO actions of Step 4), however, the suction valve would not be throttled closed if the pump were running.

D. Correct. With Spent Fuel Pool Cooling Pump discharge pressure less than 20 psig, the crew directs the local operator to throttle the pump discharge valve until pressure is approximately 30 psig.

Page 128 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Technical Reference(s) SO23-13-23, Steps 3 & 4 Attached w/ Revision # See SO2-15-61.C.07 Comments / Reference Proposed references to be provided during examination: None Learning Objective: ANALYZE normal and abnormal operations of the Fuel Storage and Spent 81752 Fuel Pool Cooling System.

Question Source: Bank # 132035 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 5, 10 55.43 Page 129 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SO23-13-23, Step 3 Revision # 10 Page 130 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SO23-13-23, Step 4 Revision # 10 Page 131 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SO2-15-61.C.07 Revision # 6 Page 132 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 2 K/A # 045 A1.06 Importance Rating 3.3 Main Turbine Generator System: Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating MTG System controls including: Expected response of secondary plant parameters following TG trip Proposed Question: Common 36 Which ONE (1) of the following is the normal response of the Main Generator Automatic Voltage Regulator (AVR) during a normal Main Turbine trip and coast down?

A. The AVR Channels remain in their pre-trip configuration. As Turbine Speed drops below 1800 rpm, the Generator Excitation is reduced to a minimum. Below 1440 rpm, excitation is removed by tripping both AVR AC Supply (Input) Breakers (29VA/29VB).

B. Both AVR Channels trip to EMERGENCY MANUAL Mode. "Suppression of Excitation" is initiated and Excitation is immediately reduced to minimum. Below 1440 rpm, excitation is removed by inhibiting Thyristor firing.

C. The AVR Channels remain in their pre-trip configuration. As Turbine speed drops below 1800 rpm, the Generator Voltage is reduced proportionally to speed. Below 1440 rpm, excitation is removed by inhibiting Thyristor firing.

D. Both AVR Channels trip to EMERGENCY MANUAL Mode. "Suppression of Excitation" is initiated and excitation is immediately removed by tripping both AVR AC Supply (Input) Breakers (29VA/29VB).

Proposed Answer: C Explanation:

A. Incorrect. Plausible because all other indications are correct with the exception of the AVR AC Input Breakers opening.

B. Incorrect. Plausible because it could be thought that the AVR senses a condition that requires a trip to EMERGENCY MANUAL C. Correct. This is the action of the AVR post-trip.

D. Incorrect. Plausible because an overvoltage condition when the Unit is not synchronized will trip the AVR AC Supply Breakers.

Technical Reference(s) SD-SO23-105, Pages 4, 18, 20, 31, & 32 Attached w/ Revision # See Comments / Reference Proposed references to be provided during examination: None Page 133 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Learning Objective: ANALYZE normal and abnormal operations of the Turbine Protection and 83809 Auxiliaries Systems.

Question Source: Bank # 78049 Modified Bank # (Note changes or attach parent)

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

Reference:

From SD-SO23-105, Page 31 Revision # 12 Page 134 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SD-SO23-105, Page 20 Revision # 12 Comments /

Reference:

From SD-SO23-105, Page 4 Revision # 12 Page 135 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SD-SO23-105, Page 32 Revision # 12 Comments /

Reference:

From SD-SO23-105, Page 18 Revision # 12 Page 136 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 2 K/A # 055 K3.01 Importance Rating 2.5 Condenser Air Removal System: Knowledge of the effect that a loss or malfunction of the CARS will have on the following: Main condenser Proposed Question: Common 37 Given the following conditions:

• The Unit is at 100% power.

• HS-3331A, Condenser Vacuum Pump hand switch on CR-53 is in AUTO.

• Condenser Vacuum starts to degrade slowly and is at approximately 4 inches Hg absolute pressure.

Which ONE (1) of the following describes the design of the Condenser Vacuum System that will prevent continued degradation of Condenser vacuum?

P-054, Condenser Vacuum Pump will START, A. PV-3335B, Alternate Suction Valve is OPEN (HOLD), and PV-3335A, Primary Suction Valve (HOG) is CLOSED placing the Air Assisted Air Ejectors in service.

B. and operate in the HOGGING mode after the inlet diaphragm valves are OPEN.

C. PV-3335A, Primary Suction Valve will OPEN (HOG), and PV-3335B, Alternate Suction Valve will CLOSE (HOLD) placing the Air Assisted Air Ejectors in service.

D. then immediately shift to the HOGGING mode of operation to prevent further loss of vacuum.

Proposed Answer: A Explanation:

A. Correct. At 4 Hg absolute, the Alternate Suction Valve is open and the Primary Suction Valve is closed to place the Air Assisted Air Ejectors in service.

B. Incorrect. Plausible because the Vacuum Pump will start with the handswitch in AUTO, however, the inlet diaphragm valves do not open until a low differential pressure is sensed across the valve.

C. Incorrect. Plausible because this is the reverse of the correct condition. When vacuum is low (i.e.,

during start up) the Primary Suction Valve is opened and the Alternate Suction Valve is closed.

D. Incorrect. Plausible because this is the condition of the Vacuum Pump upon normal startup. It is in the HOGGING mode at startup; it does not shift to this mode.

Page 137 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Technical Reference(s) SD-SO23-190, Figure 1 Attached w/ Revision # See SD-SO23-190, Pages 13 Comments / Reference Proposed references to be provided during examination: None Learning Objective: IDENTIFY the alarms, automatic actions, and effect on other plant systems 53207 and equipment for a Loss of Condenser Vacuum.

Question Source: Bank # 126686 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam 2003 NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Page 138 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SO23- Revision # 11 Page 139 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SD-SO23-190, Page 8 Revision # 11 Comments /

Reference:

From SD-SO23-190, Page 13 Revision # 11 Page 140 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SD-SO23-190, Page 14 Revision # 11 Page 141 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SD-SO23-190, Page 15 Revision # 11 Page 142 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 2 K/A # 071 K5.04 Importance Rating 2.5 Waste Gas Disposal System: Knowledge of the operational implication of the following concepts as they apply to the Waste Gas Disposal System: Relationship of hydrogen / oxygen concentrations to flammability Proposed Question: Common 38 In accordance with Licensee Controlled Specification 3.3.107, the Waste Gas Holdup System Hydrogen and Oxygen Monitor alarm setpoints are set to ensure which ONE (1) of the following?

A. Hydrogen concentration does not exceed 2% when Oxygen concentration is greater than 2%.

B. Oxygen concentration does not exceed 2% when Hydrogen concentration is greater than 2%.

C. Hydrogen concentration does not exceed 2% when Oxygen concentration is greater than 4%.

D. Oxygen concentration does not exceed 2% when Hydrogen concentration is greater than 4%.

Proposed Answer: D Explanation:

A. Incorrect. Plausible because it could be thought that hydrogen concentration must be kept at the same low value as oxygen concentration.

B. Incorrect. Plausible because the oxygen concentration is correct, however, the hydrogen concentration is not high enough at this point.

C. Incorrect. Plausible because it could be thought that hydrogen concentration must be kept low when oxygen concentration is high.

D. Correct. This coincides with the setpoint of annunciator 61B17 and the information listed in the LCS.

Technical Reference(s) Licensee Controlled Specification 3.3.107 Attached w/ Revision # See SO23-15-61.B.17 Comments / Reference Proposed references to be provided during examination: None Learning Objective: Given indications of abnormal waste gas system oxygen or hydrogen levels, 52668 DETERMINE the action required per SO23-15-61.B and explain the reason for the actions taken.

Page 143 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Question Source: Bank # 126434 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10, 13 55.43 Comments /

Reference:

From Licensee Controlled Specification 3.3.107 Revision # 2 Page 144 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SO23-15-61.B.17 Revision # 12 Page 145 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 008 AA1.08 Importance Rating 3.8 Pressurizer Vapor Space Accident: Ability to operate and/or monitor the following as they apply to the Pressurizer Vapor Space Accident:

PRT level, pressure, and temperature Proposed Question: Common 39 Given the following conditions following a Unit trip and SIAS actuation:

• Pressurizer pressure is 1720 psia and slowly LOWERING.

• Quench Tank pressure is 20 psig and STABLE.

• Pressurizer level is 28% and RISING.

• Reactor Coolant System Tcold is 545°F and STABLE.

• Steam Generator narrow range levels are 21% and slowly RISING.

• Steam Generator pressures are 1000 psia and STABLE.

• Containment temperature is 125°F and slowly RISING.

• Containment pressure is 3 psig and slowly RISING.

• Containment humidity is RISING.

Which ONE (1) of the following could be the cause of the above conditions?

A. Reactor Coolant System Cold Leg is leaking.

B. Pressurizer steam space is leaking.

C. Pressurizer Safety Valve is leaking.

D. Steam Generator level condensing pot is leaking.

Proposed Answer: B Explanation:

A. Incorrect. Plausible because Pressurizer pressure is lowering and Containment humidity level is rising, however, Pressurizer level is also rising which would be indicative of a Pressurizer steam space leak.

B. Correct. Given the conditions listed, a Pressurizer steam space leak is occurring.

C. Incorrect. Plausible because Pressurizer pressure is lowering, however, Quench Tank pressure is not changing.

D. Incorrect. Plausible because Containment humidity level is rising and Steam Generator levels are rising, however, if the condensing pot were leaking one would expect indicated narrow range level to be much higher and Steam Generator pressures would not be stable.

Page 146 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Technical Reference(s) SO23-3-1.11, L&S 1.2 Attached w/ Revision # See Comments / Reference Proposed references to be provided during examination: None Learning Objective: Given plant conditions, PREDICT and EXPLAIN the response of containment 55326 instrumentation to adverse containment conditions during a Loss of Coolant Accident.

Question Source: Bank # 126590 Modified Bank # (Note changes or attach parent)

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

Reference:

From SO23-3-1.11, L&S 1.2 Revision # 10-1 Page 147 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 009 EK2.03 Importance Rating 3.0 Small Break LOCA: Knowledge of the interrelations between the small break LOCA and the following: SGs Proposed Question: Common 40 Given the following conditions:

• A Small Break Loss of Coolant Accident is in progress.

• The Safety Injection Actuation Signal has actuated.

• All systems are operating as expected.

Per the stated conditions, which ONE (1) of the following is the basis for maintaining a secondary heat sink?

A. To minimize boron stratification of the RCS.

B. To provide for Containment temperature and pressure control.

C. Reflux boiling is the primary means of heat removal prior to voiding in the hot legs.

D. RCS pressure may remain so high that cooling from the injection flow alone is inadequate to remove decay heat.

Proposed Answer: D Explanation:

A. Incorrect. Plausible because boron stratification is a concern, however, more so during a large break LOCA.

B. Incorrect. Plausible because it could be thought that if Containment pressure remains below 3.4 psig then a Containment Isolation Actuation Signal would not be initiated. The issue here however, is Containment cooling which is actuated by the SIAS.

C. Incorrect. Plausible because this statement is true if talking about a large break LOCA.

D. Correct. Any Reactor Coolant System pressure that remains above the shutoff head of the High Pressure Safety Injection Pumps will result in a lowering of inventory without the attendant makeup. The Steam Generators provide decay heat removal until the system is cooled down and depressurized.

Technical Reference(s) SO23-14-3, Section 2.0 Attached w/ Revision # See Comments / Reference Proposed references to be provided during examination: None Page 148 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Learning Objective: PREDICT and EXPLAIN the response of major plant systems, equipment and 53006 parameters to a loss of forced circulation Question Source: Bank # 75386 Modified Bank # (Note changes or attach parent)

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

Reference:

From SO23-14-3, Section 2.0 Revision # 8 Page 149 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 011 EK3.03 Importance Rating 4.1 Large Break LOCA: Knowledge of the reasons for the following responses as they apply to the large break LOCA: Starting auxiliary feed pumps and flow, EDG, and service water pumps Proposed Question: Common 41 Given the following conditions:

• Unit 3 is in a normal at power alignment when a Large Break Loss of Coolant Accident occurs.

Which ONE (1) of the following is the reason the Unit 3 Emergency Diesel Generators start but do not load on a Safety Injection Actuation Signal caused by a Large Break Loss of Coolant Accident?

In this condition...

A. 1E 4 kV Bus voltage must degrade below 3796 volts before the diesels will load.

B. the Unit 2 1E 4 kV Buses are the Normal Preferred Source.

C. 1E 4 kV Bus voltage must degrade below 4200 volts before the diesels will load.

D. the Unit 3 Reserve Auxiliary Transformers are the Normal Preferred Source.

Proposed Answer: D Explanation:

A. Incorrect. Plausible because under certain conditions it could be thought that the Loss of Voltage Signal (<3796 V) must actuate, however, this might only be true if the EDG breaker was already closed (See Reference for annunciator 63B05).

B. Incorrect. Plausible because Unit 2 1E 4 kV Buses are the Alternate Preferred Source.

C. Incorrect. Plausible because nominal bus voltage is 4360 VAC and the degraded grid relay will not actuate until the setpoint is 4154 volts.

D. Correct. The Normal Preferred Source is that Units Reserve Auxiliary Transformer and it remains that way as long as voltage does not degrade.

Technical Reference(s) SD-SO23-120, Page 136 Attached w/ Revision # See SO23-15-63.B.05, Step 1.0 Comments / Reference Proposed references to be provided during examination: None Page 150 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Learning Objective: Given an operational condition of the 1E 4.16 KV, SDVS (Sustained Degraded 52761 Signal) or DGVSS (Degraded Grid Voltage with Stained SIAS) addressed by a procedural precaution, limitation, or administrative requirement, STATE the limiting condition and the basis for that limiting condition.

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

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

Reference:

From SD-SO23-120, Page 136 Revision # 19 Comments /

Reference:

From SO23-15-63.B.05, Step 1.0 Revision # 12 Page 151 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SO23-15-63.B.05, Step 1.0 Revision # 12 Comments /

Reference:

From SD-SO23-120, Page 136 Revision # 19 Page 152 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 015 / 17 G 2.1.20 Importance Rating 4.6 RCP Malfunctions: Conduct of Operations: Ability to interpret and execute procedure steps Proposed Question: Common 42 Given the following conditions during a plant heatup from MODE 4:

• P-003 Reactor Coolant Pump lower thrust bearing is at 225°F and rising at approximately 1°F per minute.

• P-003 upper thrust bearing is 170°F and rising at 1°F every 3 minutes.

• P-003 upper and lower guide bearings are at 170°F and rising at 1°F per minute.

• P-003 anti-reverse rotation bearing is at 170°F and rising at 1°F per minute.

• The Reactor Trip Circuit Breakers are closed.

Which ONE (1) of the following describes the plant condition and procedural action required?

A Reactor Coolant Pump bearing temperature limit...

A. has been exceeded. Start the Oil Lift and Anti-Reverse Rotation Device Pumps and stop the RCP per SO23-5-1.3, Plant Startup from Cold Shutdown to Hot Standby.

B. has been exceeded. Ensure the associated Steam Generator Low Flow Trip has been bypassed then stop the RCP per SO23-3-1.7, Reactor Coolant Pump Operation.

C. will be exceeded in approximately 15 minutes. When the limit is exceeded, stop the RCP per SO23-3-1.7, Reactor Coolant Pump Operation.

D. will be exceeded in approximately 15 minutes. When the limit is exceeded, stop the heatup and determine the cause per SO23-5-1.3, Plant Startup from Cold Shutdown to Hot Standby.

Proposed Answer: B Page 153 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible because this action is outlined in Step 6.4 for RCP P-002 for high upper thrust bearing temperatures, however, it is an automatic response when the RCP is stopped.

B. Correct. This is the required action per SO23-3-1.7 when in MODES 3, 4, or 5.

C. Incorrect. Plausible because the upper and lower guide bearings will exceed 185°F in 15 minutes and exceed the maximum allowable temperatures per SO23-3-1.7, Attachment 2, however, the Steam Generator Low Flow Trip must be bypassed in this MODE.

D. Incorrect. Plausible because there are RCP temperature limits in SO23-5-1.3 (see Reference),

however, they refer to CCW from the Seal Water Heat Exchanger and Controlled Bleedoff (CBO) flow. Additionally, the upper and lower guide bearings will exceed 185°F in 15 minutes and exceed the maximum allowable temps per SO23-3-1.7, Attachment 2.

Technical Reference(s) SO23-3-1.7, Steps 6.3 & 6.4 Attached w/ Revision # See SO23-5-1.3, Attachment 10 Comments / Reference SO23-3-1.7, Attachment 2 Proposed references to be provided during examination: None Learning Objective: DESCRIBE the operation of the following alarms associated with the Reactor 55277 Coolant System including setpoint, possible causes, and effects on system and overall plant operation, as applicable:

- RCP THRUST BEARINGS TEMP HI alarm Question Source: Bank # 127313 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7, 10 55.43 Page 154 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SO23-3-1.7, Step 6.3 Revision # 32 Page 155 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SO23-3-1.7, Attachment 2 Revision # 32 Comments /

Reference:

From SO23-3-1.7, Step 6.4 Revision # 32 Page 156 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SO23-5-1.3, Attachment 10 Revision # 31 Page 157 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 022 AK1.02 Importance Rating 2.7 Loss of Reactor Coolant Makeup: Knowledge of the operational implications of the following concepts as they apply to the Loss of Reactor Coolant Makeup: Relationship of charging flow to pressure differential between charging and RCS Proposed Question: Common 43 Given the following conditions at 100% power:

• Pressurizer pressure is 2275 psia and rising.

• Volume Control Tank pressure is 15 psig and lowering.

• Volume Control Tank level is 15% and lowering.

• Charging Pump P-191 is in service.

• Charging Pump P-191 Suction Pulsation Dampener bladder pressure is lowering.

Which ONE (1) of the following identifies the response of Charging Pump flow to the listed conditions?

Charging Pump flow will...

A. decrease in 15 gpm increments due to lowering VCT pressure.

B. remain the same until Charging Pump suction shifts to the RWST.

C. decrease in response to increasing Pressurizer pressure.

D. remain the same as long as the suction bladder pressure remains above 17 psig.

Proposed Answer: A Page 158 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Explanation:

A. Correct. When VCT pressure drops below 25 psig, Charging Pump flow is affected in 15 gpm increments. This corresponds to one of three plungers in the Charging Pump losing suction pressure.

B. Incorrect. Plausible because the VCT suction is ready to shift to the RWST and under normal circumstances this results in a loss of suction pressure because of the VCT cover gas. For the conditions listed, the suction pressures are very close to being the same but the underlying reason has to do with Charging Pump design.

C. Incorrect. Plausible because it could be thought that an increase in Pressurizer pressure will affect Charging Pump flow with the conditions listed, however, flow does decrease but not for this reason.

D. Incorrect. Plausible because this is the correct minimum value for the suction pulsation dampener.

Lowering dampener pressure will only affect the Charging Pump suction pressure fluctuations.

Charging pump flow is unaffected by the suction or discharge bladder pressures.

Technical Reference(s) SO23-3-2.2, L&S 1.3, 2.10, & 2.11 Attached w/ Revision # See SO23-3-2.1.2, L&S 1.12 & 2.1 Comments / Reference Proposed references to be provided during examination: None Learning Objective: Given a malfunction or misoperation of the charging pumps or associated 53437 support systems, PREDICT the expected response of the associated components, instruments and controls.

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 5, 6, 7 55.43 Page 159 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SO23-3-2.2, L&S 1.3 Revision # 27 Comments /

Reference:

From SO23-3-2.2, L&S 2.10 Revision # 27 Comments /

Reference:

From SO23-3-2.2, L&S 2.11 Revision # 27 Page 160 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SO23-3-2.1.2, L&S 2.1 Revision # 2 Comments /

Reference:

From SO23-3-2.1.2, L&S 1.12 Revision # 2 Page 161 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 025 AK2.02 Importance Rating 3.2 Loss of RHR System: Knowledge of the interrelations between the Loss of Residual Heat Removal System and the following: LPI or Decay Heat Removal / RHR pumps Proposed Question: Common 44 Which ONE (1) of the following would be an indication that a Shutdown Cooling Pump was not meeting its required operating parameters while in SO23-13-15, Loss of Shutdown Cooling?

Shutdown Cooling Pump...

A. flow is less than 2600 gpm.

B. amperage is fluctuating greater than +/- 10 amps.

C. suction leg level is 22 inches in the Hot Leg.

D. suction leg pressure is less than 10 psig.

Proposed Answer: B Explanation:

A. Incorrect. Plausible because it could be thought that this flow rate was not appropriate, however, the Tech Spec minimum flow rate is 2200 gpm and the Exit Conditions for SO23-13-15 require flow 2500 gpm.

B. Correct. With amperage fluctuating greater than +/- 10 amps, the Shutdown Cooling Pump should be secured per SO23-13-15.

C. Incorrect. Plausible because it could be thought that Hot Leg level is too low, however, the pump can be vented either running or shutdown as long as level is above 21 inches.

D. Incorrect. Plausible because it could be thought that a minimum pressure exists for the Shutdown Cooling Pump, however, it is minimum level in the Hot Leg that is of concern.

Technical Reference(s) SO23-13-15, Step 4h Attached w/ Revision # See SO23-13-15, Exit Conditions Comments / Reference SO23-13-15, Attachments 2 & 3 Proposed references to be provided during examination: None Learning Objective: Per the Loss of Shutdown Cooling procedure, SO23-13-15, DESCRIBE:

55323 - The basis for each step, caution, or note.

- The expected plant response for each step.

Page 162 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Question Source: Bank #

Modified Bank # (Note changes or attach parent)

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

Reference:

From SO23-13-15, Step 4h Revision # 18 Comments /

Reference:

From SO23-13-15, Exit Conditions Revision # 18 Page 163 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SO23-13-15, Attachment 2 Revision # 18 Comments /

Reference:

From SO23-13-15, Attachment 3 Revision # 18 Page 164 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 026 AK3.03 Importance Rating 4.0 Loss of Component Cooling Water: Knowledge of the reasons for the following responses as they apply to the Loss of Component Cooling Water: Guidance actions contained in EOP for loss of CCW Proposed Question: Common 45 Given the following conditions:

• Unit 2 is at 1% power with a power ascension in progress.

• Component Cooling Water is lost to the Non-Critical Loop.

• It is determined that flow cannot be restored.

Which ONE (1) of the following actions is required per AOI SO23-13-7, Loss of Component Cooling Water / Saltwater Cooling?

A. Cross connect the Component Cooling Water Non-Critical Loop to Unit 3, perform a Normal Plant Shutdown and then secure the Reactor Coolant Pumps.

B. Perform a Rapid Plant Shutdown and then secure the Reactor Coolant Pumps.

C. Cross connect the Component Cooling Water Non-Critical Loop to Unit 3, perform a Rapid Plant Shutdown and then secure the Reactor Coolant Pumps.

D. Trip the Reactor and then secure the Reactor Coolant Pumps.

Proposed Answer: D Explanation:

A. Incorrect. Plausible because it could be thought that the Non-Critical Loops could be cross connected, however, the valves are interlocked (when one set of supply and return valves are opened the other set will go closed). Cross connecting CCW Loops is not allowed until MODE 5.

B. Incorrect. Plausible because in MODES 3-5 the affected RCPs would be stopped, however, operation in MODES 1 or 2 requires a Reactor trip.

C. Incorrect. Plausible because it could be thought that the Non-Critical Loops could be cross connected, however, the valves are interlocked (when one set of supply and return valves are opened the other set will go closed). Cross connecting CCW Loops is not allowed until MODE 5.

D. Correct. Per Step 4 of SO23-13-7, the Reactor would be tripped and 5 seconds after the CEAs are inserted the RCPs would be tripped.

Page 165 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Technical Reference(s) SO23-13-7, Step 4 Attached w/ Revision # See SD-SO23-400, Page 24 Comments / Reference SD-SO23-400, Figure 3 SO23-13-7, Attachment 4 Proposed references to be provided during examination: None Learning Objective: Given an operational condition of the Component Cooling Water System, 55868 DESCRIBE the precaution, limitation or administrative requirement applicable to the situation.

Question Source: Bank # 126611 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam 2003 NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 4, 7, 10 55.43 Comments /

Reference:

From SO23-13-7, Step 4 Revision # 13 Page 166 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SD-SO23-400, Page 24 Revision # 18 Page 167 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SD-SO23-400, Figure 3 Revision # 18 Comments /

Reference:

From SO23-13-7, Attachment 4 Revision # 13 Page 168 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 027 AA1.01 Importance Rating 4.0 Pressurizer Pressure Control System Malfunction: Ability to operate and/or monitor the following as they apply to the Pressurizer Pressure Control Malfunctions: Pressurizer heaters, sprays, and PORVs Proposed Question: Common 46 Given the following conditions at 100 % power:

• Pressurizer pressure is 2190 psia.

• Pressurizer pressure setpoint is at 2250 psia.

• Pressurizer pressure Channel Y has failed high.

• Pressurizer pressure Channel X is in control.

• All Pressurizer heaters are in AUTO.

Which ONE (1) of the following is correct for the stated conditions?

A. One set of 1E Pressurizer Heaters is OFF; all other Pressurizer Heaters are ON.

B. All Proportional, Backup and 1E Pressurizer Heaters are ON.

C. Both sets of 1E Pressurizer Heaters are ON; all other Pressurizer Heaters are OFF.

D. All Proportional and Backup Pressurizer heaters are OFF.

Proposed Answer: B Explanation:

A. Incorrect. Plausible because it could be thought that given the difference between actual pressure and setpoint that not all the Proportional Heaters are energized. This, in conjunction with the misconception that the 1E Pressurizer Heaters are the Proportional Heaters could yield this response.

B. Correct. Given the conditions listed, all PZR Heaters would be energized.

C. Incorrect. Plausible because of the misconception that the 1E Pressurizer Heaters are the Proportional Heaters and given the fact that with pressure less than 2225 psia, all Proportional Heaters are fully energized.

D. Incorrect. Plausible because if Channel Y were selected then all PZR Heaters would trip via the position of HS-0100A and the high failure.

Technical Reference(s) SD-SO23-360, Page 78 Attached w/ Revision # See SO23-3-1.10, Attachment 8 Comments / Reference SO23-3-1.10, L&S 1.1 SD-SO23-360, Figures III-8, III-9 & III-10 Page 169 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Proposed references to be provided during examination: None Learning Objective: DESCRIBE the cause/effect relationships associated with the Pressurizer 56419 Pressure and Level Control System and an increasing or decreasing Pressurizer pressure and/or level.

Question Source: Bank # 112949 Modified Bank # (Note changes or attach parent)

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

Reference:

From SD-SO23-360, Page 78 Revision # 17 Comments /

Reference:

From SD-SO23-360, Page 79 Revision # 17 Page 170 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SO23-3-1.10, Attachment 8 Revision # 18 Comments /

Reference:

From SO23-3-1.10, L&S 1.1 Revision # 18 Page 171 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SD-SO23-360, Figure III-9 Revision # 17 Page 172 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SD-SO23-360, Figure III-8 Revision # 17 Comments /

Reference:

From SD-SO23-360, Figure III-10 Revision # 17 Page 173 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 029 EK2.06 Importance Rating 2.9 ATWS: Knowledge of the interrelations between the following and an ATWS: Breakers, relays, and disconnects Proposed Question: Common 47 Activation of the Anticipated Transient Without Scram / Diversified Scram System (ATWS /

DSS) on High-High Pressurizer pressure directly results in the OPENING of which ONE (1) of the following components?

A. Reactor Trip Breakers.

B. MG Set Supply Breakers.

C. MG Set Output Contactors.

D. CEDMCS Power Switch Assemblies.

Proposed Answer: C Explanation:

A. Incorrect. Plausible because it could be thought that a signal separate from the shunt and UV coils is sent to the Reactor Trip Breakers since they are downstream of the MG Set Output Contactors.

B. Incorrect. Plausible because there is a misconception that the MG Set Supply Breakers trip. This is because circuit control power is supplied from the upstream (supply breaker) side of the MG Set.

See Figure 3.

C. Correct. This is the component that opens on an ATWS signal.

D. Incorrect. Plausible because it could be thought that this is what opens, however, the Power Switch Assemblies direct power to the CEDM coils. See Figure 1A.

Technical Reference(s) SD-SO23-520, Page 2 Attached w/ Revision # See SD-SO23-520, Figure 3 Comments / Reference SD-SO23-510, Figure 1A Proposed references to be provided during examination: None Learning Objective: STATE the functions of the ATWS/DSS System.

52389 Question Source: Bank # 73940 Modified Bank # (Note changes or attach parent)

New Page 174 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Question History: Last NRC Exam 1998 NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7, 8 55.43 Comments /

Reference:

From SD-SO23-520, Page 2 Revision # 6 Comments /

Reference:

From SD-SO23-520, Figure 3 Revision # 6 Page 175 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SD-SO23-510, Figure 1A Revision # 10 Page 176 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 038 EA2.13 Importance Rating 3.1 Steam Generator Tube Rupture: Ability to determine or interpret the following as they apply to a SGTR: Magnitude of rupture Proposed Question: Common 48 Which ONE (1) of the following parameters would assist the operator in determining the size of a Steam Generator Tube Rupture while at 10% power?

A. Main Feedwater flow.

B. Steam Generator pressure.

C. Main Steam flow.

D. Steam Generator level.

Proposed Answer: A Explanation:

A. Correct. Main Feedwater flow experiences a decrease in proportion to the size of the rupture while at power.

B. Incorrect. Plausible because Steam Generator pressure may experience a minor change, however, it could not be used to determine the size of the rupture at power.

C. Incorrect. Plausible because Main Steam flow may experience a minor change, however, it could not be used to determine the size of the rupture at power.

D. Incorrect. Plausible because Steam Generator level may experience a minor change, however, it is not well suited to determine the size of the rupture at power.

Technical Reference(s) SO23-14-4, Attachment 1, Event Attached w/ Revision # See Description 2.0 Comments / Reference Proposed references to be provided during examination: None Learning Objective: Given plant conditions, PREDICT and EXPLAIN the response of major plant 52660 systems, equipment and parameters to a Steam Generator Tube Rupture.

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Page 177 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 5 55.43 Comments /

Reference:

From SO23-14-4, Attachment 1, Event Description 2.0 Revision # 7 Page 178 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 055 EK1.01 Importance Rating 3.3 Station Blackout: Knowledge of the operational implications of the following concepts as they apply to the Station Blackout: Effect of battery discharge rates on capacity Proposed Question: Common 49 Given the following conditions during a Station Blackout:

• Battery D5 is discharging during a Station Blackout.

Assuming the load on DC Bus 5 does not change, which ONE (1) of the following statements correctly describes Battery D5 discharge rate (amps) as the battery is expended?

As D5 Battery voltage lowers, the discharge rate in amps will...

A. increase until the design battery capacity is exhausted.

B. be fairly constant until the design battery capacity (amp-hours) is exhausted and then will rapidly decrease.

C. decrease until the design battery capacity is exhausted.

D. initially decrease until approximately 50% design capacity had been expended and then increase until the battery has been exhausted.

Proposed Answer: A Explanation:

A. Correct. For DC systems, P = IE, therefore, as battery voltage drops the discharge amps will rise.

Additionally, internal energy losses and the batteries internal resistance generated while discharging leads to an increase in the discharge rate.

B. Incorrect. Plausible if thought that battery voltage does not lower during discharge.

C. Incorrect. Plausible because load is decreased on the battery due to the load reductions that are performed at 45 and 90 minutes, however, the question stem states that load does not change.

D. Incorrect. Plausible because load is decreased on the battery due to the load reductions that are performed at 45 and 90 minutes, however, the question stem states that load does not change.

Technical Reference(s) SO23-6-15, L&S 1.1 Attached w/ Revision # See Comments / Reference Proposed references to be provided during examination: None Page 179 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Learning Objective: DESCRIBE the configuration and operational characteristics of 1E 125 VDC 80605 and 120 VAC Electrical System components.

Question Source: Bank # 151673 Modified Bank # (Note changes or attach parent)

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

Reference:

From SO23-6-15, L&S 1.1 Revision # 26 Page 180 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 056 AK3.02 Importance Rating 4.4 Loss of Offsite Power: Knowledge of the reasons for the following responses as they apply to the Loss of Offsite Power: Actions contained in EOP for loss of offsite power Proposed Question: Common 50 Given the following conditions:

• Unit 3 was at 100% power.

• A Loss of Offsite Power occurred 15 minutes ago.

• 1E 4 kV Buses have been reenergized from the Emergency Diesel Generators.

• There are no Reactor Coolant Pumps running.

• Letdown flow has been restored.

• Pressurizer level is 68% and slowly lowering.

• Pressurizer pressure is 2320 psia and slowly lowering.

• Main Steam Isolation Valves are closed.

In accordance with SO23-12-7, Loss of Forced Circulation / Loss of Offsite Power, which ONE (1) of the following actions is required?

Operate the...

A. Auxiliary Feedwater System to establish 200 gpm flow rate to each Steam Generator for at least 5 minutes.

B. Main Feedwater System to establish at least one intact Steam Generator level between 40% and 80% on the narrow range indication.

C. Auxiliary Feedwater System to establish at least one intact Steam Generator level between 40% and 80% on the narrow range indication.

D. Main Feedwater System and use Main Steam Safety Valves to steam the Steam Generators to establish natural circulation and decay heat removal.

Proposed Answer: C Page 181 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible because this action would be required if Auxiliary Feedwater flow were lost, however, based on the information in the stem there is no indication that this occurred. If flow were lost, the value would be 130 to 150 for 5 minutes then raise to 200 gpm.

B. Incorrect. Plausible because this is the correct range for Steam Generator level indication, however, Main Feedwater System is not available given the conditions listed.

C. Correct. This is the required action per Step 9 of SO23-12-7.

D. Incorrect. Plausible because the Atmospheric Dump Valves may not be available to control Reactor Coolant System temperature, however, neither is Main Feedwater.

Technical Reference(s) SO23-12-7, Step 9 Attached w/ Revision # See SO23-12-6, Step 7 Comments / Reference Proposed references to be provided during examination: None Learning Objective: DESCRIBE the general sequence of events during a Loss of Offsite Power or 52844 Station Blackout event.

Question Source: Bank # 126657 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam 2003 NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 Comments /

Reference:

From SO23-12-7, Step 9 Revision # 20 Page 182 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SO23-12-6, Step 7 Revision # 21 Page 183 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 057 AK3.01 Importance Rating 4.1 Loss of Vital AC Instrument Bus: Knowledge of the reasons for the following responses as they apply to the Loss of Vital AC Instrument Bus:

Actions contained in EOP for loss of vital AC electrical instrument bus Proposed Question: Common 51 Given the following conditions:

• Vital Bus Y-02 has been lost due to Inverter failure with the Unit operating at power.

• Vital Bus Y-02 has not been energized from the Alternate Source.

Which ONE (1) of the following actions is required and the reason for that action?

E. TRANSFER Pressurizer Level Setpoint to Setpoint LS1 to allow restoring the Pressurizer Level Controller to AUTO.

F. Manually ACTUATE Train A FHIS, TGIS, CRIS and CPIS due to Train B FHIS, TGIS, CRIS and CPIS actuation.

G. VERIFY Auxiliary Feedwater flow to Steam Generator E-088 due to EFAS Trip Paths 2 and 4 actuation.

H. CLOSE the failed open Atmospheric Dump Valve (HV-8421) due to loss of the pressure input from the Main Steam pressure transmitter.

Proposed Answer: A Explanation:

E. Correct. Loss of Y-02 will require transfer of the PZR Level Setpoint to LS1 to allow restoration of the Pressurizer Level Controller to AUTO.

F. Incorrect. Plausible because Train B FHIS, TGIS, CRIS and CPIS will actuate, however, there is no requirement to actuate the Train A components.

G. Incorrect. Plausible because five (5) AFW Valves open due to EFAS Trip Path actuation, however, there is no associated pump start.

H. Incorrect. Plausible because this valve fails for the reason listed, however, the valve fails closed.

Technical Reference(s) SO23-13-18, Attachments 1 & 2 Attached w/ Revision # See Comments / Reference Proposed references to be provided during examination: None Learning Objective: ANALYZE normal and abnormal operations of the 1E 125 VDC and 120 VAC Page 184 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 80607 Electrical System.

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 8, 10 55.43 Page 185 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SO23-13-18, Attachment 2 Revision # 8 Page 186 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SO23-13-18, Attachment 2 Revision # 8 Page 187 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 058 AK1.01 Importance Rating 2.8 Loss of DC Power: Knowledge of the operational implications of the following concepts as they apply to Loss of DC Power: Battery charger equipment and instrumentation Proposed Question: Common 52 Given the following conditions:

• Annunciator 63A34 - 2D3 125 VDC BUS TROUBLE is in alarm.

• Annunciator 63A54 - 2D3 CHARGER TROUBLE is also in alarm.

• 2D3 Bus voltage initially rose to 157 VDC for one (1) minute and is now 110 VDC and lowering.

• There are no other Control Room alarms at this time.

Which ONE (1) of the following is the likely cause of these conditions?

2D3 Battery...

A. Breaker has opened.

B. Charger has shutdown on low voltage.

C. Charger has an AC PHASE FAILURE.

D. Charger has shutdown on high voltage.

Proposed Answer: D Explanation:

A. Incorrect. Plausible because this could create the conditions listed, however, there would be an Annunciator associated with the Battery Breaker being open.

B. Incorrect. Plausible because the Battery Charger has shutdown, however, there is no shut down associated with low voltage (alarm condition only).

C. Incorrect. Plausible because this condition could shut down the Battery Charger, however, AC Phase Failure is initiated from a low voltage condition on the 480 VAC bus.

D. Correct. Once battery charger voltage rises above 152 VDC for greater than 10 seconds, the battery charger will shutdown. This condition brings in both annunciators listed.

Technical Reference(s) SO23-63.A.34 & 54 Attached w/ Revision # See Comments / Reference Page 188 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Proposed references to be provided during examination: None Learning Objective: ANALYZE normal and abnormal operations of the 1E 125 VDC and 120 VAC 80607 Electrical System.

Question Source: Bank # 127373 Modified Bank # (Note changes or attach parent)

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

Reference:

From SO23-63.A.34 Revision # 8 Page 189 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SO23-63.A.54 Revision # 8 Page 190 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SO23-63.A.44 Revision # 8 Page 191 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 065 AK3.04 Importance Rating 3.0 Loss of Instrument Air: Knowledge of the reasons for the following responses as they apply to the Loss of Instrument Air: Cross-over to backup air supplies Proposed Question: Common 53 Given the following conditions:

• 2/3 PI-5344A, Instrument Air pressure, has been steadily decreasing from 105 psig as read on CR-61.

• Pressure continues to decrease until it stabilizes at 87 psig.

Which ONE (1) of the following is the reason that Instrument Air pressure stabilized?

A. The LAG 1 Instrument Air Compressor started and is fully loaded.

B. The LEAD Instrument Air Compressor started and is fully loaded.

C. PCV-5458, Nitrogen Backup to Instrument Air supply valve has opened.

D. PCV-5354, Respiratory and Service Air cross-connect valve has opened Proposed Answer: D Explanation:

A. Incorrect. Plausible because this compressor is available and will automatically start at 98 psig, however, the setpoint for full loading is ~94 psig.

B. Incorrect. Plausible because this compressor is available and will automatically start at 106 psig, however, the setpoint for full loading is ~102 psig.

C. Incorrect. Plausible because this valve is available and will automatically open, however, the setpoint for PCV-5458 is 83 psig.

D. Correct. This valve opens at 88 psig and will attempt to maintain pressure >84 psig.

Technical Reference(s) SO23-13-5, Attachment 6, L&S 1.2 Attached w/ Revision # See SO23-1-1, Attachment 22, L&S 1.6 Comments / Reference Proposed references to be provided during examination: None Learning Objective: DESCRIBE the configuration and operational characteristics of Instrument and 72865 Respiratory & Service Air Systems components.

Page 192 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Question Source: Bank # Not assigned Modified Bank # (Note changes or attach parent)

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

Reference:

From SO23-13-5, Attachment 6, L&S 1.2 Revision # 7 Page 193 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SO23-1-1, Attachment 22, L&S 1.6 Revision # 19 Page 194 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # AA2.02 Importance Rating 3.5 Generator Voltage and Electric Grid Disturbances: Ability to determine and interpret the following as they apply to Generator Voltage and Electric Grid Disturbances: Voltage outside the generator capability curve Proposed Question: Common 54 The Main Generator is operating outside (above and to the right) of the Generator Capability Curve.

Which ONE (1) of the following corrective actions could restore acceptable operating conditions?

A. Boost (raise) voltage or raise Turbine load.

B. Buck (lower) voltage or lower Turbine load.

C. Boost (raise) voltage or lower Turbine load.

D. Buck (lower) voltage or raise Turbine load.

Proposed Answer: B Explanation:

A. Incorrect. Plausible if thought that this condition was improving the operating characteristics of the Main Generator, however, in both instances conditions are worsening.

B. Correct. Lowering voltage will reduce Generator reactive load and move the operating point left on the X-axis of the Generator Capability Curve. Lowering Turbine load will reduce real load and move the operating point down on the Y-axis of the Generator Capability Curve.

C. Incorrect. Plausible because lowering Turbine load will work, however, raising voltage will drive the operating point to the right on the X-axis of the Generator Capability Curve. Additionally, examinee must recognize that the BOOST and BUCK values shown on the Generator Capability Curve reflect the conditions necessary to drive generator MVARS in that specific direction.

D. Incorrect. Plausible because lowering voltage will work, however, raising Turbine load will move the operating point up on the Y-axis of the Generator Capability Curve.

Technical Reference(s) SO23-6-28, Attachment 5 Attached w/ Revision # See Comments / Reference Proposed references to be provided during examination: SO23-6-28, Attachment 5 Learning Objective: ANALYZE normal and abnormal operations of the Main Generator and 22 kV 73267 System.

Page 195 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Question Source: Bank # 77138 Modified Bank # (Note changes or attach parent)

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

Reference:

From SO23-6-28, Attachment 5 Revision # 13 Page 196 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # CE-E05 EA2.02 Importance Rating 3.4 Steam Line Rupture - Excessive Heat Transfer: Ability to determine and interpret the following as they apply to the Excess Steam Demand:

Adherence to appropriate procedures and operation within the limitations in the facilities license and amendments Proposed Question: Common 55 Given the following conditions:

• An unisolable Excess Steam Demand Event has occurred outside Containment.

• The affected Steam Generator has blown dry.

• RCS temperature is 350°F.

• Pressurizer pressure is 1290 psia.

Which ONE (1) of the following actions is required for the listed conditions?

A. Secure all running RCPs.

B. Stabilize RCS temperature and depressurize the RCS.

C. Reduce RCS temperature to establish 20°F subcooling.

D. Stabilize RCS pressure and temperature at current values.

Proposed Answer: B Explanation:

A. Incorrect. Plausible because PZR pressure is low, however, two RCPs have already been secured and NPSH requirements are being met.

B. Correct. Given the conditions listed, this is the next set of actions to be performed.

C. Incorrect. Plausible because this is used throughout the EOIs, however, it is not required in the ESDE procedure.

D. Incorrect. Plausible because temperature is stabilized at the current value to prevent a PTS condition from occurring, however, RCS pressure must be reduced.

Page 197 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Technical Reference(s) SO23-14-5, Attachment 1, Step 3 Attached w/ Revision # See Comments / Reference Proposed references to be provided during examination: None Learning Objective: PREDICT and EXPLAIN the response of major plant systems, equipment and 54695 parameters to an excess steam demand event.

Question Source: Bank # 73369 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 Page 198 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SO23-14-5, Attachment 1, Step 3 Revision # 8 Page 199 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # CE-E06 EK3.03 Importance Rating 3.7 Loss of Main Feedwater: Knowledge of the reasons for the following responses as they apply to the Loss of Feedwater: Manipulation of controls required to obtain desired operating results during abnormal and emergency situations Proposed Question: Common 56 Given the following conditions:

• Feedwater restoration is in progress per SO23-12-6, Loss of Feedwater, Step 7, Establish AFW Flow to at least one Steam Generator following a loss of Main Feedwater.

• Direction is given to OVERRIDE and OPEN the available motor driven Auxiliary Feedwater Pump Discharge Bypass Valves 35%.

Which ONE (1) of the following is the reason for performing this action?

To establish a feedwater flow rate between...

E. 200 and 220 gpm. The lower value is based on twice refilling the Steam Generator feedring in a five-minute period. The higher value prevents damage due to excessive refill flow to a drained feedring.

F. 130 and 150 gpm. The lower value is based on twice refilling the Steam Generator feedring in a five-minute period. The higher value prevents damage due to excessive refill flow to a drained feedring.

G. 130 and 150 gpm. The lower value is based on collapsing a drained Steam Generator feedring. The higher value is based on overstressing the Steam Generator feedring J-tubes.

H. 200 and 220 gpm. The lower value is based on collapsing a drained Steam Generator feedring. The higher value is based on overstressing the Steam Generator feedring J-tubes.

Proposed Answer: B Page 200 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible because the reasons are correct, however, the range is 130 to 150 GPM.

B. Correct. This is the correct fill rate per the RNO column for Step 7g. These are the correct reasons per the EOI bases document.

C. Incorrect. Plausible because the feed rate is correct, however, the reasons are incorrect.

D. Incorrect. Plausible because one of the feed rates matches an old value used at SONGS (200 gpm) and collapsing a feedring is a concern, however, they do not match the EOI bases document.

Technical Reference(s) SO23-12-6, Step 7g Attached w/ Revision # See SO23-14-6, Step 7g Comments / Reference Proposed references to be provided during examination: None Learning Objective: DESCRIBE the cause/effect relationships associated with the following 55262 Auxiliary Feedwater System conditions/operations:

- EFFECT on Steam Generator levels of operation of the Auxiliary Feedwater Pump Discharge Bypass Valve.

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 Page 201 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SO23-12-6, Step 7g Revision # 21 Page 202 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SO23-14-6, Step 7g Revision # 8 Page 203 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 2 K/A # 001 AK3.01 Importance Rating 3.2 Continuous Rod Withdrawal: Knowledge of the reasons for the following responses as they apply to the Continuous Rod Withdrawal:

Manually driving rods into position that existed before start of casualty Proposed Question: Common 57 Given the following conditions at Unit 2 End-Of-Core conditions:

• Part-Length Control Element Assemblies (PLCEAs) were withdrawn 3 inches while performing Axial Shape Index control.

• When the Reactor Operator (RO) released the Manual Control Switch, the PLCEAs continued to move an additional three (3) inches.

• CEA movement stopped when the RO placed the CEDMCS Mode Selector Switch to OFF.

Which ONE (1) of the following is the reason for returning the PLCEAs to their Pre-Continuous Rod Withdrawal Incident position?

A. Minimize Local Power Density anomalies due to pellet clad interaction.

B. Avoid DNBR penalty factors inserted by the Core Protection Calculators.

C. Restore Tcold to within the Program Band per the Reload Ground Rules.

D. Preclude COLSS from generating an out-of-specification Azimuthal Tilt.

Proposed Answer: A Explanation:

A. Correct. CEA withdrawal or insertions are kept to less than 3 inches per minute to help control or minimize pellet clad interaction (PCI) per SO23-5-1.7 (Step 4.1). Because CEAs were moved 6 inches the chance of a high Local Power Density incurring PCI is greatly enhanced.

B. Incorrect. Plausible because CPC penalty factors could be generated, however, it would require a CEA deviation of 5 inches or greater for a single or multiple CEAs in a subgroup.

C. Incorrect. Plausible because Tcold should remain in the Program Band per the Reload Ground Rules, however, in this situation it is pellet clad interaction that is the concern.

D. Incorrect. Plausible because this is what COLSS calculates, however, the symmetry associated with the PLCEAs would preclude this from happening.

Page 204 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Technical Reference(s) SO23-5-1.7, Attachment 6 Attached w/ Revision # See SO23-3-2.13, Attachment 3 Comments / Reference SO23-3-2.21, L&S 4.11 & 4,12 SO23-5-1.7, L&S 1.7.2 Proposed references to be provided during examination: None Learning Objective: ANALYZE normal and abnormal operations of the Control Element Drive 81789 Mechanism Control System (CEDMCS).

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

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

Reference:

From SO23-5-1.7, Attachment 6 Revision # 38-1 Page 205 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SO23- 3-2.13, Attachment 3 Revision # 14-1 Comments /

Reference:

From SO23-3-2.21, L&S 4.11 & 4,12 Revision # 24 Comments /

Reference:

From SO23-5-1.7, L&S 1.7.2 Revision # 38-1 Page 206 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 2 K/A # 003 AK2.03 Importance Rating 3.1 Dropped Control Rod: Knowledge of the interrelations between the Dropped Control Rod and the following: Metroscope Proposed Question: Common 58 Which ONE (1) of the following indications is used to determine dropped CEA position?

A. Pulse Counter information from the Plant Computer System and Core Mimic Display (Rod Bottom light indication).

B. Reed Switch and Pulse Counter information from both the Plant Computer System and Core Mimic Display (Rod Bottom light indication).

C. Reed Switch and Pulse Counter information from both the Secondary Rod Position CRT Display and the Plant Computer System.

D. Reed Switch indication from the Secondary Rod Position CRT Display and Core Mimic Display (Rod Bottom light indication).

Proposed Answer: D Explanation:

A. Incorrect. Plausible because Pulse Counter is available from PCS and Reed Switch is available from Core Mimic Display, however, Pulse Counter does not update when a CEA is dropped.

B. Incorrect. Plausible because Reed Switch information is available from PCS and the Core Mimic Display is Reed Switch driven, however, Pulse Counter is only available from PCS.

C. Incorrect. Plausible because Reed Switch and Pulse Counter information are both available from PCS, however, only Reed Switch is available from Secondary Rod Position CRT Display.

D. Correct. This information is correct as shown.

Technical Reference(s) SD-SO23-510, Figure 1A Attached w/ Revision # See SD-SO23-510, Page 7 Comments / Reference Proposed references to be provided during examination: None Learning Objective: INTERPRET instrumentation and controls utilized in the Control Element Drive 81788 Mechanism Control System (CEDMCS).

Page 207 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Question Source: Bank #

Modified Bank # (Note changes or attach parent)

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

Reference:

From SD-SO23-510, Figure 1A Revision # 10 Comments /

Reference:

From SD-SO23-510, Page 7 Revision # 10 Page 208 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SD-SO23-510, Figure 11 Revision # 10 Page 209 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SD-SO23-510, Figure 12A Revision # 10 Page 210 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 2 K/A # 028 AK3.02 Importance Rating 2.9 Pressurizer Level Control Malfunction: Knowledge of the reasons for the following responses as they apply to the Pressurizer Level Control Malfunctions: Relationships between PZR pressure increase and reactor makeup / letdown imbalance Proposed Question: Common 59 Given the following conditions while at Hot Zero Power (HZP):

• Pressurizer Level is on program at 38%.

• Pressurizer Level Control Channel X is selected.

• Pressurizer Level Control Channel X failed to 35%.

• Pressurizer Pressure is 2260 psia and rising.

Which ONE (1) of the following is the reason for the response of the Reactor Coolant System?

Pressurizer Pressure is rising because...

A. one (1) Backup Charging Pump has started with Letdown flow at minimum.

B. both (2) Backup Charging Pumps have started with Letdown flow at minimum.

C. one (1) Backup Charging Pump has started with Letdown flow at normal.

D. both (2) Backup Charging Pumps have started with Letdown flow normal.

Proposed Answer: A Explanation:

A. Correct. At Hot Zero Power with Pressurizer Level on program, level is 38%. With level dropping 3% it will cause the start of one Backup Charging Pump (due to -2.5% deviation from program) and Letdown flow will go to minimum (due to -1.1% deviation from program).

B. Incorrect. Plausible because this action would occur if Pressurizer level drops below 34%.

C. Incorrect. Plausible because one Backup Charging Pump will start, however, level is far enough off program for Letdown to go to minimum.

D. Incorrect. Plausible because both Backup Charging Pumps would start with an additional 1% level deviation, however, Letdown flow would be at minimum.

Technical Reference(s) SO23-13-27, Attachments 2, 3, & 4 Attached w/ Revision # See Comments / Reference Page 211 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Proposed references to be provided during examination: None Learning Objective: DESCRIBE the Pressurizer Level Control System expected response to 55219 Pressurizer level transient and the required operator action if the expected plant response is not obtained in accordance with SO23-13-27.

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Page 212 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SO23-13-27, Attachment 2 Revision # 3 Page 213 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SO23-13-27, Attachment 4 Revision # 3 Page 214 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SO23-13-27, Attachment 3 Revision # 3 Page 215 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 2 K/A # 032 AA2.02 Importance Rating 3.6 Loss of Source Range Nuclear Instrumentation: Ability to determine and interpret the following as they apply to the Loss of Source Range Nuclear Instrumentation: Expected change in source range count rate when rods are moved Proposed Question: Common 60 Given the following conditions during a Reactor Startup:

• Excore Nuclear Instrument Startup Channel 1 has failed as is at 30 cps.

• Excore Nuclear Instrument Startup Channel 2 is reading 30 cps when the Regulating Group 1 CEAs are withdrawn from 0 to 90 inches.

• Keff is 0.980.

Which ONE (1) of the following is the expected count rate as read on Excore Nuclear Instrument Startup Channel 2?

A. 32 cps B. 34 cps C. 37 cps D. 40 cps Proposed Answer: B Explanation:

A. Incorrect. Plausible if graph is misread and it is thought that only 0.125% K/K was inserted.

B. Correct. 0 to 90 inches = 0.25% K/K therefore, CR2 / CR1 = 1-Keff1 / 1-Keff2 = 34 cps.

C. Incorrect. Plausible if wrong graph is read (HFP vice HZP) and it is thought that 0.37% K/K was inserted.

D. Incorrect. Plausible if math error is made and Keff2 is calculated at 0.985 vice 0.9825.

Technical Reference(s) OPS Physics Summary Figures 4.1 & 4.2 Attached w/ Revision # See Comments / Reference Proposed references to be provided during examination: OPS Physics Summary Figures 4.1 & 4.2 Learning Objective: IDENTIFY characteristics of the effective neutron multiplication factor, Keff, 54218 including:

- Reactor power (fission rate) response when its value is less than one, equal to one, or greater than one.

Page 216 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 1, 5 55.43 Page 217 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From OPS Physics Summary Figure 4.1 Revision # 55 Page 218 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From OPS Physics Summary Figure 4.2 Revision # 55 Page 219 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 2 K/A # 059 AK2.01 Importance Rating 2.7 Accidental Liquid Radwaste Release: Knowledge of the interrelations between the Accidental Liquid Radwaste Release and the following:

Radioactive liquid monitors Proposed Question: Common 61 Given the following conditions:

• An accidental radioactive liquid release is occurring from 2/3T-058, Radwaste Secondary Tank at 2 gpm due to leakage to the Circulating Water System Outfall.

Which ONE (1) of the following is the reason that Annunciator 61A09 - U2 & COMMON INSTRUMENT FAILURE for RT-7813, Liquid Radwaste Radiation Monitor would not be in alarm?

A. The minimum sensitivity of RT-7813, Liquid Radwaste Radiation Monitor requires greater than 30 gpm flow.

B. The low flow interlock has disabled the RT-7813, Liquid Radwaste Radiation Monitor alarm function.

C. The Data Acquisition System (DAS) high radiation alarm is set too high.

D. Less than 2 gpm flowrate will not maintain the Marineli Canister around the radiation detector full of sample fluid.

Proposed Answer: B Explanation:

A. Incorrect. Plausible because there is a misconception between the minimum sensitivity of RT-7813 and low sample flow. Additionally, the minimum flow is 60 gpm.

B. Correct. Flow through radiation monitor RT-7813 is dependent upon the differential pressure generated across flow orifice FO-7626. Low sample flow through RT-7813 is not enabled when the process discharge flow rate is less than 60 gpm. This corresponds to 2 gpm sample flow in RT-7813. If this occurred, the operator would still be able to monitor RT-7813 on the Data Acquisition System.

C. Incorrect. Plausible if the contents of the tank were exceptionally clean, however, annunciator 61A09 measures flow through the Radiation Monitor.

D. Incorrect. Plausible because it could be thought that this was the reason the alarm would not function, however, it is due to low sample flow.

Page 220 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Technical Reference(s) SO23-8-7, L&S 1.2 Attached w/ Revision # See SD-SO23-650, Figure 5 Comments / Reference SO23-15-61.A1.09 Proposed references to be provided during examination: None Learning Objective: EXPLAIN the precautions, limitations and administrative requirements 55859 associated with the operation of the Miscellaneous Waste System.

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 11, 13 55.43 Comments /

Reference:

From SO23-8-7, L&S 1.2 Revision # 17 Page 221 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SD-SO23-650, Figure 5 Revision # 12 Page 222 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SO23-15-61.A1.09 Revision # 1 Page 223 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 2 K/A # G 2.2.40 Importance Rating 3.4 Loss of Containment Integrity: Equipment Control: Ability to apply Technical Specifications for a system Proposed Question: Common 62 Which ONE (1) of the following represents a Loss of Containment Integrity as described in Technical Specifications during MODE 1?

A. While performing OPERABILITY tests on the redundant Containment Loop 1 Hot Leg Sample valves (HV-0508 and HV-0509), one of the two valves (HV-0508) fails to completely close.

B. The outer Containment Airlock Door fails its leak rate test and has to be opened to facilitate repairs that will take 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

C. A mechanic blocks open the outer Containment Airlock Door for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to perform maintenance activities on the INOPERABLE inner Containment door.

D. The Nitrogen Supply Isolation Valve to the Safety Injection Tanks Isolation Valve (HV-5434) fails its required stroke time.

Proposed Answer: C Explanation:

A. Incorrect. Plausible because if this would impact technical specification LCO 3.6.3, however, there is a redundant valve in the line that is OPERABLE.

B. Incorrect. Plausible because one Containment airlock door has failed, however, the other door is OPERABLE.

C. Correct. Per Technical Specification LCO 3.6.2.

D. Incorrect. Plausible because some valves such as the MSIVs must have OPERABLE stroke times per the FSAR, however, this particular valve does not meet that requirement.

Technical Reference(s) Unit 2 Technical Specifications LCO 3.6.2 Attached w/ Revision # See Unit 2 Technical Specifications LCO 3.6.3 Comments / Reference Unit 2 Tech Spec LCO 3.6.3 Bases Proposed references to be provided during examination: None Learning Objective: EXPLAIN the responsibility of the Operations Department for controlling 55076 containment access and integrity including:

- Technical Specification requirements and basis for Containment Integrity.

Page 224 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Question Source: Bank # 73339 Modified Bank # (Note changes or attach parent)

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

Reference:

From Unit 2 Technical Specifications LCO 3.6.2 Amendment # 127 Page 225 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From Unit 2 Technical Specifications LCO 3.6.3 Bases Amendment # 127 Page 226 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From Unit 2 Technical Specifications LCO 3.6.2 Amendment # 127 Page 227 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From Unit 2 Technical Specifications LCO 3.6.3 Amendment # 127 Page 228 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Page 229 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 2 K/A # 074 EA1.15 Importance Rating 3.9 Inadequate Core Cooling: Ability to operate and monitor the following as they apply to an Inadequate Core Cooling: Hot-leg and cold-leg temperature recorders Proposed Question: Common 63 Given the following conditions during a Large Break Loss of Coolant Accident:

• Critical Function Monitoring System (CFMS) is not available.

• Qualified Safety Parameter Display System (QSPDS) is not available.

• Both Thot Temperature Recorders (2TR-0111 and 2TR-0121) are pegged high at 625°F.

• Both Tcold Temperature Recorders (2TR-0115 and 2TR-0125) are reading 640°F.

• Both Loop Thot Wide Range Temperatures (2TI-0911X1 and 2TI-0911X2) are pegged high reading greater then 700°F.

• Both Loop Tcold Wide Range Temperatures (2TI-0911Y1 and 2TI-0911Y2) are reading 638°F.

• Pressurizer pressure is 500 psia.

Given the conditions listed, which ONE (1) of the following describes the condition of the Reactor Coolant System?

A. Adequate Core cooling exists as long as narrow range level is maintained in one (1)

Steam Generator.

B. Inadequate Core cooling exists whenever the implied temperature of the Core Exit Thermocouples is > 700ºF.

C. Adequate Core cooling exists as long as Reactor Vessel Plenum level is greater than 20%.

D. Inadequate Core cooling exists whenever reflux cooling is removing heat from the Reactor Coolant System.

Proposed Answer: B Page 230 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible because maintaining narrow range level in one Steam Generator will assure that reflux cooling was occurring, however, given the temperature in the Hot and Cold Legs, inadequate core cooling exists.

B. Correct. Given a T greater than 60°F, one would determine that natural circulation conditions do not exist. With Thot and Tcold temperature recorders reading as high as they are one could surmise that Core Exit Thermocouple temperatures have exceeded 700°F. This is an Inadequate Core Cooling condition. Additionally, one could calculate that superheat conditions exist in the RCS.

C. Incorrect. Plausible because this statement is true with respect to Plenum level, however, there is no way to determine what Plenum level is without QSPDS or CFMS and Core temps are > 700°F.

Additionally, Plenum level must be 41% for Adequate Core Cooling to exist.

D. Incorrect. Plausible because Inadequate Core Cooling does exist, however, the presence of reflux cooling does not automatically determine if Adequate (or Inadequate) Core Cooling exists.

Technical Reference(s) SO23-12-11, Attachment 2, FS-3 Attached w/ Revision # See SO23-14-10, Attachment 1, SF-8 Bases Comments / Reference SO23-14-3, Attachment 1 Proposed references to be provided during examination: None Learning Objective: PREDICT and EXPLAIN the response of major plant systems, equipment and 53006 parameters to a loss of forced circulation.

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 Page 231 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SO23-12-11, Attachment 2, FS-3 Revision # 6 Page 232 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SO23-14-10, Attachment 1, SF-8 Bases Revision # 2 Comments /

Reference:

From SO23-14-3, Attachment 1 Revision # 8 Page 233 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 2 K/A # 076 AA2.02 Importance Rating 2.8 High Reactor Coolant Activity: Ability to determine and interpret the following as they apply to the High Reactor Coolant Activity: Corrective actions required for high fission product activity in the RCS Proposed Question: Common 64 Given the following conditions:

• Unit 2 has high Reactor Coolant System activity and is in the process of cooling down on Shutdown Cooling.

• With Shutdown Cooling Purification System in service it is necessary to replace 2F-020, CVCS Purification Filter cartridge.

Which ONE (1) of the following is used to fill and vent the CVCS Purification Filter, 2F-020, following filter cartridge replacement?

A. Primary Plant Makeup Water with Purification flow diverted to Radwaste.

B. Nuclear Service Water with Purification flow aligned to the VCT.

C. Primary Plant Makeup Water with Purification flow aligned to the VCT.

D. Nuclear Service Water with Purification flow diverted to Radwaste.

Proposed Answer: D Explanation:

A. Incorrect. Plausible because the fluid can be diverted to Radwaste, however, Nuclear Service Water is the fluid used for filling and flushing.

B. Incorrect. Plausible because use of Nuclear Service Water is correct and it is possible to align to the VCT during normal operations, however, when placing the Shutdown Cooling Purification Loop in service the VCT is bypassed.

C. Incorrect. Plausible because it could be thought that since Primary Makeup Water is used for normal makeup that it would also be used for this evolution. Aligning to the VCT is procedurally allowed, however, not when placing the Shutdown Cooling Purification Loop in service.

D. Correct. Nuclear Service Water is used and it would be diverted to Radwaste. Refer to provided Shutdown Cooling Figure for general detail and CVCS Figure for greater detail.

Technical Reference(s) SO23-3-2.1, Attachment 7, Section 2.2 Attached w/ Revision # See SD-SO23-740, Figure 1 Comments / Reference SD-SO23-390, Figure 1 Page 234 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Proposed references to be provided during examination: None Learning Objective: ANALYZE normal and abnormal operations of the Nuclear Service Water and 81831 Primary Plant Make-Up Water System.

Question Source: Bank # 77244 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam 2000 NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 Comments /

Reference:

From SD-SO23-740, Figure 1 Revision # 17 Page 235 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SO23-3-2.1, Attachment 7, Section 2.2 Revision # 27 Page 236 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Page 237 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SD-SO23-390, Figure 1 Revision # 17 Page 238 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 2 K/A # CE-A13 G 2.4.31 Importance Rating 4.2 Natural Circulation: Emergency Procedures / Plan: Knowledge of annunciator alarms, indications, or response procedures Proposed Question: Common 65 A natural circulation cooldown is in progress. You have been assigned to perform SO23-12-11, EOI Supporting Attachments, Attachment 2, Floating Step 3, Monitor Natural Circulation Established.

The following parameters are observed:

• Pressurizer Pressure is 1050 psia and stable.

• Thot 540°F and slowly decreasing.

• Tcold 505°F and slowly decreasing.

• Pressurizer level 36% and stable.

• Reactor Vessel Head level is 48%.

• Annunciator 56B45 - RCS SUBCOOLED MARGIN LO is in alarm.

Which ONE (1) of the following action(s) should be taken in response to these conditions?

A. Continue cooldown at present rate as all criteria are satisfied.

B. Raise steaming and feeding rates to recover Natural Circulation criteria.

C. Energize Pressurizer Heaters to collapse Reactor Vessel Head voids to recover Natural Circulation criteria.

D. Lower steaming and feeding rates to recover Natural Circulation criteria.

Proposed Answer: B Page 239 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible because it could be thought that all Natural Circulation conditions are being met. Examinee must recognize that Annunciator 56B45 directly impacts Core Exit Saturation Margin (CESM) conditions despite the wording used.

B. Correct. Core Exit Saturation Margin conditions are not being met with this annunciator in alarm.

RNO actions require increasing steaming and feeding to recover Natural Circulation criteria.

C. Incorrect. Plausible because energizing Pressurizer Heaters would be a desirable condition to restore CESM, however, in this condition it is being used to collapse voids which are not part of the Natural Circulation criteria.

D. Incorrect. Plausible because lowering steaming and feeding rates could lower the rate of pressure reduction which would in turn restore CESM, however, this is not the correct action.

Technical Reference(s) SO23-15-56.B.45 Attached w/ Revision # See SO23-12-11, Attachment 2, FS-3 Comments / Reference Proposed references to be provided during examination: None Learning Objective: Per the EOI Attachments procedure, SO23-12-11, DESCRIBE:

55279 / 53429 - The CEN-152 basis or reason for these steps.

STATE how the operator actions in the EOIs promote or recover natural circulation/two phase heat removal processes.

Question Source: Bank #

Modified Bank # 137810 (Note changes or attach parent)

New Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 4, 5, 10 55.43 Comments /

Reference:

From SO23-15-56.B.45 Revision # 6 Page 240 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SO23-12-11, Attachment 2, FS-3 Revision # 6 Comments /

Reference:

From Exam Bank #137810 Revision 04/23/07 A natural circulation cooldown is in progress. You have been assigned to monitor natural circulation by initiating Floating Step 3 of SO23-12-11, Attachment 2.

The following parameters are observed:

• RCS Pressure1903 psia and stable

• T-hot 549oF and slowly decreasing

• T-cold 506oF and slowly decreasing

• REPCET 557oF and slowly decreasing

• Pressurizer level 36% and stable

• Reactor Vessel level 48% (RV Head)

What action(s) should be taken in response to these conditions?

A. Continue cooldown as directed. All natural circulation guidelines are satisfied.

B. Raise steaming and feeding rates to recover natural circulation criteria.

C. Lower steaming and feeding rates to recover natural circulation criteria.

D. Start all available charging pumps and initiate aux spray flow to collapse RV Head voids.

Page 241 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 3 Group # 1 K/A # G 2.1.14 Importance Rating 3.1 Conduct of Operations: Knowledge of criteria or conditions for plant wide announcements, such as pump starts, reactor trips, mode changes, etc.

Proposed Question: Common 66 Given the following conditions:

• Operations Personnel are standing by at each location listed.

Which ONE (1) of the following conditions would require a plant wide Public Address System announcement?

Starting a...

A. Condensate Pump 2P-051.

B. Emergency Diesel Generator 2G-002.

C. Heater Drain Pump 2P-058.

D. HPSI Pump 2P-017.

Proposed Answer: B Explanation:

A. Incorrect. Plausible because Control Room personnel would be notified, however, with an operator standing by, a plant wide PA announcement is not required. At one time, all 4 kV motor start announcements were made.

B. Correct. Even with several operators standing by in the Diesel Generator Room, this action is required per procedure anytime the Emergency Diesel Generator is started.

C. Incorrect. Plausible because Control Room personnel would be notified, however, with an operator standing by, a plant wide PA announcement is not required. At one time, all 4 kV motor start announcements were made.

D. Incorrect. Plausible because Control Room personnel would be notified, however, with an operator standing by, a plant wide PA announcement is not required. At one time, all 4 kV motor start announcements were made.

Technical Reference(s) SO23-3-3.23, Attachment 1 Attached w/ Revision # See SO123-0-A1, Step 6.2.4 Comments / Reference Proposed references to be provided during examination: None Page 242 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Learning Objective: EXPLAIN the importance of the Emergency Diesel Generator Mechanical 73242 Systems with regard to plant safety.

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

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

Reference:

From SO23-3-3.23, Attachment 1 Revision # 33 Page 243 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SO123-0-A1, Step 6.2.4 Revision # 14 Page 244 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 3 Group # 1 K/A # G 2.1.9 Importance Rating 2.9 Conduct of Operations: Ability to direct personnel activities inside the control room Proposed Question: Common 67 According to SO123-0-A3, Procedure Use, if an emergency occurs which is NOT covered by an approved Emergency Operating Instruction, then additional actions shall be approved by the:

A. Control Room Supervisor B. Plant Superintendent C. Operations Manager D. Shift Manager Proposed Answer: D Explanation:

A. Incorrect. Plausible because the CRS would do this for an AOI but not an EOI (shown procedurally as the SRO Operations Supervisor, this could be the CRS).

B. Incorrect. Plausible because it could be though that this individual must approve these types of issues but it is not procedurally required.

C. Incorrect. Plausible because the Operations Manager authorizes many administrative issues, however, they may not be in the Control Room and waiting for their response to an EOI issue could put the plant in jeopardy.

D. Correct. Per the procedure.

Technical Reference(s) SO123-0-A3, Section 6.7 Attached w/ Revision # See Comments / Reference Proposed references to be provided during examination: None Learning Objective: FOLLOW the steps of SO123-0-A3 to prepare an abnormal alignments and 184777 evolutions document to perform a plant manipulation not covered by a procedure.

Question Source: Bank # 75634 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam 2000 NRC Exam Page 245 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Comments /

Reference:

From SO123-0-A3, Section 6.7 Revision # 6 Comments /

Reference:

From SO23- Revision #

Page 246 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 3 Group # 2 K/A # G 2.2.14 Importance Rating 3.9 Equipment Control: Knowledge of the process for controlling equipment configuration or status Proposed Question: Common 68 Per SO123-0-A4, Configuration Control, which ONE (1) of the following describes the situation in which Peer Checking may be substituted for Independent Verification and what documentation is required?

A. Manipulation of switches in the Control Room.

Peer Checker may sign as Independent Verifier.

B. Manipulation of switches in the Control Room.

Independent Verification signature must be marked N/A.

C. Verification of locked valve position outside the Control Room.

Peer Checker may sign as Independent Verifier.

D. Verification of locked valve position outside the Control Room. Independent Verification signature must be marked N/A.

Proposed Answer: A Explanation:

A. Correct. Per the procedure when switches are manipulated in the Control Room the Peer Checker may sign as Independent Verifier.

B. Incorrect. Plausible because this is when Peer Checking is used, however, the Independent Verification signature is signed by the Peer Checker.

C. Incorrect. Plausible because the Peer Checker may sign as Independent Verifier, however, Peer Checking is not allowed for verifying locked valve positions.

D. Incorrect. Plausible because it could be thought that a Peer Checker could verify the position of a locked valve by observation, however, not only must the Independent Verification signature be signed but the individual would not be performing the position of a Peer Checker for this evolution.

Technical Reference(s) SO123-0-A4, 6.5.4 Attached w/ Revision # See Comments / Reference Proposed references to be provided during examination: None Learning Objective: Follow the steps of SO123-0-A4 to coordinate performance of system 191593 alignments.

Page 247 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Question Source: Bank # 127026 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam 2006 NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Comments /

Reference:

From SO123-0-A4, 6.5.4 Revision # 10 Page 248 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 3 Group # 2 K/A # G 2.2.38 Importance Rating 3.6 Equipment Control: Knowledge of conditions and limitations in the facility license Proposed Question: Common 69 SO23-3-3.25, Once a Shift Surveillance, requires verifying that Steam Generator level channels agree within 5% of each other.

Which ONE (1) of the following terms defines a qualitative assessment of the instrument channel's behavior during operation by visually comparing the indication to independent instrument channels measuring the same parameter?

A. Channel Functional Test B. Channel Verification C. Channel Calibration D. Channel Check Proposed Answer: D Explanation:

A. Incorrect. Plausible because a Channel Functional Test requires observation, however, it also requires insertion of a simulated signal.

B. Incorrect. Plausible because it sounds like a Channel Check, however, it is undefined with respect to Technical Specifications.

C. Incorrect. Plausible because a Channel Calibration requires observation, however, it also includes a Channel Functional Test as well as any adjustment required.

D. Correct. Per the definitions in Technical Specifications.

Technical Reference(s) Tech Specs Section 1.0 Definitions Attached w/ Revision # See Comments / Reference Proposed references to be provided during examination: None Learning Objective: EVALUATE plant status against Technical Specification requirements.

189963 / 55353 STATE the Technical Specification Definition for the following:

- Channel Check Page 249 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Question Source: Bank # 74073 Modified Bank # (Note changes or attach parent)

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

Reference:

From Unit 2 Technical Specifications Amendment # 127 Page 250 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From Unit 2 Technical Specifications Amendment # 127 Page 251 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 3 Group # 3 K/A # G 2.3.7 Importance Rating 3.5 Radiation Control: Ability to comply with the radiation work permit requirements during normal or abnormal conditions Proposed Question: Common 70 Along with a Radiation Exposure Permit (REP), which ONE (1) of the following is required to enter a High Radiation Area greater than 1000 mrem/hr for work of unknown duration?

E. Special REP Training.

F. Continuous HP Coverage.

G. Written permission from HP Supervision.

H. A respirator and special protective clothing.

Proposed Answer: B Explanation:

E. Incorrect. Plausible because it could be thought that this is required, however, there are no conditions in the Stem that would identify special REP training.

F. Correct. In lieu of a dose rate monitoring device or an alarming dosimeter, continuous HP coverage is required.

G. Incorrect. Plausible because it could be thought that this is required, however, HP supervision has already provided written permission if one has been able to sign on to an REP.

H. Incorrect. Plausible because this condition could be required, however, more information would be required in the Stem.

Technical Reference(s) SO123-VII-20, Section 6.10 Attached w/ Revision # See Comments / Reference Proposed references to be provided during examination: None Learning Objective: DEFINE a High Radiation Area.

172858 Question Source: Bank # 74608 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam Page 252 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 12 55.43 Comments /

Reference:

From SO123-VII-20, Section 6.10 Revision # 12-2 Page 253 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 3 Group # 3 K/A # G 2.3.14 Importance Rating 3.4 Radiation Control: Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or emergency conditions or activities Proposed Question: Common 71 Given the following conditions:

• Unit is operating normally at 100% power.

• NO radioactive releases are in progress.

• Radwaste Secondary Tank T-057 is leaking onto the floor from a vent line.

Which ONE (1) of the following radiation monitors will be the FIRST to indicate a high activity?

A. Liquid Waste Discharge Monitor RE-7813.

B. BPS Neutralization Sump Monitor RE-7817.

C. Plant Vent Stack Wide Range Monitor RE-7865.

D. Radwaste Condensate Return Monitor RE-7812.

Proposed Answer: C Explanation:

A. Incorrect. Plausible because this is in the normal flowpath alignment, however, flow is not through the discharge piping.

B. Incorrect. Plausible because this tank is located in Radwaste, however, the fluid would not flow to the BPS Neutralization Sump.

C. Correct. Activity would be seen in atmosphere and picked up through Plant Vent Stack discharge.

D. Incorrect. Plausible because Radwaste Condensate Return does alarm for various processes, however, tank does not drain to Radwaste in an area that would be picked up by the monitor.

Technical Reference(s) SD-SO23-690, Figures 2, 6, 12 & 13 Attached w/ Revision # See Comments / Reference Proposed references to be provided during examination: None Learning Objective: EXPLAIN the interfaces between the Radiation Monitoring System and other 103328 plant systems.

Page 254 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Question Source: Bank # 146209 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam 2006 NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 11, 13 55.43 Comments /

Reference:

From SD-SO23-690, Figure 2 Revision # 16 Page 255 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SD-SO23-690, Figure 6 Revision # 16 Comments /

Reference:

From SD-SO23-690, Figure 12 Revision # 16 Page 256 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SD-SO23-690, Figure 13 Revision # 16 Page 257 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 3 Group # 3 K/A # G 2.3.5 Importance Rating 2.9 Radiation Control: Ability to use radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc.

Proposed Question: Common 72 Which ONE (1) of the following radiation detectors should be selected when performing a body frisk for beta contamination and how should the frisk be performed?

Select the...

A. RO2 / RO2A with a single air filled Ion Chamber and slowly move the instrument over the frisking area, pausing when instrument response is detected.

B. ASP-1 with frisker with a hand-held Geiger-Mueller pancake probe and slowly move the instrument within 1/2 inch of the frisking area, pausing when instrument response is detected.

C. ASP-1 with AC3-7 with a hand-held ZnS (Ag) Scintillation probe and slowly move the instrument within 1/2 inch of the frisking area, pausing when instrument response is detected.

D. ASP-1 with NRD probe with a BF3 proportional counter and slowly move the instrument over the frisking area, pausing when instrument response is detected.

Proposed Answer: B Explanation:

A. Incorrect. Plausible because this instrument can detect low to medium beta-gamma radiation, however, it is too bulky to perform a frisk and the instrument should be within 1/2. Additionally, this could be used to frisk a swipe.

B. Correct. This is the correct instrument and method to use when performing a beta contamination frisk.

C. Incorrect. Plausible because this instrument reads out and counts per minute and the method is correct, however, it is used for detecting alpha radiation.

D. Incorrect. Plausible because the ASP-1 is also used, however, this instrument equipped with a BF3 proportional counter is used for detecting neutron radiation the instrument should be within 1/2.

Technical Reference(s) Lesson Plan 2LC898 HO, Slides 11 & 12 Attached w/ Revision # See Comments / Reference Proposed references to be provided during examination: None Page 258 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Learning Objective: USE portable survey instruments.

185456 Question Source: Bank #

Modified Bank # (Note changes or attach parent)

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

Reference:

From Lesson Plan 2LC898 HO, Slide 12 Revision # 1-1 Page 259 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From Lesson Plan 2LC898 HO, Slide 11 Revision # 1-1 Page 260 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 3 Group # 4 K/A # G 2.4.13 Importance Rating 4.0 Emergency Procedures / Plan: Knowledge of crew roles and responsibilities during EOP usage Proposed Question: Common 73 Which ONE (1) of the following identifies the Balance of Plant Operator (21 / 31) responsibilities during implementation of Emergency Operating Instructions (EOI)?

A. Assume the Plant Monitoring responsibility during performance of the Standard Post Trip Actions when not reporting to the CRS.

B. Relinquish Plant Monitoring responsibility once the immediate action steps of Emergency Operating Instructions are completed.

C. Coordinate the EOI Brief and ensures all instructions are understood and integrated into the shift plan along with priorities set by the CRS.

D. Delegate performance of specific EOI Attachments and reject the performance of Attachments that would prevent monitoring of Secondary Plant Safety Functions.

Proposed Answer: A Explanation:

A. Correct. Per SO123-0-A2, these are the Balance of Plant Operator responsibilities.

B. Incorrect. Plausible because it could be thought that Plant Monitoring responsibility is the assignment of the Control Operator (22/32), however, both individuals are responsible for Plant Monitoring.

C. Incorrect. Plausible because this individual coordinates the post-turnover brief, however, the EOI Brief is performed by the Control Room Supervisor.

D. Incorrect. Plausible because it could be thought that it is appropriate to continue monitoring Primary Plant Safety Functions, however, there are some Control Room emergency conditions where the 21 / 31 would perform actions on the secondary side of the plant.

Technical Reference(s) SO123-0-A2, Step 6.17.5 Attached w/ Revision # See SO123-0-A2, Step 6.18.4 Comments / Reference Proposed references to be provided during examination: None Learning Objective: DESCRIBE Control Board Operator's authority, duties, and responsibilities per 182045 SO123-0-A2.

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

Page 261 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Comments /

Reference:

From SO123-0-A2, Step 6.17.5 Revision # 9-1 Page 262 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SO123-0-A2, Step 6.18.4 Revision # 9-1 Page 263 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 3 Group # 4 K/A # G 2.4.39 Importance Rating 3.9 Emergency Procedures / Plan: Knowledge of RO responsibilities in emergency plan implementation Proposed Question: Common 74 Given the following conditions:

• The Control Room Supervisor has assigned you Operations Leader Duties.

Which ONE (1) of the following is an Operations Leader responsibility during Emergency Plan implementation?

A. Within 90 minutes of event declaration or as directed by the Emergency Coordinator, perform the Public Address & Coordination per Attachment 1.

B. Inform onsite personnel of emergency conditions, plant status, and on-going work related to the emergency using the Yellow Phone Network.

C. If an Unusual Event is declared, prohibit eating, drinking, and smoking in the Control Room until clearance has been given by Health Physics.

D. Within one (1) hour of an Alert or higher classification, activate the Emergency Response Data System to the NRC Operations Center.

Proposed Answer: D Explanation:

A. Incorrect. Plausible because as the Operations Leader this action must be performed, however, it must take place immediately (within 15 minutes).

B. Incorrect. Plausible because all of these conditions are performed, however, it is the Ivory Phone Network that is used for command and control. The Yellow Phone Network is used for incidental communications between groups.

C. Incorrect. Plausible because the Operations Leader is taxed with performing this action, however, only if a Site Area Emergency is declared.

D. Correct. Per the procedure, the ERDS is activated to the NRC Operations Center per SO23-3-2.32, Critical Functions Monitoring System.

Page 264 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Technical Reference(s) SO123-VIII-30, Step 6.1.12 Attached w/ Revision # See SO123-VIII-30, Step 6.1.1.1 Comments / Reference SO123-VIII-30, Step 6.2.1 & 6.2.1.1 SO123-VIII-30, Step 6.5.1 Proposed references to be provided during examination: None Learning Objective: OPERATE the Emergency Notification System.

154107 Question Source: Bank #

Modified Bank # (Note changes or attach parent)

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

Reference:

From SO123-VIII-30, Step 6.1.12 Revision # 14 Page 265 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SO123-VIII-30, Step 6.1.1.1 Revision # 14 Comments /

Reference:

From SO123-VIII-30, Step 6.2.1 & 6.2.1.1 Revision # 14 Page 266 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SO123-VIII-30, Step 6.5.1 Revision # 14 Page 267 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 3 Group # 4 K/A # G 2.4.17 Importance Rating 3.9 Emergency Procedures / Plan: Knowledge of EOP terms and definitions Proposed Question: Common 75 While performing the actions of SO23-12-3, Loss of Coolant Accident, how often are Safety Function Status checks required to be verified?

A. Continuously.

B. Every 15 minutes.

C. Every 30 minutes.

D. Every 60 minutes.

Proposed Answer: B Explanation:

A. Incorrect. Plausible because they are required to be performed continuously when in the Functional Recovery procedure.

B. Correct. Per the SFSC Basis Document.

C. Incorrect. Plausible because they are required to be performed continuously and be completed every 30 minutes when in the Functional Recovery procedure.

D. Incorrect. Plausible because it could be thought they are performed hourly since it usually takes 15 minutes to complete the first 11 steps of the SPTAs.

Technical Reference(s) SO23-14-10, Attachment 1 Attached w/ Revision # See Comments / Reference Proposed references to be provided during examination: None Learning Objective: PERFORM Safety Function Status Check for accident mitigation.

192229 Question Source: Bank # 127835 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X Page 268 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC RO Written Exam Worksheet Form ES-401-5 10 CFR Part 55 Content: 55.41 10 55.43 Comments /

Reference:

From SO23-14-10, Attachment 1, Step 4.4 Revision # 2 Page 269 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5

`

Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 009 EA2.39 Importance Rating 4.7 Small Break LOCA: Ability to determine or interpret the following as they apply to a Small Break LOCA: Adequate core cooling Proposed Question: SRO 76 Given the following conditions during a Small Break Loss of Coolant Accident:

• Pressurizer Level is 0%.

• Pressurizer Pressure is 900 psia.

• Representative Core Exit Thermocouple is 530°F.

• Reactor Vessel Level Monitoring System indicates 60% in the Reactor Vessel Plenum with three (3) detectors OPERABLE.

• Reactor Vessel Level Monitoring System indicates 0% in the Reactor Vessel Head with one (1) detector OPERABLE.

• SIAS / CIAS / CCAS / CSAS have all actuated.

• Natural Circulation has not been established.

Which ONE (1) of the following identifies the current condition of the Core and what action is required?

I. Adequate core cooling exists. Initiate SO23-12-11, EOI Supporting Attachments, Attachment 5, Core Exit Saturation Margin Control and raise Steam Generator levels and increase steaming rate to bring Core Exit Saturation Margin to > 20°F.

J. Inadequate core cooling exists. Transition to SO23-12-9, Functional Recovery, Attachment FR-5, Recovery - Heat Removal and initiate HR-2, S/G + ECCS actions.

K. Inadequate core cooling exists. Initiate SO23-12-11, EOI Supporting Attachments, Attachment 5, Core Exit Saturation Margin Control and raise Steam Generator levels and increase steaming rate to bring Core Exit Saturation Margin to > 20°F.

L. Adequate core cooling exists. Initiate SO23-12-11, EOI Supporting Attachments, Attachment 4, Alternate Reactor Vessel Level Determination and verify Core is covered.

Proposed Answer: A Page 270 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Explanation:

I. Correct. Even though Core Exit Saturation Margin is < 20°F adequate core cooling exists. In this condition actions of Attachment 5 are implemented.

J. Incorrect. Plausible because Core Exit Saturation Margin is < 20°F. This guidance is directed when REP CET temperature is > 700°F.

K. Incorrect. Plausible because the action directed is correct, however, adequate core cooling does exist because the correct number of OPERABLE detectors is available in the Reactor Vessel Head and Plenum and current indication is that the Core is covered.

L. Incorrect. Plausible because adequate core cooling does exist. It could be thought that the minimum number of OPERABLE detectors is not being met to adequately assess whether or not the Core is covered, however, there is a sufficient number available.

Technical Reference(s) SO23-12-11, Attachments 4 & 5 Attached w/ Revision # See SO23-12-10, Attachment 1, SF-8 Comments / Reference SO23-12-11, Floating Step 16 Proposed references to be provided during examination: Steam Tables Learning Objective: STATE how the operator actions in the EOIs promote or recover natural 53429 / 52757 circulation/two phase heat removal processes.

STATE the major recovery actions in response to a LOCA event.

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 5 Page 271 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SO23-12-10, Attachment 1, SF-8 Revision # 2 Page 272 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SO23-12-11, Attachment 5 Revision # 6 Page 273 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SO23-12-11, FS-16 Revision # 6 Page 274 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SO23-12-11, Attachment 4 Revision # 6 Page 275 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SO23-12-11, Attachment 4 Revision # 6 Page 276 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 011 EA2.03 Importance Rating 4.2 Large Break LOCA: Ability to determine or interpret the following as they apply to a Large Break LOCA: Consequences of managing LOCA with loss of CCW Proposed Question: SRO 77 Given the following conditions:

• A Large Break Loss of Coolant Accident is in progress on Unit 2.

• A total loss of Component Cooling Water has occurred on Unit 2.

• The Recirculation Actuation Signal will actuate in 5 minutes.

• Bus 2A04 has tripped and locked out due to a bus fault.

• HPSI Pump P-018 is currently aligned to Train B.

• Actions of SO23-12-3, Loss of Coolant Accident are in progress.

Which ONE (1) of the following actions should be taken to mitigate the situation?

A. Remain in SO23-12-3, Loss of Coolant Accident and cross-connect Unit 2 Component Cooling Water with Unit 3 and invoke 10CFR50.54.X on Unit 2.

B. Start P-019, HPSI Pump to Train B following the Recirculation Actuation Signal in order to provide additional flow and transition to SO23-12-9, Functional Recovery, Attachment FR-5, Recovery - Heat Removal if HPSI Pump performance becomes unstable.

C. Transition to SO23-12-9, Functional Recovery, Attachment FR-5, Recovery - Heat Removal and cross-connect Unit 2 Component Cooling Water with Unit 3 and invoke 10CFR50.54.X on Unit 3.

D. Perform actions of SO23-12-11, EOI Supporting Attachments, Attachment 14, RAS Operations to raise RWST level in order to flood Containment above the 23 level to improve Net Positive Suction Head to the operating HPSI Pump.

Proposed Answer: A Page 277 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Explanation:

A. Correct. Given the conditions listed, remaining in SO23-12-3 is appropriate and cross connecting CCW places Unit 2 in a 50.54.X notification.

B. Incorrect. Plausible because the pump is available (P-018, the 3rd of a kind Pump will start on an SIAS, therefore, P-019 is available) and might be considered, however, according to Step 5 EOI Bases, it will only increase flow marginally, if at all, and one operating train is sufficient at his time.

Transitioning to the FR is plausible as this is the RNO action in Floating Step 22 if HPSI Pump flow becomes unstable.

C. Incorrect. Plausible because it could be thought that the FR is the procedure required for this condition, however, one Train is all that is required given the conditions listed and cross connecting with Unit 3 CCW places Unit 2 in a 50.54.X notification.

D. Incorrect. Plausible because refilling the RWST would be a desired action given the loss of CCW and raising level does improve HPSI Pump NPSH, however, flooding above the 225 level will impact the Emergency Cooling Unit ductwork and could complicate the loss of CCW that already exists.

Technical Reference(s) SO23-12-3, Step 5 Attached w/ Revision # See SO23-14-3, Step 5 Bases Comments / Reference SO23-12-11, FS-22 SO23-13-7, Attachment 4 SO23-12-11, Attachment 14 Proposed references to be provided during examination: None Learning Objective: STATE the major recovery actions in response to a LOCA event.

52757 / 53682 EVAULATE Component Cooling Water System conditions against Administrative and Technical Specification requirements and determine what action, if any, is required.

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 2, 5 Page 278 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SO23-12-3, Step 5 Revision # 20 Comments /

Reference:

From SO23-14-3, Step 5 Bases Revision # 8 Page 279 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SO23-14-3, Step 5 Bases Revision # 8 Comments /

Reference:

From SO23-13-7, Attachment 4 Revision # 13 Page 280 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SO23-12-11, FS-22 Revision # 6 Page 281 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SO23-12-11, Attachment 14 Revision # 6 Page 282 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 057 G 2.1.20 Importance Rating 4.6 Loss of Vital AC Instrument Bus: Conduct of Operations: Ability to interpret and execute procedure steps Proposed Question: SRO 78 Given the following Unit 2 conditions:

• Channel A Reactor Coolant System Hot Leg temperature TT-0122-1 has failed and the affected Functional Units are in BYPASS.

• Channel C Reactor Coolant System Hot Leg temperature TT-0122-3 has also failed and the affected Functional Units have been placed in TRIP.

• A loss of Vital Bus Inverter 2Y-004 occurs and the Reactor does NOT trip.

• All other plant conditions are normal.

Which ONE (1) of the following actions is required?

A. Initiate a Plant Shutdown per SO23-5-1.7, Power Operations.

B. Place Channel D RCS Hot Leg temperature TT-0122-4 Functional Units in BYPASS.

C. Trip the Reactor and enter SO23-12-1, Standard Post Trip Actions.

D. Place Vital Bus 2Y-04 on its Alternate Source within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Proposed Answer: C Explanation:

A. Incorrect. Plausible because it could be thought that only a Plant Shutdown is required, however, this condition requires a Reactor trip.

B. Incorrect. Plausible because this is required by Technical Specifications and the AOI, however, with one other channel in BYPASS and another in TRIP the Reactor should have tripped.

C. Correct. With one Functional Unit in BYPASS and the other in TRIP, it places the Reactor Protection System coincidence logic in a 1 out of 2 configuration. Given the third trip signal which is generated on loss of the Vital Bus, a Reactor trip is required.

D. Incorrect. Plausible because the Vital Inverter must be restored to OPERABLE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, however, the Alternate Source is required to be aligned within two hours per Technical Specification LCO 3.8.7.

Page 283 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Technical Reference(s) SO23-13-18, Step 5 Attached w/ Revision # See SO23-13-18, Attachment 5 Comments / Reference SO23-13-18, Step 2 Technical Specification LCO 3.8.7 Technical Specifications LCO 3.3.1 Proposed references to be provided during examination: None Learning Objective: DESCRIBE the expected response of the associated components, instruments 55254 and controls for a malfunction or misoperation of one or more of the following Plant Protection System components:

- Input Sensors Question Source: Bank # 126550 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 5 Page 284 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SO23-13-18, Step 5 Revision # 8 Page 285 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SO23-13-18, Attachment 5 Revision # 8 Comments /

Reference:

From SO23-13-18, Step 2 Revision # 8 Page 286 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From Technical Specification LCO 3.8.7 Amendment # 127 Comments /

Reference:

From Technical Specification LCO 3.3.1 Amendment # 127 Page 287 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 027 AA2.16 Importance Rating 3.9 Pressurizer Pressure Control System Malfunction: Ability to determine and interpret the following as they apply to the Pressurizer Pressure Control Malfunctions: Actions to be taken if pressurizer pressure instrument fails low Proposed Question: SRO 79 Given the following Unit 2 conditions at 10% power:

• PT-0100X, Pressurizer pressure transmitter has failed low concurrent with a loss of Bus 2B06.

• Pressurizer pressure is 2280 psia and slowly rising.

• Pressurizer Spray Valves have not responded in AUTOMATIC.

Which ONE (1) of the following actions is required?

A. Direct placing Pressurizer Channel Y in service and restore Pressurizer pressure to less than 2275 psia within four (4) hours per Technical Specification LCO 3.4.1, RCS DNB Limits.

B. Determine that Pressurizer pressure is not controlled, trip the Reactor and restore Pressurizer pressure to less than 2275 psia within four (4) hours per Technical Specification LCO 3.4.1, RCS DNB Limits.

C. Direct placing Pressurizer Channel Y in service and restore Pressurizer heaters to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> per Technical Specification LCO 3.4.9, Pressurizer.

D. Determine that Pressurizer pressure is not controlled, trip the Reactor and restore Pressurizer heaters to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> per Technical Specification LCO 3.4.9, Pressurizer.

Proposed Answer: C Page 288 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible because this action would is required, however, the Tech Spec time is 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

B. Incorrect. Plausible because this action would be required if pressure continued to rise.

Additionally, the SRO should direct manual operation of the Spray Valves before determining that pressure is not controlled.

C. Correct. Per SO23-13-27, Step 3a. Additionally, with Bus 2B06 deenergized, one Train of 1E Heaters is INOPERABLE and requires the Tech Spec ACTION listed.

D. Incorrect. Plausible because this would be the required action if pressure is not controlled, however, procedurally the SRO should direct operation of the Spray Valves in MANUAL before determining that pressure is not controlled. The Tech Spec call is correct.

Technical Reference(s) SO23-13-27, Step 3a & 3f Attached w/ Revision # See Technical Specification LCOs 3.4.1 & 3.4.9 Comments / Reference SD-SO23-360, Figure III-10 Proposed references to be provided during examination: None Learning Objective: DESCRIBE the Technical Specification operability requirements and action 56422 statements associated with the Pressurizer Pressure and Level Control System.

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 2, 5 Page 289 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SO23-13-27, Step 3a Revision # 3 Comments /

Reference:

From SO23-13-27, Step 3f Revision # 3 Page 290 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From Technical Specification LCO 3.4.1 Amendment # 149 Page 291 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From Technical Specification LCO 3.4.9 Amendment # 161 Page 292 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SD-SO23-360, Figure III-10 Revision # 17 Page 293 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 029 G 2.4.11 Importance Rating 4.2 ATWS: Emergency Procedures / Plan: Knowledge of abnormal condition procedures Proposed Question: SRO 80 Given the following conditions:

• An Anticipated Transient Without Scram has occurred on Unit 3.

• Reactor power indicates approximately 20%.

• An Emergency Boration has been initiated.

Which ONE (1) of the following parameters is used to verify Emergency Boration delivery to the RCS and what Notification is required per SO23-13-11, Emergency Boration / Inadvertent Dilution or Boration?

A. Reactor Coolant System temperature. Initiate a one (1) hour report to the NRC per SO123-0-A7, Notification and Reporting of Significant Events.

B. Boric Acid Pump flow. Initiate a one (1) hour report to the NRC per SO123-0-A7, Notification and Reporting of Significant Events C. Reactor Coolant System temperature. Initiate a four (4) hour report to the NRC per SO123-0-A7, Notification and Reporting of Significant Events.

D. Boric Acid Pump flow. Initiate a four (4) hour report to the NRC per SO123-0-A7, Notification and Reporting of Significant Events.

Proposed Answer: A Explanation:

A. Correct. With the unit still at power, monitoring Reactor Coolant System temperature reduction is the means by which boron delivery is assured. This event requires a one hour report to the NRC due to an Immediate Shutdown Action.

B. Incorrect. Plausible because the reporting requirement time is correct. Boric acid pump flow is a component of an Emergency Boration, however, there is no reactivity response verification associated with this parameter.

C. Incorrect. Plausible because the parameter identified is correct, however, this event requires a one hour notification to the NRC. A four hour report is plausible because this action is required for Reactor Protection System activation when the Reactor is critical. It could be thought that the RPS activated, however, it was a Reactor Trip Breakers that failed to open.

D. Incorrect. Plausible because Boric acid pump flow is a component of an Emergency Boration, however, there is no reactivity response verification associated with this parameter.

Page 294 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Technical Reference(s) S023-13-11, Step 2j & 9 Attached w/ Revision # See SO123-0-A7, Attachment 1 Comments / Reference Proposed references to be provided during examination: None Learning Objective: ANALYZE normal and abnormal operations of the Chemical and Volume 102551 / 192858 Control System.

REPORT significant events to the Nuclear Regulatory Commission.

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

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

Reference:

From S023-13-11, Step 2j Revision # 12 Comments /

Reference:

From S023-13-11, Step 9 Revision # 12 Page 295 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SO123-0-A7, Attachment 1 Revision # 7 Comments /

Reference:

From SO123-0-A7, Attachment 1 Revision # 7 Page 296 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # G 2.4.45 Importance Rating 4.3 Reactor Trip / Stabilization / Recovery: Ability to prioritize and interpret the significance of each annunciator or alarm.

Proposed Question: SRO 81 Given the following conditions:

• Following a Reactor trip, a Reactor Trip Override (RTO) malfunction has caused an excessive cooldown.

• Manual control of Feedwater has stopped the overfeed condition.

• The following annunciators are now in alarm:

• 56A51 - SG2 E088 PRESS LO PRETRIP

• 56A53 - SG1 E089 PRESS LO PRETRIP

• Steam Generator pressure is lowering.

Which ONE (1) of the following actions should be taken by the Control Room Supervisor?

A. Declare a cooldown in progress and reset the MSIS setpoints and direct the Shift Technical Advisor to continue the Safety Function Status Checks.

B. Close the Main Steam Isolation Valves and continue monitoring SG pressure while initiating a two-prong attack.

C. Stop the cooldown by manual isolation or initiation of MSIS and redirect the Shift Technical Advisor to restart the Safety Function Status Checks.

D. Verify the Main Steam Safeties are closed, reset the MSIS setpoints and direct the Shift Technical Advisor to perform the RCS Heat Removal Safety Function.

Proposed Answer: C Explanation:

A. Incorrect. Plausible because resetting the setpoints would prevent an MSIS, however, an MSIS needs to be initiated. The Shift Technical Advisor action is appropriate for these conditions.

B. Incorrect. Plausible because closing the MSIVs is desirable, however, an MSIS should be initiated.

A two-pronged attack is used when the Functional Recovery procedure is implemented.

C. Correct. This is the desired action because the cooldown must be stopped or MSIS must be initiated. Having the STA restart Safety Function Status Checks would allow for a re-diagnosis.

D. Incorrect. Plausible because this action is desirable to determine the source of the cooldown, however, MSIS should be initiated. Additionally, the STA action is not per the guidance of OSM-9.

Page 297 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Technical Reference(s) SO23-12-1, Step 8 RNO Attached w/ Revision # See SO23-15-56.A.51 Comments / Reference OSM-9, Step 22 Proposed references to be provided during examination: None Learning Objective: DIRECT shift personnel actions to ensure plant safety during emergency, 192863 / 183756 abnormal, or off-normal conditions.

DIRECT response to and recovery from an Excess Steam Demand Event.

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

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

Reference:

From SO23-15-56.A.51 (56A53 same) Revision # 6 Page 298 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SO23-12-1, Step 8 RNO Revision # 21 Comments /

Reference:

From OSM-9, Step 22 Revision # 5 Page 299 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 2 K/A # 024 G 2.4.4 Importance Rating 4.7 Emergency Boration: Emergency Procedures / Plan: Ability to recognize abnormal indications for system operating parameters that are entry-level conditions for emergency and abnormal operating procedures Proposed Question: SRO 82 Given the following conditions during MODE 6:

• Unit 3 is in a Reduced Inventory Condition.

• Shutdown Cooling Purification was placed in service two (2) hours ago.

• The following annunciator goes into alarm:

• 58A09 - STARTUP CHANNEL 1 COUNTRATE HI / LO.

• The Boronometer on CR-50 is reading 2395 ppm and was verified by Chemistry sample.

Which ONE (1) of the following is required for the listed conditions?

A. Direct performance of SO23-13-11, Emergency Boration of the RCS / Inadvertent Dilution or Boration, and commence an Emergency Boration. Verify Boron concentration is within the limits specified in the COLR within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

B. Contact Reactor Engineering and request a more restrictive setpoint per SO23-3-2.15, Excore Instrumentation Operation, Section for Startup Channel Operation During Shutdown Modes.

C. Direct performance of SO23-13-11, Emergency Boration of the RCS / Inadvertent Dilution or Boration, actions for an inadvertent dilution and immediately bypass CVCS demineralizers.

D. Verify Startup Channel 2 indications are consistent with plant conditions and refer to SO23-3-2.15, Excore Instrumentation Operation, Section for Failure of a Single Startup Channel.

Proposed Answer: A Page 300 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Explanation:

A. Correct. Given the conditions listed, the SDC Purification loop has diluted the RCS below the minimum Refueling Boron concentration. SO23-13-11 Entry Conditions require an emergency boration. Technical Specifications require the associated Surveillance.

B. Incorrect. Plausible because this Note is at the bottom of the page for Annunciator 58A09 and it could be thought that 58A10 - Startup Channel 2 Countrate Hi / Lo should also be in alarm.

C. Incorrect. Plausible because it should be recognized that the SDC Purification loop has diluted the RCS and the actions in SO23-13-11, Step 5 will require bypassing of the demineralizers, however, emergency boration is required.

D. Incorrect. Plausible because Channel 2 is not in alarm and this is the required action if a Startup Channel 1 Trouble alarm (50A47) was received. It could be thought that the channel has failed and this would be an appropriate action.

Technical Reference(s) Technical Specification LCO SR 3.9.1.1 Attached w/ Revision # See SO23-15-58.A.09 Comments / Reference SO23-13-11, Entry Conditions SO23-3-2.15, Section 6.2 SO23-13-11, Step 5e SO23-15-50.A.47 Proposed references to be provided during examination: None Learning Objective: DESCRIBE the operation of Excore Nuclear Instrumentation System alarms, 56473 / 52685 including setpoint, possible causes and effects on the systems and overall plant operation.

DESCRIBE the Technical Specification operability requirements and action statements associated with the Excore Nuclear Instrumentation System.

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 2, 5, 6 Comments /

Reference:

From Technical Specification LCO SR 3.9.1.1 Amendment # 127 Page 301 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Page 302 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SO23-15-58.A.09 Revision # 9 Page 303 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SO23-13-11, Entry Conditions Revision # 12 Comments /

Reference:

From SO23-3-2.15, Section 6.2 Revision # 13 Page 304 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SO23-13-11, Step 5e Revision # 12 Comments /

Reference:

From SO23-15-50.A.47 Revision # 11 Page 305 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 2 K/A # 032 AA2.07 Importance Rating 3.4 Loss of Source Range Nuclear Instrumentation: Ability to determine and interpret the following as they apply to the Loss of Source Range Nuclear Instrumentation: Maximum allowable channel disagreement Proposed Question: SRO 83 Given the following conditions:

• Unit 3 Reactor Coolant System Tcold is 175°F with Shutdown Cooling in service.

• I&C reports the high voltage supply to Log Power Safety Channel JI-001-3 is out-of-specification low.

• The Reactor Trip Circuit Breakers are open.

Which ONE (1) of the following identifies this Channels OPERABILITY requirements and what is the maximum allowable channel disagreement between the other three (3) Channels?

Log Power Safety Channel JI-001-3 is...

A. required to be OPERABLE in this condition. Verify the other three (3) channels agree within the one and one-half (11/2) decades of each other.

B. NOT required to be OPERABLE in this condition. Verify the other three (3) channels agree within the one and one-half (11/2) decades of each other.

C. required to be OPERABLE in this condition. Verify the other three (3) channels agree within one (1) decade of each other.

D. NOT required to be OPERABLE in this condition. Verify the other three (3) channels agree within one (1) decade of each other.

Proposed Answer: D Page 306 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible because the channel would be required to be OPERABLE with the Reactor Trip Circuit Breakers closed and the other channels must agree within the 1 decade of each other.

B. Incorrect. Plausible because the channel is not required to be OPERABLE, however, the other channels must agree within one decade of each other.

C. Incorrect. Plausible because the 1 decade requirement is for MODE 5 operations, however, the channel would be required to be OPERABLE with the Reactor Trip Circuit Breakers closed.

D. Correct. The 1 decade requirement is applicable for MODE 5 operations. Per Technical Specifications, the channel is not required to be OPERABLE as long as the Reactor Trip Circuit Breakers are open while in MODES 3, 4, and 5.

Technical Reference(s) SO23-3-3.2, Attachment 4 Attached w/ Revision # See Unit 3 Tech Specs LCO 3.3.1 Bases Comments / Reference Unit 3 Technical Specification LCO 3.3.1 Proposed references to be provided during examination: None Learning Objective: DESCRIBE the Technical Specification operability requirements and action 52685 statements associated with the Excore Nuclear Instrumentation System.

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 2, 5 Page 307 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From Unit 3 Tech Specs LCO 3.3.1 Bases Amendment # 116 Comments /

Reference:

From SO23- SO23-3-3.2, Attachment 4 Revision # 14 Page 308 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From Unit 3 Technical Specification LCO 3.3.1 Amendment # 142 Page 309 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 2 K/A # 037 G 2.1.45 Importance Rating 4.3 Steam Generator Tube Leak: Emergency Procedures / Plan: Ability to prioritize and interpret the significance of each annunciator or alarm Proposed Question: SRO 84 Given the following conditions:

• 2RT-7870 Condenser Air Ejector Radiation Monitor setpoint is 1.05E-0 µci / sec and the Pre-Determined Steam Generator Tube leakrate is 25 gpd.

• Annunciator 60A46 - SECONDARY RADIATION HI has just alarmed.

• 15 minutes following 2RT-7870 HI alarm, the reading on the Data Acquisition System (DAS) has risen to 1.814E+3 µci / sec.

In addition to implementing the Attachment for Minimizing Contamination During a Steam Generator Tube Leak, which ONE (1) of the following actions is required based on the change in leak rate?

A. Commence a controlled shutdown to be in HOT STANDBY within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> per SO23-5-1.7, Section for Power Descension.

B. Manually trip the Reactor and perform actions of SO23-12-1, Standard Post Trip Actions.

C. Commence a rapid shutdown per SO23-5-1.7, Section for Guidelines for Rapid Power Reduction and trip the Reactor when less than 35% power.

D. Commence a controlled shutdown to be in HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> per SO23-5-1.7, Section for Power Descension.

Proposed Answer: C Explanation:

A. Incorrect. Plausible because the Unit must be shutdown, however, the leak rate has increased to greater than 75 gpd and is increasing by greater than 30 gpd.

B. Incorrect. Plausible because the Unit must be shutdown and Standard Post Trip Actions performed, however, only if the leak rate could not be maintained with the Charging Pumps.

C. Correct. Because the leak rate has gone from 25 gpd to ~30 gpm, a Rapid Power Reduction must be performed per Step 4c of SO23-13-15.

D. Incorrect. Plausible because the Unit must be shutdown and leakage is greater then 150 gpd, however, the leak rate is increasing by greater than 30 gpd.

Page 310 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Technical Reference(s) SO23-15-60.A2.46 Attached w/ Revision # See SO23-13-14, Steps 4c & 4p Comments / Reference SO23-13-14, Attachment 1 Proposed references to be provided during examination: SO23-13-14, Attachment 1 Learning Objective: Per the Reactor Coolant Leak procedure, SO23-13-14, DESCRIBE:

54932 - The basis for each step, caution, or note.

- The expected plant response for each step.

Question Source: Bank # 72675 Modified Bank # (Note changes or attach parent)

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

Reference:

From SO23-13-14, Step 4c Revision # 11 Page 311 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SO23-15-60.A2.46 Revision # 14 Page 312 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SO23-13-14, Step 4p Revision # 11 Page 313 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SO23-13-14, Attachment 1 Revision # 11 Page 314 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SO23-13-14, Attachment 1 Revision # 11 Page 315 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SO23-13-14, Attachment 1 Revision # 11 Page 316 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 2 K/A # 005 AA2.01 Importance Rating 4.1 Inoperable / Stuck Control Rod: Ability to determine or interpret the following as they apply to the Inoperable / Stuck Control Rod: Stuck or inoperable rod from in-core and ex-core NIS, in-core or loop temperature measurements Proposed Question: SRO 85 Given the following conditions during a Unit 2 power ascension from 90% power:

• Annunciator 50A28 - CEA DEVIATION has just alarmed.

• The crew determines that there is one (1) stuck Group 6 CEA with an 8 deviation.

• Annunciator 50A02 - COLSS ALARM is also in alarm.

• COLSS Azimuthal Tilt is reading 0.04 and stable.

• I&C Technician reports that the Group 6 CEA is stuck but trippable.

Which ONE (1) of the following actions is required?

A. Initiate SO23-13-13, Misaligned or Immovable Control Element Assembly, Attachment 2, Azimuthal Power Tilt Monitoring and reduce power below 50% within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

B. Initiate SO23-13-13, Misaligned or Immovable Control Element Assembly, and adjust the Azimuthal Power Tilt allowance in the Core Protection Calculators within two (2) hours.

C. Reduce the Azimuthal Power Tilt to less than 0.03 within two (2) hours or initiate a normal Plant Shutdown per SO23-5-1.7, Power Operations and be in MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

D. Initiate SO23-3-2.19, CEDMCS Operation, Section 6.7, Aligning a Misaligned CEA with its Group and reduce the Azimuthal Power Tilt to less than 0.03 within two (2) hours.

Proposed Answer: B Page 317 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible because these actions would be performed and are required by Technical Specifications, however, only if Azimuthal Tilt had exceeded 0.10%.

B. Correct. This is the ACTION required by Technical Specifications.

C. Incorrect. Plausible because this action is required per SO23-13-13, Step 1h RNO if it is determined that the CEA is not electrically movable or trippable.

D. Incorrect. Plausible because this action is directed by the annunciator response procedure for CEA DEVIATION, however, the deviation would have to be less than or equal to 7 inches. Additionally, the Limitations and Specifics in SO23-3-2.19 specify that CEA realignments should be performed per the AOI to avoid possible fuel damage.

Technical Reference(s) SO23-15-50.A1.28 Attached w/ Revision # See SO23-13-13, Step 4c RNO Comments / Reference SO23-13-13, Step 1h RNO Technical Specification LCO 3.2.3 SO23-12-13, Attachment 2 SO23-3-2.19, L&S 6.3 & 6.5 Proposed references to be provided during examination: None Learning Objective: Per the Misaligned CEA procedure, SO23-13-13, DESCRIBE:

54879 / 54876 - The basis for each step, caution, or note.

- The expected plant response for each step.

DETERMINE Technical Specifications that are impacted during a Misaligned CEA event.

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 2, 5 Page 318 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 him a Comments /

Reference:

From SO23-15-50.A1.28 Revision # 7 Page 319 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SO23-13-13, Step 4c RNO Revision # 11-2 Page 320 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SO23-13-13, Step 1h RNO Revision # 11-2 Page 321 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From Technical Specification LCO 3.2.3 Amendment # 127 Page 322 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SO23-12-13, Attachment 2 Revision # 11-2 Comments /

Reference:

From SO23-3-2.19, L&S 6.3 & 6.5 Revision # 20 Comments /

Reference:

From SO23-4-2.19, Section 6.7 Revision # 20 Page 323 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 026 G 2.2.46 Importance Rating 4.2 Containment Spray: Emergency Procedures / Plan: Ability to verify that alarms are consistent with the plant conditions Proposed Question: SRO 86 Given the following conditions during a Loss of Coolant Accident:

• The following annunciators are in alarm:

• 57C10 - CONTAINMENT RADIATION HI

• 56A07 - CONTAINMENT PRESS HI ESFAS CHANNEL TRIP

• 57C19 - CONTAINMENT HYDROGEN LEVEL HI

• The following annunciator is NOT in alarm:

• 56A08 - CONTAINMENT PRESS HI - HI ESFAS CHANNEL TRIP

• Containment radiation levels are 65 R/HR.

• Containment temperature is 130°F.

• Reactor Coolant System pressure is 900 psia.

Which ONE (1) of the following actions should be considered?

Evaluate...

A. initiating Containment Spray to strip Iodine from the Containment atmosphere and minimize potential for exceeding thyroid dose at the Exclusion Area Boundary.

B. securing of Containment Cooling to allow Containment humidity levels to rise and shift away from detonation / flammability area curve.

C. performing a normal purging of Containment to reduce the concentration of hydrogen to less than 5%.

D. securing Dome Air Circulators to allow Containment hydrogen to rise and concentrate away from electrical equipment.

Proposed Answer: A Page 324 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Explanation:

A. Correct. There is no Containment Spray signal at this time (otherwise, the HI - HI alarm would be in) and radiation level is > 40 R/HR. The LOCA EOI Bases and the EOI RNO recommends initiating Containment Spray to strip the atmosphere of iodine and reduce the potential for exceeding cumulative thyroid dose at the Exclusion Area Boundary.

B. Incorrect. Plausible because raising the moisture content in Containment will shift from a detonation / flammability area into the non-explosive region.

C. Incorrect. Plausible because there is procedural guidance for purging containment with a LOCA in progress, however, given current RCS pressure a normal purge would violate MODES 1 through 4 Purge Valve limitations. Additionally, there has been no accounting for Iodine levels in Containment and the H2 alarm comes in at 3% not 5%.

D. Incorrect. Plausible because this action will concentrate hydrogen at the top of Containment, however, this creates ideal conditions for a hydrogen burn.

Technical Reference(s) SO23-12-3, Step 17 & 18 Attached w/ Revision # See SO23-14-3, Attachment 1, Step 20 Comments / Reference SO23-57.C.10 & 19 SO23-15-56.A.07 & 08 SO23-3-2.28, Section 6.4 SO23-3-2.28, Attachment 1 Proposed references to be provided during examination: None Learning Objective: PREDICT and EXPLAIN the response of containment instrumentation to 55326 / 91482 adverse containment conditions during a Loss of Coolant Accident.

Per the LOCA procedure SO23-12-3, DESCRIBE the basis for each step, caution or note.

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 4, 5 Comments /

Reference:

From SO23-12-3, Step 18 Revision # 20 Page 325 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Page 326 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SO23-14-3, Attachment 1, Step 20 Revision # 6 Comments /

Reference:

From SO23-12-3, Step 17 Revision # 20 Page 327 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SO23-57.C.10 Revision # 17 Page 328 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SO23-15-56.A.07 Revision # 6 Comments /

Reference:

From SO23-15-56.A.08 Revision # 6 Page 329 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SO23-15-57.C.19 Revision # 17 Page 330 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SO23-3-2.28, Section 6.4 Revision # 14-1 Page 331 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SO23-3-2.28, Attachment 1 Revision # 14-1 Page 332 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 061 A2.03 Importance Rating 3.4 Auxiliary / Emergency Feedwater: Ability to (a) predict the impacts of the following malfunctions or operations on the AFW; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Loss of DC power Proposed Question: SRO 87 Given the following conditions on Unit 2 while in MODE 1:

• A loss of DC Bus D3 was immediately followed by a Unit trip.

Which ONE (1) of the following:

1.) Identifies the impact on the Auxiliary Feedwater System?

2.) What ACTION should be taken per Technical Specification LCO 3.7.5, Auxiliary Feedwater System?

A. 1.) Auxiliary Feedwater Pump P-140 is INOPERABLE due to the inability to open HV-4716, K-007 Turbine Inlet Steam Supply.

2.) Restore P-140, Auxiliary Feedwater Pump to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

B. 1.) Auxiliary Feedwater Pump P-140 is INOPERABLE due to the inability to open HS-4705 or HS-4706, P-140 Motor Operated Discharge Valves.

2.) Restore P-140, Auxiliary Feedwater Pump to OPERABLE status within 7 days.

C. 1.) Auxiliary Feedwater Pump P-140 is INOPERABLE due to the inability to open HS-4705 or HS-4706, P-140 Motor Operated Discharge Valves.

2.) Restore P-140, Auxiliary Feedwater Pump to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

D. 1.) Auxiliary Feedwater Pump P-140 is INOPERABLE due to the inability to open HV-4716, K-007 Turbine Inlet Steam Supply.

2.) Restore P-140, Auxiliary Feedwater Pump to OPERABLE status within 7 days.

Proposed Answer: A Page 333 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Explanation:

A. Correct. P-140 is supplied with DC power from DC Buses D1, D2, and D3. The Main Steam Inlets and Pump Discharge Valves are powered from D1 and D2. DC Bus D3 controls the K-007 Turbine Inlet Steam Supply and Governor Control Unit. Technical Specifications requires restoring one AFW Train to OPERABLE status for reasons other than CONDITION A within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

B. Incorrect. Plausible because these valves are DC powered, however, these valves are powered from DC Buses D1 and D2. The Technical Specification ACTION would be correct for the Main Steam Inlet Valves.

C. Incorrect. Plausible because the Technical Specification ACTION is correct, however, the wrong valves were selected. These valves are powered from DC Buses D1 and D2.

D. Incorrect. Plausible because DC Bus D3 controls the K-007 Turbine Inlet Steam Supply and Governor Control Unit. The Technical Specification ACTION would be correct for the Main Steam Inlet Valves.

Technical Reference(s) SD-SO23-780, Figure 1 Attached w/ Revision # See SD-SO23-780, Pages 90 & 91 Comments / Reference Unit 2 Technical Specifications LCO 3.7.5 Proposed references to be provided during examination: None Learning Objective: DESCRIBE the Technical Specification operability requirements and action 54868 / 80607 statements associated with the Auxiliary Feedwater System.

ANALYZE normal and abnormal operations of the 1E 125 VDC and 120 VAC Electrical System.

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 2, 5 Page 334 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SD-SO23-780, Figure 1 Revision # 10 Comments /

Reference:

From SD-SO23-780, Figure 1 Revision # 10 Page 335 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SD-SO23-780, Page 90 Revision # 10 Page 336 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SD-SO23-780, Page 91 Revision # 10 Comments /

Reference:

From Unit 2 Technical Specifications LCO 3.7.5 Amendment # 164 Page 337 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From Unit 2 Technical Specifications LCO 3.7.5 Amendments # 127 & 164 Page 338 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 073 G 2.1.32 Importance Rating 4.0 Process Radiation Monitoring: Conduct of Operations: Ability to explain and apply system limits and precautions Proposed Question: SRO 88 Which ONE (1) of the following is required when 2/3 FI-7643, Radwaste Discharge Process Flow Monitor is declared INOPERABLE?

When 2/3 FI-7643 Radwaste Discharge Process Flow Monitor is declared INOPERABLE then...

A. declare 2/3 RE-7813, Radwaste Discharge Radiation Monitor INOPERABLE.

B. secure all liquid radioactive releases until 2/3 FI-7643 Flow Monitor is repaired to avoid exceeding Offsite Dose Calculation Manual (ODCM) limits.

C. initiate four (4) hour channel checks of 2/3 RE-7813, Radwaste Discharge Radiation Monitor to ensure release limits are not exceeded.

D. effluent releases may continue as long as the flow rate is estimated every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Proposed Answer: A Explanation:

A. Correct. This is the action required per the Limitations and Specifics 2.1.

B. Incorrect. Plausible because the release would be secured and the ODCM referenced, however, RE-7813 must be declared INOPERABLE.

C. Incorrect. Plausible because this action would be performed if DAS alarm monitoring of RE-7813 was not available per Attachment 2, however, RE-7813 must be declared INOPERABLE.

D. Incorrect. Plausible because this action is allowed per Attachment 14, however, RE-7813 must be declared INOPERABLE and the sample time is incorrect.

Technical Reference(s) SO23-8-7, L&S 2.1 & 2.3 Attached w/ Revision # See SO23-8-7, Attachments 2 & 14 Comments / Reference Proposed references to be provided during examination: None Page 339 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Learning Objective: DIRECT radwaste processing and release activities.

187721 / 192875 ENSURE compliance with Technical Specifications and other regulatory requirements.

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 1, 4 Comments /

Reference:

From SO23-8-7, L&S 2.1 & 2.3 Revision # 16 Page 340 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SO23-8-7, Attachment 2 Revision # 16 Page 341 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SO23-8-7, Attachment 14 Revision # 16 Page 342 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 004 A2.20 Importance Rating 2.7 Chemical and Volume Control: Ability to (a) predict the impacts of the following malfunctions or operations on the CVCS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Shifting demineralizer while divert valve is lined up to VCT Proposed Question: SRO 89 Given the following conditions with Unit 2 at 100% power BOL:

• A freshly borated demineralizer is being placed in service with 2LV-0227A, Volume Control Tank Inlet Valve aligned to the Volume Control Tank when the following occurs:

Annunciator 50A02 - COLSS ALARM actuates.

• The Shift Technical Advisor has observed MSBSCAL and calculated Core Thermal Power at 100.7%.

Which ONE (1) of the following:

1.) Identifies the impact on the Facility License?

2.) What action is required?

A. 1.) Ensure the STA performs a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Notification Report to the NRC.

2.) Immediately reduce power to less than or equal to 100%.

B. 1.) Ensure Nuclear Licensing and Compliance files a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Notification Report to the NRC.

2.) Maintain current power level until it is determined that NI / T power does not exceed 100% power averaged over an 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> period.

C. 1.) Ensure Nuclear Licensing and Compliance files a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Notification Report to the NRC 2.) Immediately reduce power to less than or equal to 100%.

D. 1.) Ensure the STA performs a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Notification Report to the NRC.

2.) Maintain current power level until it is determined that NI / T power does not exceed 100% power averaged over an 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> period.

Proposed Answer: C Page 343 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible because power must be immediately reduced, however, a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> report is required.

B. Incorrect. Plausible because this action is allowed but only if power had not exceeded 100.6%. A 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> report is required and correct.

C. Correct. When power level has exceeded 100.6% it must be immediately reduced to 100%.

SO123-0-A7 requires a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> report.

D. Incorrect. Plausible because this action is allowed but only if power had not exceeded 100.6%. A 24 vice 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> report is required.

Technical Reference(s) SO23-5-1.7, Steps 2.4.3 & 6.4.5 Attached w/ Revision # See SO23-3-2.4, L&S 2.9 Comments / Reference SO23-5-1.7, L&S 2.2 SO123-0-A7, Attachments 1 & 3 Proposed references to be provided during examination: None Learning Objective: ENSURE compliance with Technical Specifications and other regulatory 192875 requirements.

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 1, 5 Page 344 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SO23-5-1.7, Step 6.4.5 Revision # 38-1 Comments /

Reference:

From SO23-3-2.4, L&S 2.9 Revision # 18 Page 345 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SO23-5-1.7, Step 2.4.3 Revision # 38-1 Comments /

Reference:

From SO23-5-1.7, L&S 2.2 Revision # 38-1 Comments /

Reference:

From SO123-0-A7, Attachment 1 Revision # 7 Comments /

Reference:

From SO123-0-A7, Attachment 3 Revision # 7 Page 346 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 103 A2.04 Importance Rating 3.6 Containment System: Ability to (a) predict the impacts of the following malfunctions or operations on the containment system; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Containment evacuation (including recognition of the alarm)

Proposed Question: SRO 90 Given the following MODE 1 conditions:

• A Containment Mini-Purge is in progress to facilitate a Containment Entry in progress.

• Heavy smoke has just been reported on the Penetration Building 63 elevation.

Which ONE (1) of the following:

1.) Identifies the impact on the Containment Mini-Purge?

2.) What action should be taken to mitigate the situation?

A. 1.) Secure Containment Mini-Purge Supply Fan MA-379 to prevent injecting smoke into Containment.

2.) Actuate the Containment Emergency Evacuation Siren and Emergency Evacuation Siren on CR-57 per SO23-13-1, Local Area Evacuation.

B. 1.) Initiate a Containment Purge Isolation Signal to facilitate opening the Personnel Hatch doors for evacuation.

2.) Actuate the Containment Emergency Evacuation Siren on CR-57 and dispatch SCBA equipped personnel to facilitate evacuation of Containment personnel per SO23-13-1, Local Area Evacuation.

C. 1.) Secure Containment Mini-Purge Exhaust Fan MA-059 to prevent violating Technical Specification limit for negative Containment pressure.

2.) Actuate the Emergency Evacuation Siren on CR-57 per SO23-13-1, Local Area Evacuation.

D. 1.) Initiate a Containment Purge Isolation Signal to facilitate opening the Personnel Hatch doors for evacuation.

2.) Actuate SIREN LOCAL at the NCO turret (PA System Siren) and Emergency Evacuation Siren and direct personnel to remain in the Personnel Hatch until smoke has been cleared per SO23-13-1, Local Area Evacuation.

Proposed Answer: B Page 347 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible because securing the Containment Mini-Purge Supply Unit would prevent injection of smoke into Containment, however, this would leave the Exhaust Unit running and a potential for a negative pressure and a personnel hazard as well as a Technical Specification violation inside Containment.

B. Correct. Initiating a Containment Purge Isolation Signal will assist in opening the personnel inner and outer Hatch doors. SO23-13-1 includes actions to dispatch SCBA equipped personnel to facilitate local evacuation.

C. Incorrect. Plausible because operating the Containment Mini-Purge Exhaust Unit without the associated Supply Unit running can result in a negative pressure inside Containment and violation of the Technical Specification negative pressure limit. Additionally, the wrong evacuation alarm is being sounded per SO23-13-1, Local Area Evacuation.

D. Incorrect. Plausible because a Containment Purge Isolation Signal is desired, however, although the listed alarms may be actuated, the Containment Evacuation alarm must be sounded in order to have personnel evacuate Containment.

Technical Reference(s) SO23-13-1, Entry Conditions Attached w/ Revision # See SO23-13-1, Step 1 Comments / Reference SO23-1-4.2, L&S 2.2 & 5.1 Technical Specification 3.6.4 Proposed references to be provided during examination: None Learning Objective: EXPLAIN the importance of the Containment Air Handling System with regard 81635 / 81640 to plant safety.

ANALYZE normal and abnormal operations of the Containment Air Handling System.

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

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

Reference:

From Technical Specification 3.6.4 Amendment # 127 Page 348 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Page 349 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SO23-13-1, Entry Conditions Revision # 2-2 Page 350 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SO23-13-1, Step 1 Revision # 2-2 Page 351 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SO23-1-4.2, L&S 2.2 & 5.1 Revision # 29 Page 352 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 2 K/A # 011 G 2.4.6 Importance Rating 4.7 Pressurizer Level Control System: Emergency Procedures / Plan: Knowledge of EOP mitigation strategies Proposed Question: SRO 91 Given the following conditions following a Unit 2 trip due to a Small Break Loss of Coolant Accident:

• The crew is performing actions of SO23-12-11, EOI Supporting Attachments, Attachment 2, FS-7, Verify SI Throttle / Stop Criteria.

• Offsite power has been maintained throughout the event.

• All High Pressure Safety Injection flow is secured.

• Pressurizer Level is 80% and rising.

Which ONE (1) of the following mitigation strategies should be employed per FS-7, Verify SI Throttle / Stop Criteria?

A. RESET SIAS per SO23-3-2.22, ESFAS Operation, then initiate FS-31, Establish CVCS Letdown Flow and FS-33, MONITOR RCS Solid Operation.

B. Immediately secure the Charging Pumps and then initiate SO23-12-11, Attachment 3, Cooldown / Depressurization.

C. Concurrently perform FS-33, MONITOR RCS Solid Operation and FS-31, Establish CVCS Letdown Flow and immediately secure the Charging Pumps.

D. Initiate FS-31, ESTABLISH CVCS Letdown Flow, initiate FS-33, MONITOR RCS Solid Operation, and then secure all but one (1) Charging Pump.

Proposed Answer: C Page 353 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible because the RNO actions are correct, however, it is not necessary to reset SIAS to establish CVCS Letdown flow.

B. Incorrect. Plausible because these actions are partially correct, however, they are out of order and do not include initiating the Floating Steps. Initiating Attachment 3 is plausible because a cooldown and depressurization would lower PZR level. Also no boration is taking place therefore cooldown is not allowed.

C. Correct. This is the RNO action required in FS-7 when Pressurizer level is greater than 80%.

Once the RNO actions are initiated, the Charging Pumps will be secured.

D. Incorrect. Plausible because it could be thought that 1 Charging Pump remain operating when placing Letdown in service, and this is the RNO action required in FS-7 when Pressurizer level is greater than 80%, however, all Charging Pumps must be secured.

Technical Reference(s) SO23-12-11, Attachment 2, FS-7 Attached w/ Revision # See Comments / Reference Proposed references to be provided during examination: None Learning Objective: Per the EOI Attachments procedure, SO23-12-11, DESCRIBE:

55279 - The basis for each step, caution or note.

- The CEN-152 basis or reason for these steps.

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 5 Page 354 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SO23-12-11, Attachment 2, FS-7 Revision # 6 Page 355 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 2 K/A # 034 A2.02 Importance Rating 3.9 Fuel Handling Equipment: Ability to (a) predict the impacts of the following malfunctions or operations on the Fuel Handling System; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Dropped cask Proposed Question: SRO 92 Given the following conditions Unit 3 in MODE 1:

• A Spent Fuel Cask containing irradiated fuel has just been dropped.

• RT-7850, Spent Fuel Cask Loading Area Radiation Monitor is in alarm.

• RE-7822 and RE-7823, Fuel Handling Building Ventilation Airborne Radiation Monitors are trending to an alarm condition.

Which ONE (1) of the following:

1.) Identifies the impact on the Fuel Handling Building and Control Room Ventilation Systems?

2.) What action should be taken to mitigate the situation?

A. 1.) Ensure FHIS and CRIS automatically actuate.

2.) Direct performance of SO23-13-20, Fuel Handling Accidents / Loss of Cavity or SFP Level Control.

B. 1.) Ensure FHIS automatically actuates and direct immediate actuation of CRIS.

2.) Direct performance of SO23-13-20, Fuel Handling Accidents / Loss of Cavity or SFP Level Control and SO23-13-1, Local Area Evacuation.

C. 1.) Ensure CRIS automatically actuates and direct immediate actuation of FHIS.

2.) Direct performance of SO23-13-20, Fuel Handling Accidents / Loss of Cavity or SFP Level Control.

D. 1.) Direct immediate actuation of FHIS and CRIS.

2.) Direct performance of SO23-13-20, Fuel Handling Accidents / Loss of Cavity or SFP Level Control and SO23-13-1, Local Area Evacuation.

Proposed Answer: D Page 356 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible because AOI procedure entry is correct but not complete. FHIS and CRIS will automatically actuate only if the Fuel Handling Building and Control Room sensed high radiation.

B. Incorrect. Plausible because CRIS immediate actuation is required per the AOI and it is also correct to enter both AOIs as the high radiation in the Fuel Handling Building will require a Local Area Evacuation. It could be thought that FHIS will automatically actuate once the Spent Fuel Cask Loading Area Radiation Monitor setpoint is reached, however, it is not actuated by this signal.

C. Incorrect. Plausible because the setpoint for CRIS actuation is the same as the setpoint for the Spent Fuel Cask Radiation Monitor high alarm. It could be thought that both systems would automatically actuate. CRIS will actuate only if the Control Room sensed high radiation. AOI procedure entry is correct but not complete.

D. Correct. Entry into both AOIs is required as the high radiation in the Fuel Handling Building will require a Local Area Evacuation and the dropped cask requires SO23-13-20 entry. FHIS requires actuation as it is trending to an alarm condition (per the RNO column) on high radiation in the Fuel Handling Building; however, CRIS must be immediately actuated per the Fuel Handling Accident AOI.

Technical Reference(s) SO23-13-20, Step 2a Attached w/ Revision # See SD-SO23-720, Figure 2H Comments / Reference SD-SO23-435, Figure 6 SO23-13-1, Step 1 SO23-15-60.B.08 SO23-15-60.A1.23 Proposed references to be provided during examination: None Learning Objective: As the SRO, DIRECT the response to Fuel Handling Accidents/Loss of Cavity 54861 or SFP Level Control per SO23-13-20.

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 7 Page 357 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SO23-13-20, Step 2a Revision # 9 Comments /

Reference:

From SO23-13-20, Step 2d Revision # 9 Page 358 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SO23-13-1, Step 1 Revision # 2-2 Page 359 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SD-SO23-720, Figure 2H Revision # 8 Page 360 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SD-SO23-435, Figure 6 Revision # 3 Comments /

Reference:

From SO23-15-60.B.08 Revision # 15 Page 361 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SO23-15-60.A1.23 Revision # 15 Page 362 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 2 K/A # 015 A2.03 Importance Rating 3.5 Nuclear Instrumentation System: Ability to (a) predict the impacts of the following malfunctions or operations on the NIS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Xenon oscillations Proposed Question: SRO 93 Given the following while at 100% power Middle-Of-Core (MOC) conditions:

• A xenon oscillation is in progress greater than 0.05 Axial Shape Index (ASI) units.

• Axial Shape Index is currently more positive than Equilibrium Shape Index (ESI).

• The oscillation is moving towards the top of the Core.

• It is determined that Core Average Axial Shape Index is not within limits.

Which ONE (1) of the following:

1.) Identifies the guidance provided to the Reactor Operator to mitigate the xenon oscillation?

2.) What Technical Specification ACTION should be taken?

A. 1.) Direct inserting PLCEAs in small, smooth frequent movements of less than 3 inches per minute to maintain ASI at ESI +/- 0.01 shape index units.

2.) Restore Axial Shape Index to within limits in two (2) hours.

B. 1.) Commence a Reactor Coolant System boration to lower temperature and drive ASI towards ESI.

2.) Restore Axial Shape Index to within limits in four (4) hours.

C. 1.) Direct inserting Group 6 CEAs in small, smooth frequent movements of less than 5 inches per minute to maintain ASI at ESI +/- 0.01 shape index units.

2.) Restore Axial Shape Index to within limits in four (4) hours.

D. 1.) Commence a Reactor Coolant System dilution to raise temperature and drive ASI towards ESI.

2.) Restore Axial Shape Index to within limits in two (2) hours.

Proposed Answer: A Page 363 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Explanation:

A. Correct. Given the fact that this is identified as a large oscillation in the stem, inserting either the Part Length or Group 6 CEAs is appropriate at the speed listed.

B. Incorrect. Plausible because all components of this answer are correct for controlling a xenon oscillation, however, given the size of the oscillation listed this is an inappropriate response.

Controlling a small xenon oscillation with temperature is appropriate however, this is a large oscillation.

C. Incorrect. Plausible because all components of this answer are correct for controlling a xenon oscillation including the use of Group 6 rods instead of the part length CEAs, however, the rods speed listed is too fast to avoid pellet clad interaction.

D. Incorrect. Plausible because all components of this answer are correct for controlling a xenon oscillation, however, commencing a dilution to raise temperature and drive ASI towards ESI only works if ASI is more negative than ESI. The condition in the stem is opposite this.

Technical Reference(s) SO23-5-1.7, Attachment 6 Attached w/ Revision # See Unit 2 Technical Specifications LCO 3.2.5 Comments / Reference Proposed references to be provided during examination: None Learning Objective: EVALUATE plant status against Technical Specification requirements.

189963 / 193001 DIRECT Operator response during Transients including monitoring of key parameters.

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 5 Page 364 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SO23-5-1.7, Attachment 6, Section 3.0 Revision # 38-1 Comments /

Reference:

From SO23-5-1.7, Attachment 6, Section 3.0 Revision # 38-1 Page 365 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From Unit 2 Technical Specifications LCO 3.2.5 Amendment # 127 Page 366 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 3 Group # 1 K/A # G 2.1.25 Importance Rating 4.2 Conduct of Operations: Ability to interpret reference materials, such as graphs, curves, tables, etc.

Proposed Question: SRO 94 The Unit is operating at 60% power when the following events occur:

• One (1) Group 2 CEA drops to the bottom of the Core.

• After Reactor Coolant System Tcold is restored and the plant stabilized, the initial power reduction from the dropped CEA is determined to be 3%.

Which ONE (1) of the following is the MINIMUM ADDITIONAL reduction in power required during the first hour after the time of the CEA drop per Licensee Controlled Specification 3.1.105, CEA Misalignment Power Reduction and what other action is required?

A. 2%; Restore the misaligned CEA to within 7 inches of its group within two hours.

B. 7%; Restore the misaligned CEA to within 7 inches of its group within two hours.

C. 2%; Restore the misaligned CEA to within 7 inches of its group within six hours.

D. 7%; Restore the misaligned CEA to within 7 inches of its group within six hours.

Proposed Answer: B Explanation:

A. Incorrect. Plausible because the Technical Specification requirement is correct, however, this is the power reduction required for a Group 6 CEA.

B. Correct. This is the correct power reduction per the LCS and the AOI. Additionally, Technical Specifications require alignment within two hours.

C. Incorrect. Plausible because the misaligned CEA must be realigned to its group within 7 inches, however, the time restriction is two hours. The six hour requirement is applicable if the CEA cannot be realigned but also requires MODE 3 entry. Additionally, this is the power reduction required for a Group 6 CEA.

D. Incorrect. Plausible because the power reduction is correct, however, the CEA must be aligned within two hours. The six hour requirement is applicable if the CEA cannot be realigned but also requires MODE 3 entry.

Technical Reference(s) LCS Figures 3.1.105-1 Attached w/ Revision # See SO23-13-13, Step 2 Comments / Reference Technical Specification 3.1.5 Page 367 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Proposed references to be provided during examination: None Learning Objective: EVALUATE plant status against Technical Specification requirements.

189963 / 193001 DIRECT Operator response during Transients including monitoring of key parameters.

Question Source: Bank #

Modified Bank # 72622 (Note changes or attach parent)

New Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 1, 5 Comments /

Reference:

From LCS Figure 3.1.105-1 Revision # 2/19/97 Page 368 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From LCS Figure 3.1.105-2 Revision # 2/19/97 Page 369 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SO23-13-13, Step 2 Revision # 11-2 Page 370 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From Technical Specification 3.1.5 Amendment # 200 Page 371 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 3 Group # 1 K/A # G 2.1.40 Importance Rating 3.9 Conduct of Operations: Knowledge of refueling administrative requirements Proposed Question: SRO 95 The Unit is in MODE 6 with the Reactor Vessel Head removed. The following conditions exist:

• Fuel transfer activities have been stopped while a problem with the Fuel Transfer Carriage is being fixed.

• The Refueling SRO is attending an outage planning brief in the Outage Control Center.

• Nuclear Fuels Management asks the Control Room to direct the Refueling Bridge Operator to uncouple a CEA for weighing.

Which ONE (1) of the following is correct concerning this situation?

A. The Refueling SRO is not required to be present at the Refueling Pool, since this is not considered a CORE ALTERATION.

B. A Control Room SRO must oversee this activity in the Control Room since this is a CORE ALTERATION. No other requirements are necessary.

C. The Refueling SRO must be present at the Refueling Pool since this is a CORE ALTERATION.

D. A Control Room SRO and Reactor Engineer must oversee this activity in the Control Room, but it is not considered a CORE ALTERATION.

Proposed Answer: C Explanation:

A. Incorrect. Plausible because it does not involve fuel movement, a reactivity change, or fuel damage, however, it would be considered a CORE ALTERATION.

B. Incorrect. Plausible because this activity is a CORE ALTERATION, however, the Refueling SRO must be present at the Refueling Pool for this manipulation.

C. Correct. The Refueling SRO must be present at the Refueling Pool since this is a CORE ALTERATION.

D. Incorrect. Plausible because the answer includes a Senior Reactor Operator and Reactor Engineer overseeing this manipulation, however, this action is considered a CORE ALTERATION.

Technical Reference(s) SO23-5-1.8, Attachment 5 Attached w/ Revision # See Comments / Reference Page 372 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Proposed references to be provided during examination: None Learning Objective: EVALUATE plant status against Technical Specification requirements.

189963 / 54767 REVIEW the duties and responsibilities of the Refueling Supervisor including:

- Responsibility for supervision of Core Alterations.

Question Source: Bank #

Modified Bank # 73793 (Note changes or attach parent)

New Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 6, 7 Page 373 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SO23-5-1.8, Attachment 5 Revision # 18 Page 374 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SO23-5-1.8, Attachment 5 Revision # 18 Page 375 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From Exam Bank #73793 Revision 7/15/03 The plant is in Mode 6 with the reactor vessel head removed. The following conditions exist:

• Fuel transfer activities have been stopped while a problem with the upender is being fixed.

• The Refueling SRO is attending an outage planning brief

• The Nuclear Fuels Mgmt people in the Control Room ask the Refueling Bridge Operator to reposition a portable light in the vessel.

Which of the following is correct concerning this situation?

A. The Refueling SRO is not required to be present at the Refueling Pool, since this is not considered a core alteration.

B. A Control Room SRO must oversee this activity in the Control Room since this is a core alteration.

No other requirements are necessary.

C. A Control Room SRO and Reactor Engineer must oversee this activity in the Control Room, but it it not considered a core alteration.

D.The Refueling SRO must be present at the Refueling Pool since this is a core alteration Page 376 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 3 Group # 2 K/A # G 2.2.5 Importance Rating 3.2 Equipment Control: Knowledge of the process for making design or operating changes to the facility Proposed Question: SRO 96 A condition has been identified that requires a Non-Conformance Report (NCR). The NCR has been dispositioned as a non-permanent repair. Several Control Room stick-file drawings will be impacted by the repair.

Which ONE (1) of the following is an appropriate means of controlling the Temporary Modification?

A. An Operations Abnormal Evolution should be generated.

B. The Non-Conformance Report may serve as the controlling document.

C. A Temporary Engineering Change Package should be generated.

D. An ISCO Engineering Change Package should be generated.

Proposed Answer: C Explanation:

A. Incorrect. Plausible because this could be used to procedurally control the temporary modification, however, a Temp ECP is required to control the change itself.

B. Incorrect. Plausible because an NCR is generated, however, it must include a Temp ECP.

C. Correct. As outlined in SO123-XV-5.

D. Incorrect. Plausible because an ECP must be generated, however, a Temp ECP is required.

Technical Reference(s) SO123-XV-5.1, Step 6.4 & 6.9 Attached w/ Revision # See Comments / Reference Proposed references to be provided during examination: None Learning Objective: DESCRIBE the process for installation and removal of a temporary ECP and 55192 the responsibilities of operations during this process, including the following:

- Conditions where a Temporary ECP would be used.

Question Source: Bank # 73208 Modified Bank # (Note changes or attach parent)

New Page 377 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Question History: Last NRC Exam 2005A NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 3 Comments /

Reference:

From SO123-XV-5.1, Step 6.9 Revision # 8 Page 378 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SO123-XV-5.1, Step 6.4 Revision # 8 Page 379 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 3 Group # 2 K/A # G 2.2.37 Importance Rating 4.6 Equipment Control: Ability to determine operability and/or availability of safety related equipment Proposed Question: SRO 97 When the Reactor Coolant System in a Reduced Inventory Condition it is required to have at least two (2) means of keeping the Reactor Core covered. One of these is an "AVAILABLE" HPSI or Containment Spray Pump.

Which ONE (1) of the following best describes "AVAILABLE" in this context?

A. Pump is operating in the Recirculation Mode.

B. Ready for use within a short enough time to meet the intended need.

C. Completely OPERABLE with all support systems per the Technical Specification definition.

D. INOPERABLE but required to be made OPERABLE through Control Room manipulations only.

Proposed Answer: B Explanation:

A. Incorrect. Plausible because with the Reactor Coolant System in a Reduced Inventory Condition it could be thought that the Containment Spray Pump should be running and available for immediate injection, however, the pump need not be operating.

B. Correct. Per the Attachments in SO23-3-1.8, Draining the Reactor Coolant System to a Reduced Inventory Condition.

C. Incorrect. Plausible because one HPSI Pump must be OPERABLE, however, the additional HPSI Pump or a Containment Spray Pump need only be AVAILABLE.

D. Incorrect. Plausible because the definition of AVAILABLE discusses instrumentation required either in the Control Room or where it is intended to function, however, the pump is not INOPERABLE.

Technical Reference(s) SO23-3-1.8, Attachments 11 & 14 Attached w/ Revision # See Comments / Reference Proposed references to be provided during examination: None Learning Objective: ENSURE compliance with Technical Specifications and other regulatory 192875 requirements.

Page 380 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Question Source: Bank # 65418 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 1, 5 Comments /

Reference:

From SO23-3-1.8, Attachment 14 Revision # 26 Page 381 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SO23-3-1.8, Attachment 11 Revision # 26 Page 382 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 3 Group # 3 K/A # G 2.3.15 Importance Rating 3.1 Radiation Control: Knowledge of radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc.

Proposed Question: SRO 98 While in MODE 1, the Reactor Operator reports to you that both of the Control Room Airborne (CRIS) Radiation Monitors, RE-7824-1 and RE-7825-2, have failed low.

The I&C Technician indicates that it will take five days to replace the detectors.

Which ONE (1) of the following actions must be taken in accordance with Technical Specifications?

A. Return the monitors to service, or be in HOT STANDBY within six (6) hours.

B. Establish a portable Control Room Gaseous Process Radiation Monitoring station.

C. Perform a surveillance test on the Control Room Emergency Air Cleanup System.

D. Place one Control Room Emergency Air Cleanup System Train in Emergency Mode within one (1) hour.

Proposed Answer: D Explanation:

A. Incorrect. Plausible because this action would be required by Technical Specification 3.0.3, however, the provisions of that Technical Specification are not applicable to this LCO (see NOTES in LCO).

B. Incorrect. Plausible because it could be thought that establishing a portable radiation monitoring station would allow the control room to maintain a normal ventilation alignment without initiating CREACUS since the manual actuation is not affected. Because the provisions of Technical Specification LCO 3.0.3 do not apply, CREACUS must be placed in Emergency Mode.

C. Incorrect. Plausible because the CRIS logic consists of manual trip, high airborne radiation detection and actuation logic functions. It could be thought that performing a Surveillance Test on CREACUS would fulfill the requirement with both detectors INOPERABLE.

D. Correct. This is the required ACTION per Technical Specification Technical Reference(s) Technical Specification LCO 3.3.9 & Bases Attached w/ Revision # See Comments / Reference Proposed references to be provided during examination: None Page 383 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Learning Objective: Given a system/component inoperability, DETERMINE the applicable LCO(s),

55307 required action, and action times using the Tech Specs and the LCS.

Question Source: Bank # 73813 Modified Bank # (Note changes or attach parent)

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

Reference:

From Technical Specification LCO 3.3.9 Amendment # 132 Page 384 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From Technical Specification LCO 3.3.9 Amendment # 132 Page 385 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From Technical Specification LCO 3.3.9 Bases Amendment # 127 Page 386 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 3 Group # 4 K/A # G 2.4.11 Importance Rating 4.2 Emergency Procedures / Plan: Knowledge of abnormal condition procedures Proposed Question: SRO 99 Given the following conditions:

• A seismic event has occurred.

Which ONE (1) of the following personnel are qualified to perform Attachment 3, Fire Zone Inspections per SO23-13-3, Earthquake?

A. Housekeeping personnel or Fire Watch Personnel.

B. Fire Watch personnel or Shift Captain.

C. Shift Captain or Plant Operators.

D. Equipment Control Coordinator or Plant Operators.

Proposed Answer: C Explanation:

A. Incorrect. Plausible because it could be thought Fire Watch personnel are qualified but they are not.

B. Incorrect. Plausible because Shift Captain is correct, however, Fire Watch personnel are not.

C. Correct. These individuals are qualified per SO23-13-3, Attachment 3.

D. Incorrect. Plausible because Plant Operators is correct, however, the Equipment Control Coordinator is not.

Technical Reference(s) SO23-13-3, Attachment 3 Attached w/ Revision # See Comments / Reference Proposed references to be provided during examination: None Learning Objective: DIRECT operator response to a fire in the plant.

188590 Question Source: Bank # 66797 Modified Bank # (Note changes or attach parent)

New Page 387 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 5 Comments /

Reference:

From SO23-13-3, Attachment 3 Revision # 6 Page 388 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 3 Group # 4 K/A # G 2.4.23 Importance Rating 4.4 Emergency Procedures / Plan: Knowledge of the bases for prioritizing emergency procedure implementation during emergency operations Proposed Question: SRO 100 Given the following conditions during a Steam Generator Tube Rupture:

• The crew is performing steps in SO23-12-4, Steam Generator Tube Rupture for cooling and depressurizing the Reactor Coolant System.

Which ONE (1) of the following identifies when reducing Reactor Coolant System pressure reduction takes priority over maintaining a minimum 20°F Core Exit Saturation Margin requirement and what action is required?

A. Pressurizer level is rapidly increasing. Initiate FS-7, Verify SI Throttle / Stop Criteria when Pressurizer level exceeds 30% per SO23-12-11, EOI Supporting Attachments.

B. Ruptured Steam Generator level is rapidly increasing. Continue lowering Pressurizer pressure using Normal Spray flow per SO23-12-4, Steam Generator Tube Rupture.

C. Feedwater is unavailable to the intact Steam Generator. Secure Reactor Coolant Pumps and initiate Auxiliary Spray flow per SO23-12-4, Steam Generator Tube Rupture.

D. Ruptured Steam Generator level cannot be maintained above the level of the U-tubes. Initiate Attachment 18, Backflowing then Feeding a Ruptured Steam Generator per SO23-12-11, EOI Supporting Attachments.

Proposed Answer: B Page 389 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible because it could be thought that Pressurizer overfill could create a Pressurized Thermal Shock event especially if Reactor Coolant System temperature were not controlled as Reactor Coolant System pressure was being reduced. Initiating FS-3 would be desirable as this Pressurizer level is one of the criteria that must be met.

B. Correct. When the ruptured Steam Generator level is rising rapidly it is expected that the crew will reduce RCS pressure and sacrifice Core Exit Saturation Margin in order to avoid overfilling the Steam Generator. This is performed by remaining in SO23-12-4 and initiating Normal Spray flow.

C. Incorrect. Plausible because if feedwater was not available to the intact Steam Generator it could be thought that reducing RCS pressure takes priority to minimize leakage across the primary to secondary boundary, however, cooldown using the ruptured Steam Generator is not procedurally restricted. Securing Reactor Coolant Pumps would minimize heat input and initiating Auxiliary Spray flow would lower RCS pressure.

D. Incorrect. Plausible because maintaining level above the U-tubes is desirable for Iodine scrubbing purposes, however, not for the conditions listed. This appears plausible if one thought that this would help raise level, however, this Attachment is used to cool and depressurize the Steam Generator in order to align Shutdown Cooling.

Technical Reference(s) SO23-14-4, Step 12 Caution Attached w/ Revision # See SO23-12-04, Step 12 Comments / Reference SO23-12-11, FS-7 SO23-12-11, Attachment 18 Proposed references to be provided during examination: None Learning Objective: Per the EOI Attachments procedure, SO23-12-04, DESCRIBE:

53000 - The basis for each step, caution or note.

- The CEN-152 basis or reason for these steps.

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 5 Page 390 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SO23-14-4, Step 12 Caution Revision # 7 Page 391 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SO23-12-04, Step 12 Revision # 21 Page 392 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SO23-12-11, FS-7 Revision # 6 Page 393 of 394 Rev 1

ES-401 SONGS Sept 2008 NRC SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SO23-12-11, Attachment 18 Revision # 6 Page 394 of 394 Rev 1