ML14174B412

2014-06 Final Written Exam

ML14174B412
Person / Time
Site:	Comanche Peak
Issue date:	06/16/2014
From:	Vincent Gaddy Operations Branch IV
To:	Luminant Generation Co
References
50-445/OL-14, 50-446/OL-14
Download: ML14174B412 (483)

Text

ES-401 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. Date: 3/24/2014 Tier 1 Change: 4 Group 1 K/A 007 EK1.02 Level of Difficulty: 2 Importance Rating 3.4 Reactor Trip-Stabilization-Recovery: Knowledge of the operational implications of the following concepts as they apply to the Reactor Trip: Shutdown margin Proposed Question: 1 Why are Control Rod insertion limits established for power operation?

A. Minimizes the worth of a postulated dropped Control Rod.

B. Maintains a negative Moderator Temperature Coefficient.

C. Provides adequate shutdown margin after a Reactor Trip.

D. Ensures sufficient positive reactivity to offset the Power Defect.

Proposed Answer: C Explanation:

A. Incorrect. Plausible because it could be thought that if RILs are high enough the effect of a dropped rod would be minimized by the remaining rods being above the RIL.

B. Incorrect. Plausible because a positive moderator temperature coefficient will reduce shutdown margin due to the positive reactivity added during a heatup. So it could be thought that RILs provide a mechanism for ensuring a negative MTC is maintained.

C. Correct. SDM requirements provide sufficient reactivity margin to ensure that acceptable fuel design limits will not be exceeded for normal shutdown and anticipated operational occurrences (AOOs). As such, the SDM defines the degree of subcriticality that would be obtained immediately following the insertion or scram of all shutdown and control rods, assuming that the single rod cluster assembly of highest reactivity worth is fully withdrawn.

D. Incorrect. Plausible because it could be thought that RILs ensure power defect is offset, however rods are essentially fully withdrawn and boron concentration is reduced to offset the power defect.

Technical Reference(s) Technical Specification LCO 3.1.1 Bases Attached w/ Revision: See Technical Specification LCO 3.1.6 Bases Comments / Reference LO21.GFR.PHY, Page 4 Proposed references to be provided during examination: None Learning Objective: EXPLAIN reactor response to a control rod insertion.

DESCRIBE the basic design of the Rod Insertion Limit (RIL) Monitor System.

Page 1 of 51 CPNPP 2014 NRC RO Written Exam Worksheet 01 to 10

ES-401 RO Written Exam Worksheet Form ES-401-5 Question Source: Bank ILOT1413 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 1 55.43 Page 2 of 51 CPNPP 2014 NRC RO Written Exam Worksheet 01 to 10

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

Technical Specification LCO 3.1.1 Bases Revision: 68 Page 3 of 51 CPNPP 2014 NRC RO Written Exam Worksheet 01 to 10

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

Technical Specification LCO 3.1.6 Bases Revision: 68 Page 4 of 51 CPNPP 2014 NRC RO Written Exam Worksheet 01 to 10

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

LO21.GFR.PHY, Page 4 Revision: 12/17/07 Page 5 of 51 CPNPP 2014 NRC RO Written Exam Worksheet 01 to 10

ES-401 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. Date: 3/24/2014 Tier 1 Change: 5 Group 1 K/A 008 AA1.08 Level of Difficulty: 3 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: 2 Given the following conditions:

Unit 1 is in MODE 1.

1-8010B, Pressurizer Safety Relief Valve B, is partially open.

The following parameters are observed:

Reactor power is 99.8%.

Pressurizer pressure is 2185 psig and slowly lowering.

Pressure Relief Tank (PRT) pressure is 1 psig and slowly rising.

Which of the following lists the approximate temperature and phase of the fluid flowing to the PRT from the open relief valve?

A. 102F, saturated B. 102F, superheated C. 216F, saturated D. 216F, superheated Proposed Answer: C Page 6 of 51 CPNPP 2014 NRC RO Written Exam Worksheet 01 to 10

ES-401 RO Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible because 102F is the saturation temperature for the downstream pressure if not converted to psia. Fluid phase is correct.

B. Incorrect. Plausible because 102F is the saturation temperature for the downstream pressure if not converted to psia. Additionally, if the downstream pressure was not converted to psia, fluid phase would appear to be superheated.

C. Correct. The steam tables are used to determine the enthalpy of the downstream fluid. Enthalpy is determined at absolute pressure. Since the ideal throttling process in isenthalpic, enthalpy does not change across the relief valve. Enthalpy at 2200 psia (2185 psig + 15 psi) = 1122 Btu/lbm.

With downstream pressure at 16 psia, the enthalpy of saturated vapor is approximately 1152 Btu/lbm. Fluid at this pressure that is below this value is saturated. Therefore the fluid on the downstream side of the relief valve is a saturated, or wet, vapor. Temperature of the fluid is equal to the saturation temperature of the fluid pressure (16 psia), which is approximately 216F.

D. Incorrect. Plausible because fluid temperature is correct. However, if the downstream pressure was not converted to psia, fluid phase would appear to be superheated.

Technical Reference(s) Steam Tables Attached w/ Revision: See Comments / Reference Proposed references to be provided during examination: Steam Tables Learning Objective: SOLVE throttling process problems, applying the General Energy Equation.

Question Source: Bank Modified Bank ILOT2314 (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 14 55.43 Page 7 of 51 CPNPP 2014 NRC RO Written Exam Worksheet 01 to 10

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From Mollier Diagram Revision: N/A 2200 psia 216F 6 psia 1122 BTU/lbm Original Question: CPNPP Exam Bank ILOT2314 What is the approximate temperature and phase of the fluid downstream of the pressurizer relief valve if it sticks partially open with 2,200 psia in the pressurizer and a 50 psia backpressure?

A. 281ºF, saturated B. 281ºF, superheated C. 332ºF, saturated D. 332ºF, superheated Answer: A Page 8 of 51 CPNPP 2014 NRC RO Written Exam Worksheet 01 to 10

ES-401 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. Date: 5/09/2014 Tier 1 Change: 6 Group 1 K/A 009 EA2.14 Level of Difficulty: 4 Importance Rating 3.8 Small Break LOCA: Ability to determine and interpret the following as they apply to the small break LOCA: Actions to be taken if PTS limits are violated Proposed Question: 3 Given the following conditions:

Unit 1 is responding to a Small Break Loss of Coolant Accident (SBLOCA).

An Orange path on INTEGRITY is being addressed in accordance with FRP-0.1A, Response to Imminent Pressurized Thermal Shock Condition.

All Reactor Coolant Pumps were stopped in response to the SBLOCA.

Auxiliary Feedwater (AFW) Flow is 70 gpm and stable to Steam Generators (SG) 1-01 and 1-02.

Auxiliary Feedwater (AFW) Flow is 110 gpm and stable to Steam Generators (SG) 1-03 and 1-04.

Reactor Coolant System (RCS) temperature is 240F with a steady cooldown rate of 30F/hr.

Steam Generator 1-01 Atmospheric Relief Valve, 1-PV-2325 indicates fully open.

Steam Dump Group 1, Bank 1 indicates partially open.

Containment Pressure is 7 psig and slowly rising.

SG levels are as follows:

SG 1-01 46% and stable SG 1-02 48% and stable SG 1-03 53% and stable SG 1-04 55% and stable Which of the following action(s) are(is) required by FRP-0.1A?

A. Increase AFW flow to SG 1-01 and 1-02, Maintain AFW flow to SG 1-03 and 1-04.

B. Increase AFW flow to SG 1-01 and 1-02, Increase AFW flow to SG 1-03 and 1-04.

C. Close or isolate Steam Generator 1-01 Atmospheric Relief Valve, 1-PV-2325.

D. Take manual control of Steam Dumps and close Group 1, Bank 1 dump valves.

Proposed Answer: C Page 9 of 51 CPNPP 2014 NRC RO Written Exam Worksheet 01 to 10

ES-401 RO Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible because increasing the AFW flow in SGs 1 and 2 could be considered a reasonable action as both SGs are still below the minimum heat sink level of 50% for Adverse Containment conditions. However, as two SGs are above the minimum level AFW flow should not be increased but controlled to stop the RCS cooldown. SGs 3 and 4 both meet the minimum heat sink level requirement; reducing the AFW flow to these two SGs is the correct action.

B. Incorrect. Plausible because increasing the AFW flow in SGs 1 and 2 could be considered a reasonable action as both SGs are still below the minimum heat sink level of 50% for Adverse Containment conditions. However, as two SGs are above the minimum level AFW flow should not be increased but controlled to stop the RCS cooldown. SGs 3 and 4 both meet the minimum heat sink level requirement, increasing the current AFW flow to these two SGs would not comply with FRP-0.1A to control AFW flow to stop the RCS cooldown as a substantial RCS cooldown rate currently exist with the stable AFW flow of 110 gpm to SGs 3 and 4.

C. Correct. Closing or isolating SG 1-01 ARV is the required action based on istopping the RCS cooldown.

D. Incorrect. Plausible because ensuring steam dumps are closed is an action required by FRP-0.1A to stop RCS cooldown, however in this situation Containment pressure is above 6.3 psig so the MSIVs are closed and closing the steam dumps is immaterial.

Technical Reference(s) FRP-0.1A, Step 2 Attached w/ Revision: See FRP-0.1A, Step 2 Bases Comments / Reference Proposed references to be provided during examination: None Learning Objective: Given a procedural Step, NOTE, or CAUTION, DISCUSS the reason or basis for the Step, NOTE, or CAUTION in FRP-0.1, Response to Imminent Pressurized Thermal Shock 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 10 55.43 Page 10 of 51 CPNPP 2014 NRC RO Written Exam Worksheet 01 to 10

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

FRP-0.1A, Step 2 Revision: 8 Page 11 of 51 CPNPP 2014 NRC RO Written Exam Worksheet 01 to 10

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

FRP-0.1A, Step 2 Bases Revision: 8 Page 12 of 51 CPNPP 2014 NRC RO Written Exam Worksheet 01 to 10

ES-401 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. Date: 5/9/2014 Tier 1 Change: 5 Group 1 K/A 011 EA2.03 Level of Difficulty: 3 Importance Rating 3.7 Large Break LOCA: Ability to determine and interpret the following as they apply to the large break LOCA: Consequences of managing LOCA with loss of CCW Proposed Question: 4 Given the following conditions:

A Large Break Loss of Coolant Accident has occurred inside Unit 1 Containment.

Reactor Coolant System and Containment pressure are both approximately 30 psig.

Component Cooling Water (CCW) Pump 1-01 has tripped.

Unit 1 Refueling Water Storage Tank level is 47% and lowering.

CCW trains are split and CANNOT be cross-tied.

1-8812B, RWST TO RHRP SUCT VLV is de-energized and OPEN.

With the current status of CCW, the Residual Heat Removal (RHR) system can be operated in what Modes of Emergency Core Cooling?

A. Train A RHR can be operated in the Injection Mode ONLY.

Train B RHR can be operated in BOTH the Injection and Recirculation Modes.

B. Train A RHR can be operated in the Injection Mode ONLY.

Train B RHR can be operated in the Injection Mode ONLY.

C. Train A RHR can be operated in the Injection and Recirculation Modes if the RHR trains are cross-tied.

Train B RHR can be operated in BOTH the Injection and Recirculation Modes.

D. Train A RHR can be operated in the Injection and Recirculation Modes if the RHR trains are cross-tied.

Train B RHR can be operated in the Injection Mode ONLY.

Proposed Answer: A Page 13 of 51 CPNPP 2014 NRC RO Written Exam Worksheet 01 to 10

ES-401 RO Written Exam Worksheet Form ES-401-5 Explanation:

A. Correct. When an RHR train is operating in Injection Mode, it pumps water from the Refueling Water Storage Tank (RWST). The temperature of the water from the RWST is < 120F, which is the limit for using a RHR pump (see EOS-1.3A, Transfer to Cold Leg Recirculation, Attachment 3, Bases for Step 2). Therefore, Train A RHR may only operate in the Injection Mode. Train B may be operated in any mode since CCW is still available to the Train B RHR system. The inability to remotely close 1-8812B does not preclude placing Train B in Cold Leg Recirculation, but does require actions later in EOS-1.3A to manually close the valve to place an additional barrier between the containment sump and the RWST.

B. Incorrect. Plausible because Train A RHR can only be operated in the Injection Mode, but Train B may be operated in any mode since CCW is still available to the Train B RHR system. The inability to remotely close 1-8812B does not preclude placing Train B in Cold Leg Recirculation, but does require actions later in EOS-1.3A to manually close the valve to place an additional barrier between the containment sump and the RWST.

C. Incorrect. Plausible because Train B RHR can be operated in the Injection or Recirculation Mode since it has CCW available, but Train A can only be operated in the Injection Mode without CCW available.

D. Incorrect. Plausible because both Trains of RHR can be operated in the Injection Mode, but Train B may be operated in any mode since CCW is still available to the Train B RHR system. The inability to remotely close 1-8812B does not preclude placing Train B in Cold Leg Recirculation, but does require actions later in EOS-1.3A to manually close the valve to place an additional barrier between the containment sump and the RWST.

Technical Reference(s) EOS-1.3A, Attachment 3, Step 2 Bases Attached w/ Revision: See FRC-0.1A, Step 1 CAUTION Comments / Reference EOS-1.3A, Step 3b EOS-1.3A, Step 5i Proposed references to be provided during examination: None Learning Objective: Given a procedural step, or sequence of steps from EOS-1.3, Transfer to Cold Leg Recirculation STATE the purpose/basis for the step(s).

Question Source: Bank ILOT5905 Modified Bank (Note changes or attach parent)

New Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 8 55.43 Page 14 of 51 CPNPP 2014 NRC RO Written Exam Worksheet 01 to 10

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

EOS-1.3A, Attachment 3, Step 2 Bases Revision: 8 Page 15 of 51 CPNPP 2014 NRC RO Written Exam Worksheet 01 to 10

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

FRC-0.1A, Step 1 CAUTION Revision: 8 Comments /

Reference:

EOS-1.3A, Step 3b Revision: 8 Page 16 of 51 CPNPP 2014 NRC RO Written Exam Worksheet 01 to 10

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

EOS-1.3A, Step 5i Revision: 8 Page 17 of 51 CPNPP 2014 NRC RO Written Exam Worksheet 01 to 10

ES-401 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. Date: 5/9/2014 Tier 1 Change: 5 Group 1 K/A 015/017 G 2.2.22 Level of Difficulty: 4 Importance Rating 4.0 RCP Malfunctions: Equipment Control: Knowledge of limiting conditions for operations and safety limits Proposed Question: 5 Given the following conditions:

Unit 1 is in MODE 3 following a Refueling outage.

All four Reactor Coolant System (RCS) Loops are OPERABLE, with all four Reactor Coolant Pumps (RCP) in operation.

Both Control Rod Drive Motor Generators are energized and Reactor Trip Breakers are CLOSED.

If RCP 1-02 trips, which of the following identifies the MINIMUM RCP requirements in accordance with LCO 3.4.5, RCS Loops -- MODE 3 and IPO-001A, Plant Heatup from Cold Shutdown to Hot Standby, Attachment 5, Checklist Required Prior to Closing Reactor Trip Breakers?

A. LCO 3.4.5 is satisfied with two RCPs OPERABLE, with one RCP in operation.

IPO-001A is satisfied with two RCPs in operation.

B. LCO 3.4.5 is satisfied with two RCPs OPERABLE, with two RCPs in operation.

IPO-001A is NOT satisfied with only three RCPs in operation.

C. LCO 3.4.5 is satisfied with two RCPs OPERABLE, with one RCP in operation.

IPO-001A is NOT satisfied with only three RCPs in operation.

D. LCO 3.4.5 is satisfied with two RCPs OPERABLE, with two RCPs in operation.

IPO-001A is satisfied with two RCPs in operation.

Proposed Answer: B Page 18 of 51 CPNPP 2014 NRC RO Written Exam Worksheet 01 to 10

ES-401 RO Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible because in MODE 3 with the Rod Control System capable of rod withdrawal two RCS loops are required operable with two in operation, however, only one is required in operation if the Rod Control System is not capable of rod withdrawal. IPO-001A only requires two RCS loops remain in operation in MODE 3 if the Rod Control System is not capable of rod withdrawal. This answer would be correct if the Rod Control System were not capable of rod withdrawal in the current conditions.

B. Correct. In MODE 3 with the Rod Control System capable of rod withdrawal, two RCS loops shall be OPERABLE with 2 loops in operation IAW TS 3.4.5 LCO. IPO-001A requires that four RCS loops be in operation with the Rod Control System capable of rod withdrawal and thus IPO-001A is not satisfied.

C. Incorrect. Plausible because in MODE 3 with the Rod Control System capable of rod withdrawal two RCS loops are required operable with two in operation, however, only one is required in operation if the Rod Control System is not capable of rod withdrawal. IPO-001A requires that four RCS loops be in operation with the Rod Control System capable of rod withdrawal and thus IPO-001A is not satisfied.

D. Incorrect. Plausible because in MODE 3 with the Rod Control System capable of rod withdrawal, two RCS loops shall be OPERABLE with 2 loops in operation IAW TS 3.4.5 LCO.. IPO-001A only requires two RCS loops remain in operation in MODE 3 if the Rod Control System is not capable of rod withdrawal. This answer would be correct if the Rod Control System were not capable of rod withdrawal in the current conditions.

Technical Reference(s) Technical Specification LCO 3.4.5 Attached w/ Revision: See IPO-001A, Attachment 5 Comments / Reference Proposed references to be provided during examination: None Learning Objective: APPLY the administrative requirements of the Reactor Coolant system including Technical Specifications, TRM and ODCM.

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

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 Page 19 of 51 CPNPP 2014 NRC RO Written Exam Worksheet 01 to 10

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

IPO-001A, Attachment 5 Revision: 22 Page 20 of 51 CPNPP 2014 NRC RO Written Exam Worksheet 01 to 10

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

IPO-001A, Attachment 5 Revision: 22 Page 21 of 51 CPNPP 2014 NRC RO Written Exam Worksheet 01 to 10

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

Technical Specification LCO 3.4.5 Amendment: 161 Page 22 of 51 CPNPP 2014 NRC RO Written Exam Worksheet 01 to 10

ES-401 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. Date: 5/9/2014 Tier 1 Change: 4 Group 1 K/A 025 AK2.01 Level of Difficulty: 3 Importance Rating 2.9 Loss of RHR System: Knowledge of the interrelations between the Loss of RHR System and the following: RHR heat exchangers Proposed Question: 6 Given the following:

Unit 1 is in MODE 4 entering a Refueling outage.

Residual Heat Removal (RHR) Train A is in Shutdown Cooling Mode.

Which of the following would result in the loss of Train A RHR heat removal capability?

A. A loss of power to 1-HV-606, U1 RHR HX 1-01 FLO CTRL VLV.

B. Closing 1-HV-4572, RHR HX 1-01 CCW RET VLV.

C. A loss of power to 1-FCV-618, RHR HX 1-01 BYP FLO CTRL VLV.

D. Closing 1CC-0109, RHR HX 1-01 CCW SPLY ISO VLV.

Proposed Answer: B Explanation:

A. Incorrect. Plausible if thought that the RHR Heat Exchanger Flow Control Valve will close on a loss of power. However, this valve fails open on a loss of power allowing maximum flow through the RHR Heat Exchanger and thus increases the RHR heat removal capability.

B. Correct. As the plant is in Shutdown Cooling Mode, Component Cooling Water flow has been adjusted to RHR Heat Exchanger 01. Closing the CCW return valve would remove RHR heat removal capability.

C. Incorrect. Plausible if thought that the RHR Heat Exchanger Bypass Flow Control Valve will open on a loss of power. However, this valve fails closed on a loss of power which forces all RHR flow through the RHR Heat Exchanger and thus increases the RHR heat removal capability.

D. Incorrect. Plausible because a typical isolation valve would terminate flow when closed and result in a complete loss of heat removal capability. However, the CCW Inlet Isolation valve for the RHR Heat Exchangers are of special design with an orifice in the disc that allows sufficient flow for the RHR system to meet design basis accident criteria when closed. Thus closing of this valve would reduce the flow of CCW through the RHR Heat Exchanger but would not result in a loss of RHR heat removal capability.

Technical Reference(s) LO21.SYS.RH1, Page 17 Attached w/ Revision: See SOP-102A, Step 5.4.B Comments / Reference SOP-102A, Step 2.1.C.1)

Page 23 of 51 CPNPP 2014 NRC RO Written Exam Worksheet 01 to 10

ES-401 RO Written Exam Worksheet Form ES-401-5 Proposed references to be provided during examination: None Learning Objective: DEMONSTRATE an understanding of the components of the Residual Heat Removal system including interrelations with other systems to include interlocks and control loops.

Question Source: Bank Modified Bank ILOT8002 (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 Page 24 of 51 CPNPP 2014 NRC RO Written Exam Worksheet 01 to 10

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From LO21.SYS.RH1, Page 17 Revision: 4-14-2011 Page 25 of 51 CPNPP 2014 NRC RO Written Exam Worksheet 01 to 10

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SOP-102A , Step 5.4.B Revision: 19 Page 26 of 51 CPNPP 2014 NRC RO Written Exam Worksheet 01 to 10

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

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From SOP-102A, Step 2.1.C.1) Revision: 19 Page 27 of 51 CPNPP 2014 NRC RO Written Exam Worksheet 01 to 10

ES-401 RO Written Exam Worksheet Form ES-401-5 Original Question: CPNPP Exam Bank ILOT8002 During a Design Basis Accident, which of the following would prevent the Residual Heat Removal Heat Exchanger 1-01 from performing its design function?

A. A loss of air to 1-HV-606, U1 RHR HX 1-01 FLO CTRL VLV..

B. Closing 1-HV-4572, RHR HX 1 CCW Return Valve C. Closing 1-HCV-0128, U1 RHR LTDN FLO CTRL VLV.

D. A loss of air to 1-FCV-618, RHR HX 1-01 BYP FLO CTRL VLV.

Answer: B Page 28 of 51 CPNPP 2014 NRC RO Written Exam Worksheet 01 to 10

ES-401 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. Date: 5/9/2014 Tier 1 Change: 4 Group 1 K/A 026 AA1.06 Level of Difficulty: 3 Importance Rating 2.9 Loss of Component Cooling Water: Ability to operate and/or monitor the following as they apply to a Loss of Component Cooling Water: Control of flow rates to components cooled by the CCWS Proposed Question: 7 Given the following conditions:

Unit 1 is in MODE 3 with Train A Residual Heat Removal System operating in the Shutdown Cooling Mode.

1-ALB-3B, Window 4.5 - CCW HX 1/2 SPLY FLO LO is LIT.

While performing actions in accordance with ALM-0032A, Alarm Procedure 1-ALB-3B, the crew observes the following:

CCW Pump 1-01 Discharge Pressure (1-PI-4520): 140 psig.

CCW HX 1-01 Outlet Flow (1-FI-4536A): 8000 gpm.

CCW HX 1-01 Recirculation Flow (1-FI-4536B): 8000 gpm.

CCW HX 1-01 Outlet Temperature (1-TI-4530): 100F.

CCW Surge Tank Level (1-LI-4500): 68%.

Which of the following lists the action that should be performed in response to the CCW parameters in accordance with ALM-0032A?

A. Start CCWP 1-02 to share heat load between trains.

B. Start CCWP 1-02 and secure CCWP 1-01 for pump protection.

C. Perform ABN-502, Component Cooling Water System Malfunction.

D. Close 1-HS-4536, CCWP 1-01 Recirculation Valve.

Proposed Answer: D Page 29 of 51 CPNPP 2014 NRC RO Written Exam Worksheet 01 to 10

ES-401 RO Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible because it could be thought that 100°F at the heat exchanger outlet is too high, however, the alarm for this action is at 118°F.

B. Incorrect. Plausible because if thought that 100°F would require swapping CCWPs this action is directed at 120°F.

C. Incorrect. Plausible because if total flow was less than 10,000 gpm action required is a transition to ABN-502 for either a pump trip or loss of flow. ABN-502 does not have actions for a Recirc valve being open; this action is addressed per the ALM.

D. Correct. IAW ALM-0032A with total flow > 15,500 gpm with the Recirc valve open the expected action is to close the Recirc valve.

Technical Reference(s) ALM-0032A, Window 4.5, Logic Diagram Attached w/ Revision: See ALM-0032A, Window 4.5, Steps 2, 3, & 4 Comments / Reference ALM-0032A, Window 1.5, Logic Diagram ALM-0032A, Window 1.5, Step 1 & 1.B Proposed references to be provided during examination: None Learning Objective: COMPREHEND the normal, abnormal and emergency operation of the Component Cooling Water system.

ANALYZE the response to Loss of All CCW Flow in accordance with ABN-502, Component Cooling Water System Malfunction.

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 30 of 51 CPNPP 2014 NRC RO Written Exam Worksheet 01 to 10

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

ALM-0032A, Window 4.5, Logic Diagram Revision: 7 Page 31 of 51 CPNPP 2014 NRC RO Written Exam Worksheet 01 to 10

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

ALM-0032A, Window 4.5, Steps 2, 3, & 4 Revision: 7 Page 32 of 51 CPNPP 2014 NRC RO Written Exam Worksheet 01 to 10

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

ALM-0032A, Window 1.5, Logic Diagram Revision: 7 Page 33 of 51 CPNPP 2014 NRC RO Written Exam Worksheet 01 to 10

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

ALM-0032A, Window 1.5, Step 1 & 1.B Revision: 7 Page 34 of 51 CPNPP 2014 NRC RO Written Exam Worksheet 01 to 10

ES-401 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. Date: 5/10/2014 Tier 1 Change: 4 Group 1 K/A 027 AK2.03 Level of Difficulty: 3 Importance Rating 2.6 Pressurizer Pressure Control Malfunction: Knowledge of the interrelations between Pressurizer Pressure Control Malfunctions and the following: Controllers and positioners Proposed Question: 8 Given the following conditions:

Unit 2 is at 100% power.

2-PK-455, PRZR MASTER PRESS CTRL has failed to 40% demand.

With no operator action, how does the Pressurizer Pressure Control System respond?

Pressurizer Spray valves close, Pressurizer pressure rises and...

A. ...PRZR PORV 2/1-PCV-456 opens and cycles open and closed around setpoint.

B. ...PRZR PORV 2/1-PCV-456 opens and then closes at 2185 psig.

C. ...PRZR PORV 2/1-PCV-455A opens and cycles open and closed around setpoint.

D. ...PRZR PORV 2/1-PCV-455A opens and then closes at 2185 psig.

Proposed Answer: A Explanation:

A. Correct. With all heaters energized and both spray valves incapable of opening, pressurizer pressure will rise to 2335 psig and PORV 456 will open and then close at the reset setpoint, around 2315 psig.

B. Incorrect. Plausible because with all heaters energized and both spray valves incapable of opening, pressurizer pressure will rise to 2335 psig and PORV 456 will open, the interlock with channel 457 would close PORV 456 at 2185 psig, however the PORV will actually reclose at the reset setpoint, approximately 2315 psig.

C. Incorrect. Plausible because it could be thought that with all heaters energized and both spray valves incapable of opening, pressurizer pressure will rise to 2335 psig and PORV 455A will open and then close at the reset setpoint, approximately 2315 psig, however with the master pressure controller at 40% demand PORV 455A is incapable of automatically opening.

D. Incorrect. Plausible because with all heaters energized and both spray valves incapable of opening, pressurizer pressure will rise to 2335 psig and PORV 455A would open and the interlock with channel 458 would close PORV 455A at 2185 psig, however the PORV would actually reclose at the reset setpoint, approximately 2315 psig. With the master pressure controller at 40% demand PORV 455A is incapable of automatically opening.

Page 35 of 51 CPNPP 2014 NRC RO Written Exam Worksheet 01 to 10

ES-401 RO Written Exam Worksheet Form ES-401-5 Technical Reference(s) LO21.SYS.PP1, Pages 7, 11 & 12 Attached w/ Revision: See Comments / Reference Proposed references to be provided during examination: None Learning Objective: ANALYZE the response to a Pressurizer Pressure Instrument Malfunction in accordance with ABN-705, Pressurizer Pressure Malfunction.

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 36 of 51 CPNPP 2014 NRC RO Written Exam Worksheet 01 to 10

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

LO21.SYS.PP1 Lesson Plan Revision: 9/4/13 Page 37 of 51 CPNPP 2014 NRC RO Written Exam Worksheet 01 to 10

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

LO21.SYS.PP1 Lesson Plan Revision: 9/4/13 The controlling pressure channel failing low is similar to the master pressure controller demand failing at 40% demand.

Page 38 of 51 CPNPP 2014 NRC RO Written Exam Worksheet 01 to 10

ES-401 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. Date: 5/10/2014 Tier 1 Change: 3 Group 1 K/A 029 EK1.01 Level of Difficulty: 3 Importance Rating 2.8 ATWS: Knowledge of the operational implications of the following concepts as they apply to the ATWS: Reactor nucleonics and thermal-hydraulics behavior Proposed Question: 9 Given the following conditions:

Unit 1 was operating at 100% power, MOL.

Both Main Feedwater Pumps tripped.

An automatic Reactor Trip did NOT occur.

Attempts to manually trip the Reactor have NOT been successful.

The Turbine has been manually tripped.

Both Motor Driven Auxiliary Feedwater Pumps are feeding all four Steam Generators at 1200 gpm.

The Turbine Driven Auxiliary Feedwater Pump has tripped and CANNOT be reset.

Which of the following describes the expected Reactor core and Pressurizer pressure response prior to locally tripping the Reactor?

A. Total core power LOWERS due to Moderator Temperature Coefficient.

Pressurizer pressure LOWERS due to Pressurizer PORVs and Safeties opening.

B. Total core power LOWERS due to Moderator Temperature Coefficient.

Pressurizer pressure RISES due to the available heat removal capability.

C. Total core power RISES due to Power Coefficient (Power Defect).

Pressurizer pressure LOWERS due to Pressurizer PORVs and Safeties opening.

D. Total core power RISES due to Power Coefficient (Power Defect).

Pressurizer pressure RISES due to the available heat removal capability.

Proposed Answer: B Page 39 of 51 CPNPP 2014 NRC RO Written Exam Worksheet 01 to 10

ES-401 RO Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible because as RCS density lowers due to decreased heat removal the MTC adds negative reactivity to the core causing total core power to lower and the effect from pressurizer spray on pressurizer pressure is negligible compared to the pressure rise from the heat imbalance.

The PORVs may cycle but the safeties would not open.

B. Correct. As RCS density lowers due to decreased heat removal the MTC adds negative reactivity to the core causing total core power to lower. Pressurizer pressure rises due to the imbalance between the heat source (reactor) and the heat sink (steam generators) that are being fed at well below the capability of the MFPs.

C. Incorrect. Plausible because the power coefficient will add positive reactivity to the core, however, the effect from the negative reactivity from MTC cause total core power to lower and the effect from pressurizer spray on pressurizer pressure is negligible compared to the pressure rise from the heat imbalance. The PORVs may cycle but the safeties would not open.

D. Incorrect. Plausible because the power coefficient will add positive reactivity to the core, however, the effect from the negative reactivity from MTC cause total core power to lower. Pressurizer pressure rises due to the imbalance between the heat source (reactor) and the heat sink (steam generators) that are being fed at well below the capability of the MFPs.

Technical Reference(s) LO21.MCO.MI5, Pages 12 & 15 Attached w/ Revision: See LO21.GFR.COF, Pages 4, 8 & 20 Comments / Reference Proposed references to be provided during examination: None Learning Objective: ANALYZE the Anticipated Transient Without Trip analysis.

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 55.43 Page 40 of 51 CPNPP 2014 NRC RO Written Exam Worksheet 01 to 10

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

LO21.GFR.COF, Page 4 Revision: 11/29/07 Page 41 of 51 CPNPP 2014 NRC RO Written Exam Worksheet 01 to 10

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

LO21.GFR.COF, Page 8 Revision: 11/29/07 Page 42 of 51 CPNPP 2014 NRC RO Written Exam Worksheet 01 to 10

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

LO21.GFR.COF, Page 20 Revision: 11/29/07 Page 43 of 51 CPNPP 2014 NRC RO Written Exam Worksheet 01 to 10

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

LO21.MCO.MI5, Page 12 Revision: 01/09/12 LOSS OF FEEDWATER ATWT CAUSES Loss of normal feedwater could result from a malfunction in the condensate system or its control system from such causes as a trip of a condensate pump, simultaneous trip of both main feedwater pumps (or closure of their discharge valves), or simultaneous closure of all feedwater control valves.

The vast majority of these cases would cause only a partial loss of feedwater flow. The most likely cause of a complete loss of feedwater would be loss of station power.

TRANSIENT DESCRIPTION The loss of feedwater produces a large imbalance in the heat source/ heat sink relationship. When feedwater flow to the SGs is terminated, the secondary system can no longer remove all of the heat that is generated in the reactor core. This heat buildup in the primary system is indicated by rising RCS temperature and pressure, and by increasing pressurizer water level, which is due to the insurge of expanding reactor coolant. Water level in the SGs drops as the remaining water in the secondary system is boiled off. When SG level falls to the point where the SG tubes are exposed and the primary-to-secondary heat transfer is reduced, the reactor coolant temperature and pressure begin to increase at a greater rate. This greater rate of increase is maintained as the pressurizer fills and releases water through the PORVs and safety valves. (The PORVs and safety valves have a smaller volumetric relief capacity for water than for steam.) Negative reactivity feedback, from the moderator temperature coefficient as primary temperature rises, reduces core power. Pressure begins to decrease and a steam bubble is again formed in the pressurizer. The event time line is shown in Table 1.

Page 44 of 51 CPNPP 2014 NRC RO Written Exam Worksheet 01 to 10

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

LO21.MCO.MI5, Page 15 Revision: 01/09/12 SENSITIVITY STUDIES (PARAMETRIC VARIATIONS)

The Loss of Feedwater ATWT was also subjected to sensitivity studies which analyzed changes in significant assumptions and parameters to determine their effect on RCS over pressure. The results of these studies are discussed below.

Effect of Not Tripping the Turbine Failure to trip the turbine permitted a higher steam release from the SGs. In addition, more heat was removed from the primary system early in the transient. The core power level stayed relatively high and the primary pressure attained a higher maximum value than for the case in which the turbine was tripped.

Effect of Not Opening the PORVs With only the three pressurizer safety valves available for steam and water relief, the peak pressures attained increased by about 9 percent.

Effect of Not Using Pressurizer Spray Addition of spray water into the pressurizer steam space decreased early in the transient for the base case. Thus, the use of no pressurizer spray did not significantly affect the peak pressure reached during the transient.

Effect of using Automatic Rod Control Allow the automatic insertion of control rods to compensate for rising reactor coolant temperature reduced the coolant expansion rate to the point that the pressurizer safety valves easily relieved the coolant insurge without even reaching the full open position (at 2590 psia). Also, the temperature and pressure transients were controlled to the extent that the reactor coolant pumps (RCPs) did not cavitate and the PORVs were able to limit the pressurizer pressure to about 2350 psia. When the pressurizer filled and water was released through the valves, pressure rose rapidly.

Effect of Variation in Initial Average Coolant Temperature For the purpose of determining the effect of initial average coolant temperature on the Loss of Feedwater ATWT, and in order to encompass the average coolant temperatures of a variety of Westinghouse plants, analyses were conducted assuming +8°F and -20°F variation in initial average coolant temperature.

Effect of Variation in Initial Pressurizer Level Variation of +/- 10 percent in initial pressurizer level was considered. The higher water level meant that the pressurizer filled to capacity earlier in the transient when the core power level was still relatively high. A lower than normal water level delayed the filling of the pressurizer and provided more steam for volumetric relief through the valves and resulted in a lower pressurizer pressure.

Page 45 of 51 CPNPP 2014 NRC RO Written Exam Worksheet 01 to 10

ES-401 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. Date: 5/9/2014 Tier 1 Change: 4 Group 1 K/A 040 AK3.06 Level of Difficulty: 2 Importance Rating 3.4 Steam Line Rupture: Knowledge of the reasons for the following responses as they apply to the Steam Line Rupture:

Containment temperature and pressure considerations Proposed Question: 10 Following a Steam Line Break accident inside Containment, the operator is expected to stop Auxiliary Feedwater flow to the faulted Steam Generator within 10 minutes in order to prevent exceeding...

A. ...Motor Driven Auxiliary Feedwater Pump capacity.

B. ...Containment design temperature.

C. ...Steam Generator U-tube stress limits.

D. ...Steam Generator tubesheet stress limits.

Proposed Answer: B Explanation:

A. Incorrect. Plausible because it could be thought that isolating AFW flow would prevent exceeding the capacity of the AFW pumps due to the reduced resistance to flow from a faulted SG; however, orifices in the AFW feed lines to the SGs limit flow.

B. Correct. STI-214.01 lists isolating AFW flow to a faulted SG inside CNTMT as a timed operator action that is taken within 10 minutes to ensure CNTMT design temperature is not exceeded.

C. Incorrect. Plausible because it could be thought that stopping AFW flow is to limit stress on the SG U-tubes; however, the Steam Generator flow restrictor at the top of the SG is designed to limit stress on the U-tubes.

D. Incorrect. Plausible because it could be thought that stopping AFW flow is to limit stress on the SG tubesheet; however, the Steam Generator flow restrictor at the top of the SG is designed to limit stress on the tubesheet.

Technical Reference(s) STI-214.01, Attachment 8.A Attached w/ Revision: See LO21.SYS.AF1, Page 12 Comments / Reference LO21.SYS.MR1, Page 8 Proposed references to be provided during examination: None Learning Objective: APPLY the administrative requirements of the Auxiliary Feedwater system including Technical Specifications, TRM and ODCM.

Page 46 of 51 CPNPP 2014 NRC RO Written Exam Worksheet 01 to 10

ES-401 RO Written Exam Worksheet Form ES-401-5 Question Source: Bank Modified Bank LORT0662 (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 Page 47 of 51 CPNPP 2014 NRC RO Written Exam Worksheet 01 to 10

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

STI-214.01, Attachment 8.A Revision: 0 Page 48 of 51 CPNPP 2014 NRC RO Written Exam Worksheet 01 to 10

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

LO21.SYS.AF1, Page 12 Revision: 05/11/11 Page 49 of 51 CPNPP 2014 NRC RO Written Exam Worksheet 01 to 10

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

LO21.SYS.MR1, Page 8 Revision: 09/04/13 Steam Generator Flow Restrictor Each SG outlet nozzle contains an integral flow restrictor made up of seven (7) venturi type flow nozzles having a total flow area of 1.388 ft2. During normal operation, the flow restrictor provides a 2 to 3 psi P for steam flow measurement.

Under accident conditions, a double ended rupture (DER) of the main steam line would theoretically expose a 4.6 ft2 break area to be available for blowdown flow based upon main steam line piping size.

However, the integral flow restrictor limits the maximum break area per SG to 1.388 ft2. From a safety analysis perspective, limiting break size and thus break flow provides several protective advantages:

Protects fuel integrity by limiting the primary system cooldown rate thereby reducing the reactivity addition rate ensuring a departure from nucleate boiling (DNB) condition does not occur during the reactors return to power.

Protects containment integrity by limiting the containment temperature and pressure rise.

Reduces thrust forces on main steam line piping.

Limits stresses on internal SG components, particularly the tubesheet and U-tubes.

Page 50 of 51 CPNPP 2014 NRC RO Written Exam Worksheet 01 to 10

ES-401 RO Written Exam Worksheet Form ES-401-5 Original Question: CPNPP Exam Bank LORT0662 For a steam line break accident inside Containment, the operator is expected to take manual action to stop AFW flow to the faulted SG within (1) after the initiation of the break in order to prevent exceeding (2) .

(1) (2)

A. 60 seconds pump flow capacity B. 600 seconds Containment internal pressure C. 60 minutes SG tubesheet delta T D. 600 minutes Containment flooding level Answer: B Page 51 of 51 CPNPP 2014 NRC RO Written Exam Worksheet 01 to 10

ES-401 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. Date: 3/3/2014 Tier 1 Change: 2 Group 1 K/A 055 EK3.02 Level of Difficulty: 3 Importance Rating 4.3 Station Blackout: Knowledge of the reasons for the following responses as they apply to the Station Blackout: Actions contained in EOP for loss of offsite and onsite power Proposed Question: 11 Which of the following describes the reasons for depressurizing the Steam Generators to 270 psig in accordance with ECA-0.0A, Loss of All AC Power?

A. Initiates Safety Injection System Accumulator discharge and minimizes Reactor Coolant Pump seal leakage.

B. Establishes Natural Circulation conditions and initiates Safety Injection System Accumulator discharge.

C. Establishes Natural Circulation conditions and minimizes secondary heat sink requirements if Auxiliary Feedwater inventory is limited.

D. Minimizes secondary heat sink requirements if Auxiliary Feedwater inventory is limited and minimizes RCP seal leakage.

Proposed Answer: A Explanation:

A. Correct. Lowering RCS pressure and restoring lost inventory is the reason for depressurizing.

B. Incorrect. Plausible because RCS depressurization will assist Natural Circulation, but is not the reason for depressurization to 270 psig.

C. Incorrect. Plausible because Natural Circulation will be established as a byproduct of rapid depressurization. Rapid cooldown and depressurization due to limited AFW is an action that could be taken in E-3 series procedures.

D. Incorrect. Plausible because in E-3 series procedures, rapid secondary depressurizations may be performed when there is limited makeup availability.

Technical Reference(s) ECA-0.0A, Step 18 Attached w/ Revision: See ECA-0.0A, Attachment 7, Step 18 Bases Comments / Reference Proposed references to be provided during examination: None Learning Objective: STATE the bases for operator actions, notes and cautions from ECA-0.0, Loss of All AC Power.

Page 1 of 54 CPNPP 2014 NRC RO Written Exam Worksheet 11 to 20

ES-401 RO Written Exam Worksheet Form ES-401-5 Question Source: Bank ILOT6035 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 4 55.43 Comments /

Reference:

ECA-0.0A, Step 18 Revision: 8 Page 2 of 54 CPNPP 2014 NRC RO Written Exam Worksheet 11 to 20

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

ECA-0.0A, Attachment 7, Step 18 Bases Revision: 8 Page 3 of 54 CPNPP 2014 NRC RO Written Exam Worksheet 11 to 20

ES-401 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. Date: 5/10/2014 Tier 1 Change: 5 Group 1 K/A 056 AA2.17 Level of Difficulty: 4 Importance Rating 3.4 Loss of Offsite Power: Ability to determine and interpret the following as they apply to the Loss of Offsite Power: Operational status of PZR backup heaters Proposed Question: 12 Given the following conditions:

A Loss of Offsite Power has occurred on Unit 1.

During plant recovery, the crew resets the Blackout Sequencer.

Current Pressurizer pressure is 2200 psig and slowly lowering Which of the following describes how Pressurizer Heater control is restored under these conditions?

The breakers for Pressurizer Heater Banks A. A, B and D must be closed locally; Control Bank C power is restored with NO additional operator action.

B. A, B and D must be closed locally; Control Bank C must be momentarily placed to ON.

C. A, B and D will close if left in AUTO or ON; Control Bank C power is restored with NO additional operator action.

D. A, B and D will close if left in AUTO or ON; Control Bank C must be momentarily placed to ON.

Proposed Answer: D Page 4 of 54 CPNPP 2014 NRC RO Written Exam Worksheet 11 to 20

ES-401 RO Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible if thought that automatic control is lost when the Blackout Sequencer (BOS) actuates, however, automatic control is restored when the BOS is reset.

B. Incorrect. Plausible if thought that automatic control is lost when the Blackout Sequencer (BOS) actuates, however, automatic control is restored when the BOS is reset. Placing Control Bank C momentarily in ON is the correct action.

C. Incorrect. Plausible when the BOS is reset, the backup heaters will operate as stated. Additionally, without a demand signal present. Control Bank C must be momentarily placed in ON.

D. Correct. When the BOS is reset, the backup heaters will operate as stated.

Technical Reference(s) LO21.SYS.PP1, Pages 7 & 8 Attached w/ Revision: See ABN-602, Step 10.3.9 NOTE Comments / Reference ABN-602, Attachment 1 Proposed references to be provided during examination: None Learning Objective: DEMONSTRATE an understanding of the components of the Pressurizer Pressure and Level Control System including interrelations with other systems to include interlocks and control loops.

Question Source: Bank ILOT1967 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 Page 5 of 54 CPNPP 2014 NRC RO Written Exam Worksheet 11 to 20

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

LO21.SYS.PP1, Page 7 Revision: 05/05/11 The compensated error signal from the master pressure controller controls spray valves, heaters, and PORV u-PCV-0455A in order to drive pressure to the 2235 psig setpoint. The actual pressure at which the spray valves, heaters or PORV operate is based upon the amount of time that actual pressure is off setpoint. Time passing with pressure off setpoint causes the compensated error signal to continue to increase even if the actual pressure deviation is constant. Controller output automatically changes, operating equipment as necessary to bring pressure back to 2235 psig. The master pressure controller is manually operated by setting the output signal to operate the desired equipment. The following table equates equipment operation with controller output and uncompensated error.

ACTION OUTPUT % ERROR SIGNAL NOMINAL VALUE PORV opens 81.3 +100 psig 2335 psig PORV closes 75.0 +80 psig 2315 psig Spray valves fully open 73.4 +75 psig 2310 psig Spray valves start to open 57.8 +25 psig 2260 psig Variable heaters off 54.7 +15 psig 2250 psig Normal operating pressure 50.0 2235 psig Variable heaters fully on 45.3 15 psig 2220 psig Backup heaters off 44.7 17 psig 2218 psig Backup heaters on 42.2 25 psig 2210 psig Page 6 of 54 CPNPP 2014 NRC RO Written Exam Worksheet 11 to 20

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

LO21.SYS.PP1, Page 8 Revision: 05/05/11 Pressurizer Heaters The PRZR heaters consist of 78 elements mounted vertically in the bottom of the PRZR. The PRZR heaters are divided into four groups identified as A, B, C, and D. Groups A and B each have 21 heater elements and a heat capacity of 485 KW. Groups C and D each have 18 heater elements and a heat capacity of 416 KW. The total heater capacity is 1802 KW.

Groups A, B, and D are called backup heaters. The backup heaters are energized by closing their power supply breakers in switchgear uEB2, uEB3 and uEB4. Each group has a 3-position maintained (OFF-AUTO-ON) handswitch located on u-CB-05. The backup heater power supply breaker is closed by placing the handswitch in ON or by a low pressure signal from the master pressure controller when the handswitch is in AUTO. Backup heaters in AUTO will also be energized by pressurizer level deviation of 5% above program level. Groups A and B may be operated from the Remote Shutdown Panel.

Group C is the control heaters, also called variable or proportional heaters. These heaters operate with variable output controlled by the master pressure controller. A 3-position (OFF-neutral-ON) spring-return to center handswitch operates the control heaters from u-CB-05. During normal operation, the handswitch is taken to the ON position, closing the power supply breaker in switchgear uEB1, and released to the center position. A silicon controlled rectifier (SCR) circuit supplies power to the heater elements using a time-proportioned average output voltage based on the control signal from the master pressure controller. This means that a full 480 VAC is supplied to the heaters in pulses such that the average voltage supplied over time is proportional to the pressure controller output. The control heater power supply breaker will not close automatically.

Comments /

Reference:

ABN-602, Step 10.3.9 NOTE Revision: 8 Page 7 of 54 CPNPP 2014 NRC RO Written Exam Worksheet 11 to 20

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

ABN-602, Attachment 1 Revision: 8 Page 8 of 54 CPNPP 2014 NRC RO Written Exam Worksheet 11 to 20

ES-401 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. Date: 5/9/2014 Tier 1 Change: 4 Group 1 K/A 058 AA1.03 Level of Difficulty: 3 Importance Rating 3.1 Loss of DC Power: Ability to operate and/or monitor the following as they apply to the Loss of DC Power: Vital and battery bus components Proposed Question: 13 Given the following conditions:

Unit 2 is at 100% power.

A loss of 2ED1, 125 VDC Switch Panel occurs.

Which of the following are the correct component responses?

A. Emergency Diesel Generator 2-01 CANNOT be started.

The Feedwater Isolation Valves fail close.

B. Emergency Diesel Generator 2-01 CANNOT be started.

The Feedwater Control Valves fail close.

C. Emergency Diesel Generator 2-01 auto starts.

The Feedwater Isolation Valves fail close.

D. Emergency Diesel Generator 2-01 auto starts.

The Feedwater Control Valves fail close.

Proposed Answer: B Page 9 of 54 CPNPP 2014 NRC RO Written Exam Worksheet 11 to 20

ES-401 RO Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible because DG 2-01 cannot be started as power to the start air solenoids does not exist. The Feedwater Isolation Valves do not fail close on the loss of 2ED1, but are plausible as they fail close on the loss of 2EC1.

B. Correct. DG 2-01 cannot be started as power to the start air solenoids does not exist. The Feedwater Control Valves fail closed on a loss of 2ED1 and result in a reactor trip occurring.

C. Incorrect. Plausible because the DG 2-01 start air solenoids do not have power and normally with a safety related component loss of power is to the fail safe position which would open and auto start the DG 2-01, however a loss of power prevents the DG from starting. The Feedwater Isolation Valves do not fail close on the loss of 2ED1, but are plausible as they fail close on the loss of 2EC1.

D. Incorrect. Plausible because the DG 2-01 start air solenoids do not have power and normally with a safety related component loss of power is to the fail safe position which would open and auto start the DG 2-01, however a loss of power prevents the DG from starting. The Feedwater Control Valves fail closed on a loss of 2ED1 and result in a reactor trip occurring.

Technical Reference(s) ALM-0102B, Window 1.13 Attached w/ Revision: See ABN-603, Section 3.1 Comments / Reference LO21.SYS.ED1, Page 34 Proposed references to be provided during examination: None Learning Objective: DEMONSTRATE an understanding of the components of the Solid State Protection System including interrelations with other systems to include interlocks and control loops.

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 6 55.43 Page 10 of 54 CPNPP 2014 NRC RO Written Exam Worksheet 11 to 20

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

ALM-0102B, Window 1.13 Revision: 6 Page 11 of 54 CPNPP 2014 NRC RO Written Exam Worksheet 11 to 20

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

ABN-603, Section 3.1 Revision: 8 Page 12 of 54 CPNPP 2014 NRC RO Written Exam Worksheet 11 to 20

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

LO21.SYS.ED1, Page 34 Revision: 5-2-2011 Page 13 of 54 CPNPP 2014 NRC RO Written Exam Worksheet 11 to 20

ES-401 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. Date: 5/9/2014 Tier 1 Change: 4 Group 1 K/A 062 AK3.02 Level of Difficulty: 4 Importance Rating 3.6 Loss of Nuclear Service Water: Knowledge of the reasons for the following responses as they apply to the Loss of Nuclear Service Water: The automatic actions (alignments) within the nuclear service water resulting from the actuation of the ESFAS Proposed Question: 14 Given the following conditions:

Unit 1 has experienced a complete Loss of Offsite and Onsite AC power.

The crew is performing ECA-0.0A, Loss of All AC Power.

Station Service Water (SSW) is NOT available from Unit 2, resulting in a complete loss of Unit 1 SSW.

While performing ECA-0.0A, Safety Injection actuated and was RESET.

The Safety Injection Sequencers were also RESET.

Power has been restored to Train A Safeguards Bus 1EA1 in accordance with ABN-601, Response to a 138/345 KV System Malfunction.

The Alternate Power Generators have been placed in service in accordance with SOP-614A, Alternate Power Generator Operation.

Power is being supplied to Train A Safeguards Bus 1EA1 from the Alternate Power Generators.

With NO operator action following the restoration of power, which of the following indicates the status of the SSW system when Bus 1EA1 is energized?

1-HV-4286, SSWP 1 DISCH VLV strokes from A. 10% to 100% OPEN and SSWP 1-01 STARTS.

B. 10% to 100% OPEN and SSWP 1-01 is in PULLOUT.

C. 100% to 10% OPEN and SSWP 1-01 STARTS.

D. 100% to 10% OPEN and SSWP 1-01 is in PULLOUT.

Proposed Answer: D Page 14 of 54 CPNPP 2014 NRC RO Written Exam Worksheet 11 to 20

ES-401 RO Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible because the SSW pump is normally left in AUTO after start during performance of ECA-0.0A, so that the pump will auto start when the bus is resupplied from the Diesel Generator. When power is restored to the bus from the APGs, the SSW pump is placed in PULL-OUT as the Component Cooling Water Pump is loaded on the bus prior to the SSW pump.

1-HV-4286 is a normally open valve which when in auto strokes from 10% to 100% coincident with the start of SSWP 1-01. Likewise the valve strokes from 100% to 10% when SSWP 1-01 is taken to either STOP or placed in PULL-OUT. Therefore it is plausible to think that the valve will stroke from 10% to 100% when the pump is started; however, the valve was full open when power was lost and would remain full open if the pump handswitch had remained in AUTO in accordance with ECA-0.0A.

B. Incorrect. Plausible because SSW Pump 1-01 would have been placed in PULL-OUT during performance of SOP-614A prior to energizing the bus from the APGs. 1-HV-4286 is a normally open valve which when in auto strokes from 10% to 100% coincident with the start of SSWP 1-01.

Likewise the valve strokes from 100% to 10% when SSWP 1-01 is taken to either STOP or placed in PULL-OUT. Therefore it is plausible to think that the valve will stroke from 10% to 100% with the pump handswitch in PULL-OUT, but in actuality the valve was full open when power was lost and thus will immediately stroke from 100% to 10% with the pump in PULL-OUT.

C. Incorrect. . Plausible because the SSW pump is normally left in AUTO after start during performance of ECA-0.0A, so that the pump will auto start when the bus is resupplied from the Diesel Generator. When power is restored to the bus from the APGs, the SSW pump is placed in PULL-OUT as the Component Cooling Water Pump is loaded on the bus prior to the SSW pump.

1-HV-4286 is a normally open valve which when in auto strokes from 10% to 100% coincident with the start of SSWP 1-01. Likewise the valve strokes from 100% to 10% when SSWP 1-01 is taken to either STOP or placed in PULL-OUT. Since the valve was full open when power was lost it is a misconception to believe that the valve circuitry would interpret the loss of power as stopping the SSWP and thus receive an auto close to 10% upon restoration of power to the bus.

D. Correct. SSW Pump 1-01 would have been placed in PULL-OUT during performance of SOP-614A prior to energizing the bus from the APGs. 1-HV-4286 is a normally open valve which when in auto strokes from 10% to 100% coincident with the start of SSWP 1-01. Likewise the valve strokes from 100% to 10% when SSWP 1-01 is taken to either STOP or placed in PULL-OUT. Since the valve was full open when power was lost and the pump handswitch is now in PULL-OUT the valve will immediately stroke from 100% to 10%.

Technical Reference(s) SOP-614A, Step 5.1.C, 5.1.J Attached w/ Revision: See ECA-0.0A, Attachment 7, Step 27 Bases Comments / Reference LO21.SYS.SW1, Page 18 Proposed references to be provided during examination: None Learning Objective: ANALYZE the response to a Loss of All Unit u Station Service Water in accordance with ABN-501, Station Service Water System Malfunction.

DISCUSS plant response, operator actions, and the reasons for the actions contained in ECA-0.0, Loss of All AC Power.

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

New X Question History: Last NRC Exam Page 15 of 54 CPNPP 2014 NRC RO Written Exam Worksheet 11 to 20

ES-401 RO Written Exam Worksheet Form ES-401-5 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 Comments /

Reference:

SOP-614A, Step 5.1.C Revision: 13 Page 16 of 54 CPNPP 2014 NRC RO Written Exam Worksheet 11 to 20

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

SOP-614A, Step 5.1.J Revision: 13 Page 17 of 54 CPNPP 2014 NRC RO Written Exam Worksheet 11 to 20

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

ECA-0.0A, Attachment 7, Step 27 Bases Revision: 8 Page 18 of 54 CPNPP 2014 NRC RO Written Exam Worksheet 11 to 20

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

LO21.SYS.SW1, Page 18 Revision: 4-28-2011 Page 19 of 54 CPNPP 2014 NRC RO Written Exam Worksheet 11 to 20

ES-401 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. Date: 5/10/2014 Tier 1 Change: 4 Group 1 K/A 065 G 2.4.11 Level of Difficulty: 3 Importance Rating 4.0 Loss of Instrument Air: Emergency Procedures/Plan: Knowledge of abnormal condition procedures Proposed Question: 15 Given the following conditions:

Unit 1 is at 100% power.

A break in the Instrument Air System has occurred.

ABN-301, Instrument Air System Malfunction is being implemented.

All available Air Compressors are running.

Instrument Air header pressure is 33 psig.

Control of systems has NOT been lost.

Which of the following describes the required actions for this condition in accordance with ABN-301?

A. Continue efforts to restore Instrument Air pressure and simultaneously initiate a Manual Turbine Runback per ABN-302, Feedwater, Condensate, Heater Drain System Malfunction.

B. Trip the Reactor and enter EOP-0.0A, Reactor Trip or Safety Injection. Continue in ABN-301, Instrument Air System Malfunction, and take actions to control Charging flow locally.

C. Open 1CI-0050, INST AIR RCVR 1-01 U2 XTIE VLV and simultaneously initiate a Manual Turbine Runback per ABN-302, Feedwater, Condensate, Heater Drain System Malfunction.

D. Continue efforts to restore Instrument Air pressure and when instrument air header pressure reaches 25 psig, trip the Reactor and enter EOP-0.0A, Reactor Trip or Safety Injection.

Proposed Answer: B Page 20 of 54 CPNPP 2014 NRC RO Written Exam Worksheet 11 to 20

ES-401 RO Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible because loss of Instrument Air has a significant impact on the secondary plant and it could be thought that rapidly reducing power would minimize the effects, however, ABN-301 requires a Unit trip when pressure reaches 35 psig.

B. Correct. ABN-301, Instrument Air Malfunction, requires a Unit trip when pressure reaches 35 psig.

The EOP is entered but the actions of ABN-301, Instrument Air Malfunction are still performed which includes local control of Charging flow due to failed open valves.

C. Incorrect. Plausible because loss of Instrument Air has a significant impact on the secondary plant and it could be thought that rapidly reducing power would minimize the effects, however, ABN-301 requires a Unit trip when pressure reaches 35 psig. And when instrument air header pressure falls below 85 psig, ABN-301 requires closing of 1CI-0050.

D. Incorrect. Plausible if thought that the pressure criteria for tripping were 25 psig and the actions to trip at that point would be correct.

Technical Reference(s) ABN-301, Steps 2.3.2, 2.3.3, 2.3.5, & 2.3.7 Attached w/ Revision: See Comments / Reference Proposed references to be provided during examination: None Learning Objective: ANALYZE the response to Instrument Air Compressor Trip or Header Pressure Low in accordance with ABN-301 Instrument Air System Malfunction.

Question Source: Bank ILOT8252 Modified Bank (Note changes or attach parent)

New Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Page 21 of 54 CPNPP 2014 NRC RO Written Exam Worksheet 11 to 20

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

ABN-301, Step 2.3.2 Revision: 12 Page 22 of 54 CPNPP 2014 NRC RO Written Exam Worksheet 11 to 20

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

ABN-301, Step 2.3.3 Revision: 12 Page 23 of 54 CPNPP 2014 NRC RO Written Exam Worksheet 11 to 20

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

ABN-301, Step 2.3.5 Revision: 12 Page 24 of 54 CPNPP 2014 NRC RO Written Exam Worksheet 11 to 20

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

ABN-301, Step 2.3.7 Revision: 12 Page 25 of 54 CPNPP 2014 NRC RO Written Exam Worksheet 11 to 20

ES-401 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. Date: 5/9/2014 Tier 1 Change: 3 Group 1 K/A W/E11 G 2.4.20 Level of Difficulty: 3 Importance Rating 3.8 Loss of Emergency Coolant Recirculation: Emergency Procedures/Plan: Knowledge of the operational implications of EOP warnings, cautions, and notes Proposed Question: 16 Given the following conditions:

Unit 2 has experienced a Loss of Coolant Accident.

During performance of EOP-1.0B, Loss of Reactor or Secondary Coolant, the crew determined that Residual Heat Removal Pump 2-01 was NOT available and that Bus 2EA2 was de-energized.

The Unit Supervisor transitioned to ECA-1.1B, Loss of Emergency Coolant Recirculation.

Prior to performing Step 3 to RESET Safety Injection, the Unit Supervisor read the following CAUTION:

If offsite power is lost after SI reset, manual action may be required to restore safeguards equipment to desired status.

RWST level is currently 43% and lowering.

Containment pressure is 12 psig and lowering.

While performing ECA-1.1B, Step 10, Determine Containment Spray Requirements (Suction From RWST), Offsite Power is lost to 2EA1 and the bus is re-energized by Diesel Generator 2-01.

The primary operational concern, in accordance with ECA-1.1B, Loss of Emergency Coolant Recirculation is that A. Centrifugal Charging Pump 2-01 must be manually re-started.

B. Containment Spray Pumps 2-01 and 2-03 must be manually re-started.

C. Safety Injection Pump 2-01 must be manually re-started.

D. Containment Isolation Phase A valves must be manually re-positioned.

Proposed Answer: C Page 26 of 54 CPNPP 2014 NRC RO Written Exam Worksheet 11 to 20

ES-401 RO Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible because a concern is always that cooling to the RCP seals is occurring. If the Centrifugal Charging Pump did not restart, manually restarting would be required. However, the CCP receives an auto start from the BOS.

B. Incorrect. Plausible because the Containment Spray Pumps would not restart after Step 6 is performed and Containment has not reached the pressure which would normally allow all Containment Spray flow to be stopped (3.0 psig). However, in ECA-1.1B, Step 10, all Containment Spray Pumps are stopped when Containment pressure is less than 18.0 psig to conserve RWST water. Thus the answer is incorrect in that the Containment Spray Pumps do not require restart.

C. Correct. The only ECCS Pumps available are Centrifugal Charging Pump 2-01 which would restart and Safety Injection Pump 2-01 which would need to be manually started.

D. Incorrect. Plausible because Containment Isolation Phase A would have been RESET in Step 5, but the valves should not reposition as a result of the Blackout Sequencer operation.

Technical Reference(s) EOS-1.3B, Step 5, CAUTION Attached w/ Revision: See EOS-1.3B, Attachment 3, Step 5 CAUTION Comments / Reference EOP-0.0B, Attachment 9, Item 10 ECA-1.1B, Step 3 CAUTION ECA-1.1B, Attachment 7, Step 3 CAUTION ECA-1.1B, Attachment 7, Step 9 Bases ECA-1.1B, Step 10 Proposed references to be provided during examination: None Learning Objective: Given a procedural Step, NOTE, or CAUTION, DISCUSS the reason or basis for the Step, NOTE, or CAUTION in ECA-1.1, Transfer to Cold Leg Recirculation.

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 27 of 54 CPNPP 2014 NRC RO Written Exam Worksheet 11 to 20

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

EOS-1.3B, Step 5, CAUTION Revision: 8 Page 28 of 54 CPNPP 2014 NRC RO Written Exam Worksheet 11 to 20

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

EOS-1.3B, Attachment 3, Step 5 CAUTION Revision: 8 Page 29 of 54 CPNPP 2014 NRC RO Written Exam Worksheet 11 to 20

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

EOP-0.0B, Attachment 9, Item 10 Revision: 8 Page 30 of 54 CPNPP 2014 NRC RO Written Exam Worksheet 11 to 20

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

ECA-1.1B, Step 3 CAUTION Revision: 8 Page 31 of 54 CPNPP 2014 NRC RO Written Exam Worksheet 11 to 20

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

ECA-1.1B, Attachment 7, Step 3 CAUTION Bases Revision: 8 Page 32 of 54 CPNPP 2014 NRC RO Written Exam Worksheet 11 to 20

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

ECA-1.1B, Attachment 7, Step 9 Bases Revision: 8 Page 33 of 54 CPNPP 2014 NRC RO Written Exam Worksheet 11 to 20

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

ECA-1.1B, Step 10 Revision: 8 Page 34 of 54 CPNPP 2014 NRC RO Written Exam Worksheet 11 to 20

ES-401 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. Date: 5/9/2014 Tier 1 Change: 4 Group 1 K/A W/E05 EK1.3 Level of Difficulty: 4 Importance Rating 3.9 Loss of Secondary Heat Sink: Knowledge of the operational implications of the following concepts as they apply to the Loss of Secondary Heat Sink: Annunciators and conditions indicating signals, and remedial actions associated with the Loss of Secondary Heat Sink Proposed Question: 17 Given the following conditions:

FRH-0.1A, Response to Loss of Secondary Heat Sink, is in progress on Unit 1.

All Reactor Coolant Pumps were stopped when Auxiliary Feedwater could NOT be established to at least one Steam Generator (SG).

Efforts are in progress to establish Main Feedwater to at least one SG.

Containment Pressure is 0 psig and stable.

In accordance with FRH-0.1A, Attachment 2, all SG Wide Range Levels are between 38% and 40%.

Pressurizer pressure has steadily risen over the last two minutes and both Pressurizer Power Operated Relief Valves have opened.

Reactor Coolant loop temperature differential between hot and cold legs is less than 5 ºF on all loops.

Which of the following indicates the actions required in accordance with FRH-0.1A?

A. Initiate Reactor Coolant System (RCS) bleed and feed.

B. Depressurize at least one SG to less than 500 psig.

C. Depressurize the RCS to less than 1910 psig.

D. Actuate SI and open Reactor Vessel and Pressurizer vents.

Proposed Answer: A Page 35 of 54 CPNPP 2014 NRC RO Written Exam Worksheet 11 to 20

ES-401 RO Written Exam Worksheet Form ES-401-5 Explanation:

A. Correct. In accordance with FRH-0.1A, an alternate indication of SG inventory reducing to critical levels where bleed and feed must be initiated is based on RCS pressure of 2335 psig. With both PORVs opening, an RCS pressure of at least 2335 is indicated.

B. Incorrect. Plausible because if Main Feedwater cannot be established, FRH-0.1A Step 9 requires that one SG be depressurized to less than 500 psig in order to feed the SG with condensate.

C. Incorrect. Plausible because if Main Feedwater cannot be established, FRH-0.1A Step 9 requires that the RCS be depressurized to 1910 in order to block Safety Injection and proceed with attempts to feed one SG with condensate.

D. Incorrect. Plausible because FRH-0.1A, Step 21 RNO has the operator open these vent valves if both PORVs cannot be opened. A misconception could exist that since the PORVs have opened additional pressure relief via the vent valves is required. This action is not correct in that once SI cooling had lowered RCS pressure, the PORVs would then close, thus limiting the adequate bleed path necessary to prevent core damage.

Technical Reference(s) FRH-0.1A, Steps 2, 3, 9, 11, 12 & 21 Attached w/ Revision: See Comments / Reference Proposed references to be provided during examination: None Learning Objective: Given a procedural Step, NOTE, or CAUTION, DISCUSS the reason or basis for the Step, NOTE, or CAUTION in FRH-0.1, Response to Loss of Secondary Heat Sink.

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 36 of 54 CPNPP 2014 NRC RO Written Exam Worksheet 11 to 20

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

FRH-0.1A, Steps 2 & 3 Revision: 8 Page 37 of 54 CPNPP 2014 NRC RO Written Exam Worksheet 11 to 20

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

FRH-0.1A, Step 9 Revision: 8 Page 38 of 54 CPNPP 2014 NRC RO Written Exam Worksheet 11 to 20

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

FRH-0.1A, Steps 11 & 12 Revision: 8 Page 39 of 54 CPNPP 2014 NRC RO Written Exam Worksheet 11 to 20

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

FRH-0.1A, Step 21 Revision: 8 Page 40 of 54 CPNPP 2014 NRC RO Written Exam Worksheet 11 to 20

ES-401 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. Date: 5/9/2014 Tier 1 Change: 4 Group 1 K/A 077 AK3.01 Level of Difficulty: 4 Importance Rating 3.9 Generator Voltage and Electric Grid Disturbances: Knowledge of the interrelations between Generator Voltage and Electric Grid Disturbances and the following: Reactor and turbine trip criteria Proposed Question: 18 Given the following conditions:

Unit 1 is at 45% Reactor Power.

Generator Load Target is 530 MWe.

North Texas is experiencing a winter ice storm.

Grid Frequency has just lowered from 58.7 Hz to 58.3 Hz.

Which of the following describes the REQUIRED action and the MAXIMUM allowed time, in accordance with ABN-601, Response to a 138/345 KV System Malfunction?

If grid frequency does NOT increase to greater than 58.4 Hz in A. 2 seconds, trip the turbine.

B. 30 seconds, trip the turbine.

C. 2 seconds, trip the reactor.

D. 30 seconds, trip the reactor.

Proposed Answer: D Page 41 of 54 CPNPP 2014 NRC RO Written Exam Worksheet 11 to 20

ES-401 RO Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible because if the grid frequency were to lower to less than or equal to 58.0 Hz then 2 seconds would be allowed prior to tripping the reactor. Tripping the turbine is plausible as ABN-402, Main Generator Malfunction has four sections which require a Turbine Trip rather than a Reactor Trip when reactor power is less than 50%.

B. Incorrect. Plausible because the action listed is the correct value and timing. Tripping the turbine is plausible as ABN-402, Main Generator Malfunction has four sections which require a Turbine Trip rather than a Reactor Trip when reactor power is less than 50%.

C. Incorrect. Plausible because if the grid frequency were to lower to less than or equal to 58.0 Hz then 2 seconds would be allowed prior to tripping the reactor. .

D. Correct. In accordance with ABN-601 Step 9.3, with grid frequency between 58.0 Hz and 58.4 Hz a delay of 30 seconds is allowed for frequency recovery prior to tripping the reactor.

Technical Reference(s) ABN-402, Section 2, 3, 6 & 7 Attached w/ Revision: See ABN-601, Section 9.3 Comments / Reference Proposed references to be provided during examination: None Learning Objective: COMPREHEND the normal, abnormal and emergency operation of Main Turbine and its support 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 Comprehension or Analysis 10 CFR Part 55 Content: 55.41 5 55.43 Page 42 of 54 CPNPP 2014 NRC RO Written Exam Worksheet 11 to 20

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

ABN-601, Step 9.3.3 Revision: 12 Page 43 of 54 CPNPP 2014 NRC RO Written Exam Worksheet 11 to 20

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

ABN-402, Step 2.3.5 RNO Revision: 8 Page 44 of 54 CPNPP 2014 NRC RO Written Exam Worksheet 11 to 20

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

ABN-402, Step 3.3.1 Revision: 8 Page 45 of 54 CPNPP 2014 NRC RO Written Exam Worksheet 11 to 20

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

ABN-402, Step 6.3.5 Revision: 8 Comments /

Reference:

ABN-402, Step 7.3.7 Revision: 8 Page 46 of 54 CPNPP 2014 NRC RO Written Exam Worksheet 11 to 20

ES-401 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. Date: 5/10/2014 Tier 1 Change: 3 Group 2 K/A 001 AA2.03 Level of Difficulty: 3 Importance Rating 4.5 Continuous Rod Withdrawal: Ability to determine and interpret the following as they apply to the Continuous Rod Withdrawal:

Proper actions to be taken if automatic safety functions have not taken place Proposed Question: 19 Given the following conditions:

Unit 1 is performing a Reactor Startup per IPO-002A, Plant Startup from Hot Standby twelve hours after a Reactor Trip.

The startup was delayed to troubleshoot a malfunction with the Logic Cabinet. I&C requested that 1/1-RBSS, CONTROL ROD BANK SELECT be placed in AUTO to confirm the repair is successful.

When 1/1-RBSS, CONTROL BANK SELECT is placed in AUTO, the operator identified a continuous rod withdrawal, after the Operator placed 1/1-RBSS in MAN the control rods stopped moving.

Current indications are as follows:

1-NI-31B, SR COUNT RATE CHAN I = 1.2 x 105 CPS and increasing.

1-NI-32B, SR COUNT RATE CHAN II = 8.4 x 104 CPS and increasing.

1-NI-35B, IR CURRENT CHAN I = 9 x 10-10 amps and increasing.

1-NI-36B, IR CURRENT CHAN II = 1.1 x 10-10 amps and increasing.

Which of the following actions are required following the inadvertent Continuous Rod Withdrawal?

A. Place 1/1-RTC, RX TRIP BKR in TRIP and perform the immediate actions of EOP-0.0A, Reactor Trip or Safety Injection.

B. Level Reactor Power at approximately 1 x 10-8 amps by adjusting the control rods as necessary to establish a 0 decade per minute startup rate.

C. Place 1/1-FLRM, CONTROL ROD MOTION CTRL in IN and drive rods to the desired control bank position for the 1/M plot data collection.

D. When 1-PCIP, Window 2.5 - SR RX TRIP BLK PERM P-6 is ON, place both SR RX TRIP RESET/BLK switches in BLOCK.

Proposed Answer: A Page 47 of 54 CPNPP 2014 NRC RO Written Exam Worksheet 11 to 20

ES-401 RO Written Exam Worksheet Form ES-401-5 Explanation:

A. Correct. The operator should recognize that the Reactor Trip logic of 1/2 Source Range channels greater than 105 CPS is satisfied and trip the Reactor.

B. Incorrect. Plausible because the startup concludes with the operator leveling Reactor Power at 1 x 10-8 amps. If the operator did not recognize by the indications that a Reactor Trip should have already occurred, this could be expected.

C. Incorrect. Plausible because the operator could believe that returning to the desired rod withdrawal endpoint is conservative as operator error introduced the initial problem.

D. Incorrect. Plausible because performing this action would be normal once by Intermediate Range Channels were greater than 1 x 10-10 amps and the P-6 permissive was received, therefore, if the operator did not recognize that the Source Range Reactor Trip should have already occurred, this could be expected.

Technical Reference(s) IPO-002A, Step 5.2.22 & 5.2.23 Attached w/ Revision: See EOP-0.0A, Section C.1 Comments / Reference ALM-0065A, 1-PCIP, Window 2.5 Proposed references to be provided during examination: None Learning Objective: ANALYZE the symptoms or entry conditions for EOP-0.0, Reactor Trip or Safety Injection.

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 6 55.43 Page 48 of 54 CPNPP 2014 NRC RO Written Exam Worksheet 11 to 20

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

IPO-002A, Step 5.2.22 & 5.2.23 Revision: 20 Page 49 of 54 CPNPP 2014 NRC RO Written Exam Worksheet 11 to 20

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

EOP-0.0A, Section C.1 Revision: 8 Page 50 of 54 CPNPP 2014 NRC RO Written Exam Worksheet 11 to 20

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

ALM-0065A, 1-PCIP, Window 2.5 Revision: 4 Page 51 of 54 CPNPP 2014 NRC RO Written Exam Worksheet 11 to 20

ES-401 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. Date: 5/10/2014 Tier 1 Change: 5 Group 2 K/A 003 AK1.03 Level of Difficulty: 2 Importance Rating 3.5 Dropped Control Rod: Knowledge of the operational implications of the following concepts as they apply to the Dropped Control Rod: Relationship of reactivity and reactor power to rod movement Proposed Question: 20 Given the following conditions:

Unit 1 is at 35% RTP with all systems in automatic.

Turbine Load is 350 MWe.

Control Bank D rods are at 172 steps.

Control Bank C rod K-6 drops to the bottom of the core.

NO Rod Control Urgent Failure alarms occur.

With NO operator action where will reactor thermal power and turbine load stabilize in response to the dropped rod?

Reactor Thermal Power Turbine Load A. 35% RTP 350 MWe B. 35% RTP 300 MWe C. 30% RTP 350 MWe D. 30% RTP 300 MWe Proposed Answer: A Page 52 of 54 CPNPP 2014 NRC RO Written Exam Worksheet 11 to 20

ES-401 RO Written Exam Worksheet Form ES-401-5 Explanation:

A. Correct. Power will initially decrease due to the dropped rod. As power decreases, temperature will decrease. As temperature decreases, positive reactivity is added to restore power. Bank D rods in auto will cause rods to step out and stop at C11 (223 steps). Rods stepping out and temperature lowering due to turbine control valves opening greater than the original position will restore power and turbine load to the original value.

B. Incorrect. Plausible since power will be restored due to the decrease in temperature and rod withdrawal. Control valve re-positioning will restore turbine load to 350 MWe not 300 MWe.

C. Incorrect. Plausible since power will initially decrease on the dropped rod, but power will be restored by the decreasing temperature and rods will step out. Control valve re-positioning will restore turbine load to 350 MWe.

D. Incorrect. Plausible since power will initially decrease on the dropped rod, but power will be restored by the decreasing temperature and rods will step out. Control valve re-positioning will restore turbine load to 350 MWe not 300 MWe.

Technical Reference(s) ABN-712, Section 3.2 Attached w/ Revision: See Comments / Reference Proposed references to be provided during examination: None Learning Objective: DISCUSS overall operation of the Rod Control and Digital Rod Position Indication systems.

COMPREHEND the normal, abnormal and emergency operation of the Rod 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 1 55.43 Page 53 of 54 CPNPP 2014 NRC RO Written Exam Worksheet 11 to 20

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

ABN-712, Section 3.2 Revision: 10 Page 54 of 54 CPNPP 2014 NRC RO Written Exam Worksheet 11 to 20

ES-401 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. Date: 5/10/2014 Tier 1 Change: 3 Group 2 K/A 024 AK3.02 Level of Difficulty: 3 Importance Rating 4.2 Emergency Boration: Knowledge of the reasons for the following responses as they apply to the emergency Boration: Actions contained in EOP for emergency boration Proposed Question: 21 Given the following conditions:

Unit 2 just tripped from 100% power.

The Unit was two weeks from a scheduled Refueling Outage.

Three Control Rods do NOT indicate fully inserted.

An Emergency Boration is commenced.

In accordance with the EOP-0.0B, Reactor Trip or Safety Injection Foldout Page, which of the following describes the MINIMUM required amount of boric acid to be injected and the reason for the boration?

A. 3600 gallons of boric acid. Accounts for the maximum reactivity worth of the additional two rods not assumed to stick out to ensure proper shutdown margin on the most limiting accident which is the Main Steam Line Break.

B. 3600 gallons of Boric Acid. Accounts for the maximum reactivity worth of the additional two rods not assumed to stick out to ensure proper shutdown margin on the most limiting accident which is the Large Break Loss of Coolant Accident.

C. 5400 gallons of Boric Acid. Accounts for the maximum reactivity worth of each rod stuck out to ensure proper shutdown margin on the most limiting accident which is the Main Steam Line Break.

D. 5400 gallons of Boric Acid. Accounts for the maximum reactivity worth of each rod stuck out to ensure proper shutdown margin on the most limiting accident which is the Large Break Loss of Coolant Accident.

Proposed Answer: C Page 1 of 51 CPNPP 2014 NRC RO Written Exam Worksheet 21 to 30

ES-401 RO Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible because shutdown margin is the concern, however, the required boration is 1800 gallons for each rod stuck out, not only for the two not assumed in the accident analysis.

The proper accident for the analysis is listed.

B. Incorrect. Plausible because shutdown margin is the concern, however, the required boration is 1800 gallons for each rod stuck out, not only for the two not assumed in the accident analysis. The Main Steam Line Break not the LBLOCA is the proper accident for the analysis listed.

C. Correct. A boration of 1800 gallons per rod is required to ensure shutdown margin in the event of a Main Steam Line Break per EOP-0.0A, Attachment 1.A.

D. Incorrect. Plausible because a boration of 1800 gallons per rod is required to ensure shutdown margin in the event of a Main Steam Line Break per EOP-0.0A, Attachment 1.A. The Main Steam Line Break not the LBLOCA is the proper accident for the analysis listed.

Technical Reference(s) EOP-0.0B, Attachment 1.A Attached w/ Revision: See Technical Specification LCO 3.1.1 Bases Comments / Reference Proposed references to be provided during examination: None Learning Objective: DISCUSS EOP-0.0, Reactor Trip or Safety Injection including the Purpose, Applicability, Symptoms/Entry Conditions, Operator Actions, Bases, Foldout Pages and Attachments.

Question Source: Bank ILOT8310 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 51 CPNPP 2014 NRC RO Written Exam Worksheet 21 to 30

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

EOP-0.0B, Attachment 1.A Revision: 8 Page 3 of 51 CPNPP 2014 NRC RO Written Exam Worksheet 21 to 30

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

Technical Specification LCO 3.1.1 Bases Revision: 68 Page 4 of 51 CPNPP 2014 NRC RO Written Exam Worksheet 21 to 30

ES-401 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. Date: 5/9/2014 Tier 1 Change: 3 Group 2 K/A 037 AA1.01 Level of Difficulty: 4 Importance Rating 3.7 Steam Generator Tube Leak: Ability to operate and/or monitor the following as they apply to the Steam Generator Tube Leak:

Maximum controlled depressurization rate for affected SG Proposed Question: 22 Given the following conditions:

Unit 2 has experienced a three gpm Steam Generator Tube Leak.

The Unit was shutdown to MODE 3 in accordance with ABN-106, High Secondary Activity.

EOP-0.0B, Reactor Trip or Safety Injection was completed through Step 4.

EOS-0.1B, Reactor Trip Response was completed with a transition to IPO-005B, Plant Cooldown from Hot Standby to Cold Shutdown.

The Reactor Operator is ready to commence a Reactor Coolant System cooldown in accordance with ABN-106.

Which of the following details the ABN-106, High Secondary Activity requirements for the Steam Generator (SG) depressurization prior to the Safety Injection Block at less than 1960 psig?

Dump steam to the Main Condenser A. at the maximum rate while ensuring Main Steam isolation is avoided.

B. at the maximum rate while controlling Pressurizer level 17% to 30%.

C. at less than or equal to 100ºF/hr cooldown rate while controlling Pressurizer level 17% to 30%.

D. at less than or equal to 100ºF/hr cooldown rate while ensuring Main Steam isolation is avoided.

Proposed Answer: C Page 5 of 51 CPNPP 2014 NRC RO Written Exam Worksheet 21 to 30

ES-401 RO Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible because in EOP-3.0B, Steam Generator Tube Rupture, the SG depressurization guidance is at maximum rate to avoid main steam isolation.

B. Incorrect. Plausible because maximum rate is specified in EOP-3.0B and the Pressurizer level band is used in ABN-106 until SI has been blocked. The lower band of 17% to 30% must be used until SI is blocked.

C. Correct. These are the correct values per ABN-106 until SI is blocked.

D. Incorrect. Plausible because the cooldown rate is correct but ensuring Main Steam Isolation is avoided is not considered as the plant cooldown is done in a controlled manner to maintain pressurizer level in a narrow control band and the Main Steamline rate Isolation signal is not challenged with a controlled cooldown.

Technical Reference(s) EOP-3.0B, Step 6.c Attached w/ Revision: See ABN-106, Steps 3.3.17 Comments / Reference ABN-106, Attachments 1 & 2 Proposed references to be provided during examination: None Learning Objective: ANALYZE the response to a Steam Generator Tube Leakage greater than or equal to 75 gpd in accordance with ABN-106, High Secondary Activity.

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 Page 6 of 51 CPNPP 2014 NRC RO Written Exam Worksheet 21 to 30

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

EOP-3.0B, Step 6.c Revision: 8 Page 7 of 51 CPNPP 2014 NRC RO Written Exam Worksheet 21 to 30

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

ABN-106, Step 3.3.17 Revision: 10 Page 8 of 51 CPNPP 2014 NRC RO Written Exam Worksheet 21 to 30

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

ABN-106, Attachment 1 Revision: 10 Page 9 of 51 CPNPP 2014 NRC RO Written Exam Worksheet 21 to 30

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

ABN-106, Attachment 2 Revision: 10 Page 10 of 51 CPNPP 2014 NRC RO Written Exam Worksheet 21 to 30

ES-401 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. Date: 5/14/2014 Tier 1 Change: 3 Group 2 K/A 069 AK2.03 Level of Difficulty: 4 Importance Rating 2.8 Loss of Containment Integrity: Knowledge of the interrelations between Loss of Containment Integrity and the following:

Personnel access hatch and emergency access hatch Proposed Question: 23 In accordance with OWI-801, Operations Department Local Leak Rate Testing, Acceptance Criteria, which of the following would require declaring the Containment inoperable in accordance with Technical Specification LCO 3.6.1, Containment while in MODE 1?

A. The outer door of the Personnel Air Lock is inoperable and closed and the inner door cannot be closed.

B. The Emergency Air Lock interlock mechanism is inoperable and neither door can be locked closed.

C. The Personnel Air Lock electrical interlock mechanism is inoperable and neither door can be locked closed.

D. Air leakage through the Emergency Air Lock has resulted in exceeding the overall Containment leakage rate.

Proposed Answer: D Explanation:

A. Incorrect. Plausible because this situation would make the Personnel Airlock inoperable but LCO 3.6.1 would not be entered as the overall Containment leakage rate may not be exceeded.

B. Incorrect. Plausible because this situation would make the Emergency Airlock inoperable but LCO 3.6.1 would not be entered as the overall Containment leakage rate may not be exceeded.

C. Incorrect. Plausible because this situation would make the Personnel Airlock inoperable but LCO 3.6.1 would not be entered as the overall Containment leakage rate may not be exceeded.

D. Correct. In accordance with LCO 3.6.2 Containment Air Locks NOTE 3, this condition requires declaring the Containment inoperable due to leakage Acceptance Criterias as described in OWI-801.

Technical Reference(s) Technical Specification LCO 3.6.1 & 3.6.2 Attached w/ Revision: See Technical Specification LCO 3.6.1 Bases Comments / Reference OWI-801 LO21SYSCY1 Study Guide Proposed references to be provided during examination: None Page 11 of 51 CPNPP 2014 NRC RO Written Exam Worksheet 21 to 30

ES-401 RO Written Exam Worksheet Form ES-401-5 Learning Objective: APPLY the administrative requirements of the Containment system including Technical Specifications, TRM and ODCM.

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 9 55.43 Page 12 of 51 CPNPP 2014 NRC RO Written Exam Worksheet 21 to 30

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

Technical Specification LCO 3.6.2 Amendment: 161 Page 13 of 51 CPNPP 2014 NRC RO Written Exam Worksheet 21 to 30

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

Technical Specification LCO 3.6.2 Amendment: 161 Page 14 of 51 CPNPP 2014 NRC RO Written Exam Worksheet 21 to 30

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

Technical Specification LCO 3.6.2 Amendment: 161 Page 15 of 51 CPNPP 2014 NRC RO Written Exam Worksheet 21 to 30

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

Technical Specification LCO 3.6.2 Amendment: 161 Page 16 of 51 CPNPP 2014 NRC RO Written Exam Worksheet 21 to 30

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

Technical Specification LCO 3.6.1 Bases Revision: 68 Page 17 of 51 CPNPP 2014 NRC RO Written Exam Worksheet 21 to 30

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

Technical Specification LCO 3.6.1 Amendment: 161 Page 18 of 51 CPNPP 2014 NRC RO Written Exam Worksheet 21 to 30

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

OWI-801 Revision: 6 Page 19 of 51 CPNPP 2014 NRC RO Written Exam Worksheet 21 to 30

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

LO21SYSCY1 Lesson Notes Revision: 5/2/2011 For some water tests, an air-driven hydro pump is used as the pressure source. The hydro pump is usually a self-contained rig that consists of the following:

Air-driven pump -- provides the pressure source for a hydrostatic test.

Air regulator -- determines the pressure output of the pump.

Discharge manifold -- allows installation of a calibrated discharge gauge (Hiese or Perma-cal type) and also allows controlled venting of the pressurized volume.

The higher of the two valve leakage values is considered to be the leakage from Containment through that particular penetration (most conservative assumption.) That value is added to the leakage value from all other penetrations and the total Containment Leakage sum, for Technical Specification compliance purposes, is derived.

Page 20 of 51 CPNPP 2014 NRC RO Written Exam Worksheet 21 to 30

ES-401 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. Date: 5/10/2014 Tier 1 Change: 3 Group 2 K/A 074 EA1.06 Level of Difficulty: 3 Importance Rating 3.6 Inadequate Core Cooling: Ability to operate and/or monitor the following as they apply to an Inadequate Core Cooling: RCPs Proposed Question: 24 Given the following conditions:

FRC-0.1A, Response to Inadequate Core Cooling, is in progress on Unit 1.

Efforts to depressurize the Steam Generators have been ineffective.

Which of the following is required in accordance with FRC-0.1A, Response to Inadequate Core Cooling?

Start A. only one RCP. If core cooling is insufficient PORVs are opened and the remaining RCPs are started sequentially.

B. no more than two RCPs to allow for spray flow and depressurize the RCS enough to permit RHR injection flow.

C. no more than three RCPs and reserve one RCP for future use, while the other three provide flow to the core.

D. RCPs one at a time until core exit TCs are less than 1200ºF in order to force two phase flow through the core for cooling.

Proposed Answer: D Page 21 of 51 CPNPP 2014 NRC RO Written Exam Worksheet 21 to 30

ES-401 RO Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible because in FRC-0.2A, Response to Degraded Core Cooling, Step 5, at least one RCP is verified running. At Step 7, the RCP in loop 4 is stopped to preserve it for future use.

Opening the PORVs is directed in FRC-0.1A at Step 20.b RNO once it is determined that running all RCPs has been ineffective at lowering core exit temperature below 1200°F.

B. Incorrect. Plausible because depressurizing the RCS to permit RHR injection flow is attempted earlier in FRC-0.1A at Step 13 in conjunction with the steam generator depressurization (see Step 13 bases). However, at 1200°F all RCPs are started.

C. Incorrect. Plausible because in FRC-0.2A, Response to Degraded Core Cooling RCP 4 is secured at Step 7 to preserve it for future depressurization spray flow. However, with core exit thermocouple temperatures greater than 1200°F all RCPs would be started.

D. Correct. As described in Step 20 of FRC-0.1A, starting the RCPs when core exit temperatures (CET) are greater than 1200°F will result in clearing of the water inventory in the RCS intermediate leg and permit circulation of hot gases from the overheated core into the steam generators. As stated in Steps 20.a through 20.d, idle RCPs are continuously started until CET temperatures are less than 1200°F.

Technical Reference(s) FRC-0.1A, Steps 13, 18, & 20 Attached w/ Revision: See FRC-0.1A, Attachment 5 Bases Comments / Reference FRC-0.2A, Steps 5 & 7 Proposed references to be provided during examination: None Learning Objective: Given a procedural Step, NOTE, or CAUTION, DISCUSS the reason or basis for the Step, NOTE, or CAUTION in FRC-0.1, Response to Inadequate Core Cooling.

Question Source: Bank ILOT0991 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 22 of 51 CPNPP 2014 NRC RO Written Exam Worksheet 21 to 30

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

FRC-0.1A, Step 13.c RNO Revision: 8 Page 23 of 51 CPNPP 2014 NRC RO Written Exam Worksheet 21 to 30

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

FRC-0.1A, Step 18 Revision: 8 Comments /

Reference:

FRC-0.1A, Step 20 Revision: 8 Page 24 of 51 CPNPP 2014 NRC RO Written Exam Worksheet 21 to 30

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

FRC-0.1A, Step 20 Revision: 8 Page 25 of 51 CPNPP 2014 NRC RO Written Exam Worksheet 21 to 30

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

FRC-0.1A, Attachment 5, Step 13 Bases Revision: 8 Page 26 of 51 CPNPP 2014 NRC RO Written Exam Worksheet 21 to 30

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

FRC-0.1A, Attachment 5, Step 20 Bases Revision: 8 Page 27 of 51 CPNPP 2014 NRC RO Written Exam Worksheet 21 to 30

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

FRC-0.2A, Steps 5 & 7 Revision: 8 Page 28 of 51 CPNPP 2014 NRC RO Written Exam Worksheet 21 to 30

ES-401 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. Date: 5/10/2014 Tier 1 Change: 4 Group 2 K/A 076 AK3.06 Level of Difficulty: 4 Importance Rating 3.2 High Reactor Coolant Activity: Knowledge of the reasons for the following responses as they apply to the High Reactor Coolant Activity: Actions contained in EOP for high reactor coolant activity Proposed Question: 25 Given the following conditions:

Unit 2 is at 800 MWe, 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> after a Heater Drain Pump 2-01 trip.

Chemistry reports that Reactor Coolant System (RCS) Cs-137 levels are elevated over the last shift.

Letdown flow is 75 gpm.

Which of the following indicates the actions required in accordance with ABN-102, High Reactor Coolant Activity and reason for those actions?

A. Lower letdown flow to 45 gpm to minimize dose rates in Auxiliary Building.

Contact Core Performance Engineering to determine extent of fuel failure.

B. Lower letdown flow to 45 gpm to minimize dose rates in Auxiliary Building.

Contact Chemistry to determine if a CRUD burst has occurred.

C. Raise letdown to 120-140 gpm to increase ion exchange.

Contact Core Performance Engineering to determine extent of fuel failure.

D. Raise letdown to 120-140 gpm to increase ion exchange.

Contact Chemistry to determine if a CRUD burst has occurred.

Proposed Answer: C Page 29 of 51 CPNPP 2014 NRC RO Written Exam Worksheet 21 to 30

ES-401 RO Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible because it could be thought that minimizing letdown would lower dose rates, however, raising letdown is required to remove more radioactive ions and particulates. Cs-137 is indicative of fuel failure.

B. Incorrect. Plausible because it could be thought that minimizing letdown would lower dose rates, however, raising letdown is required to remove more radioactive ions and particulates. Cs-137 is indicative of fuel failure so checking for a CRUD burst is not the correct action.

C. Correct. Raising letdown flow is the correct action to remove more radioactive ions and particulates and Core Performance Engineering should be contacted to determine the extent of fuel damage as Cs-137 is indicative of fuel failure.

D. Incorrect. Raising letdown flow is the correct action to remove more radioactive ions and particulates. Cs-137 is indicative of fuel failure so checking for a CRUD burst is not the correct action.

Technical Reference(s) ABN-102, Steps 2.3.2, 2.3.3, 2.3.6 & 2.3.7 Attached w/ Revision: See ABN-102, Step 2.3.7 NOTE, 2.3.8, & 2.3.9 Comments / Reference Proposed references to be provided during examination: None Learning Objective: ANALYZE the response to High Reactor Coolant Activity in accordance with ABN-102, High Reactor Coolant Activity.

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 55.43 Page 30 of 51 CPNPP 2014 NRC RO Written Exam Worksheet 21 to 30

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

ABN-102, Step 2.3.2, 2.3.3, & 2.3.6 Revision: 7 Page 31 of 51 CPNPP 2014 NRC RO Written Exam Worksheet 21 to 30

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

ABN-102, Step 2.3.7 NOTE, 2.3.8, & 2.3.9 Revision: 7 Page 32 of 51 CPNPP 2014 NRC RO Written Exam Worksheet 21 to 30

ES-401 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. Date: 5/10/2014 Tier 1 Change: 3 Group 2 K/A W/E02 EA2.2 Level of Difficulty: 2 Importance Rating 3.5 Safety Injection Termination: Ability to determine and interpret the following as they apply to the SI Termination: Adherence to appropriate procedures and operation within the limitations in the facilitys license and amendments Proposed Question: 26 Given the following conditions:

Following a Reactor Trip and Safety Injection, the crew has transitioned to EOS-1.1A, Safety Injection Termination.

Centrifugal Charging Pump 1-02 and both Safety Injection Pumps have been stopped and placed in standby.

Normal Charging flow has been established.

Containment pressure is 1.2 psig and stable.

Reactor Coolant System subcooling is currently 19ºF and slowly degrading.

Pressurizer level is 18% and slowly decreasing.

Which of the following actions is to be taken in accordance with the Foldout Page of EOS-1.1A, Safety Injection Termination?

Manually A. control Charging flow as necessary and continue in EOS-1.1A, Safety Injection Termination.

B. operate Emergency Core Cooling Pumps as necessary and continue in EOS-1.1A, Safety Injection Termination.

C. operate Emergency Core Cooling Pumps as necessary and transition to EOP-1.0A, Loss of Reactor or Secondary Coolant.

D. control Charging flow as necessary and transition to EOP-1.0A, Loss of Reactor or Secondary Coolant.

Proposed Answer: C Page 33 of 51 CPNPP 2014 NRC RO Written Exam Worksheet 21 to 30

ES-401 RO Written Exam Worksheet Form ES-401-5 Explanation:

A. Plausible since this is the action required if pressurizer level is below 6%, but with subcooling below the required value ECCS Pumps must be started and a transition made to EOP-1.0.

B. Plausible since ECCS Pumps are started as necessary, but a transition to EOP-1.0 is also required.

C. Based on subcooling being less than 25°F, ECCS Pumps must be started as necessary and a transition made to EOP-1.0.

D. Plausible since a transition to EOP-1.0 is required, but starting ECCS Pumps is the action to be taken instead of controlling charging flow.

Technical Reference(s) EOS-1.1A, Attachment 1.A Attached w/ Revision: See Comments / Reference Proposed references to be provided during examination: None Learning Objective: IDENTIFY the items on EOS-1.1, Safety Injection Termination Foldout Page including any equipment, parameter, set point or condition.

Question Source: Bank ILOT5805 Modified Bank (Note changes or attach parent)

New Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Page 34 of 51 CPNPP 2014 NRC RO Written Exam Worksheet 21 to 30

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

EOS-1.1A, Attachment 1.A Revision: 8 Page 35 of 51 CPNPP 2014 NRC RO Written Exam Worksheet 21 to 30

ES-401 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. Date: 3/4/2014 Tier 1 Change: 3 Group 2 K/A W/E13 G 2.4.35 Level of Difficulty: 3 Importance Rating 3.8 Steam Generator Overpressure: Emergency Procedures/Plan: Knowledge of local auxiliary operator tasks during emergency and the resultant operational effects Proposed Question: 27 Given the following conditions:

Unit 1 has experienced a Reactor Trip from a spurious Main Steam Line Isolation signal.

The crew is attempting to control Reactor Coolant System (RCS) temperature in EOS-0.1A, Reactor Trip Response.

Steam Generator (SG) 1-04 pressure is 1250 psig and stable.

SGs 1-01, 1-02, and 1-03 are 1092 psig and stable.

Field Support has reported that 1-PV-2328, SG4 ATM RLF VLV is mechanically bound and will NOT open.

The Unit Supervisor directs that FRH-0.2A, Response to Steam Generator Overpressure be performed.

Which of the following Nuclear Equipment Operator field actions are specified in FRH-0.2A, Response to Steam Generator Overpressure to reduce SG 1-04 pressure?

Open A. 1-HV-2336B, MSIV 1-04 BYP VLV.

B. 1-HV-2452-1, AFWPT STM SPLY VLV MSL 4.

C. 1-HV-2412, MSL 4 BEF MSIV D\POT ISOL VLV.

D. 1MS-0712, MSL 1-04 TO AFWPT STM SPLY VLV BYP VLV.

Proposed Answer: A Page 36 of 51 CPNPP 2014 NRC RO Written Exam Worksheet 21 to 30

ES-401 RO Written Exam Worksheet Form ES-401-5 Explanation:

A. Correct. Local actions to open this valve are included in FRH-0.2A, Steps 4 and 8.

B. Incorrect. Plausible because the opening of this valve is included in FRH-0.2A, Steps 4 and 8, but would be performed by the Reactor Operator in lieu of the NEO.

C. Incorrect. Plausible because the opening of this valve is included in FRH-0.2A, Steps 4 and 8, but would be performed by the Reactor Operator in lieu of the NEO.

D. Incorrect. Plausible because the opening of this valve would be effective in reducing the SG pressure, but this valve is not specified to be opened in FRH-0.2A, Steps 4 and 8. The Bases for Step 8 does allow the use of any plant specific means which could result in the utilization of this valve for pressure reduction.

Technical Reference(s) FRH-0.2A, Steps 4 & 8 Attached w/ Revision: See FRH-0.2A, Attachment 2, Step 8 Bases Comments / Reference Proposed references to be provided during examination: None Learning Objective: Given a procedural Step, NOTE, or CAUTION, DISCUSS the reason or basis for the Step, NOTE, or CAUTION in FRH-0.2, Response to Degraded Core Cooling.

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

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 4 55.43 Page 37 of 51 CPNPP 2014 NRC RO Written Exam Worksheet 21 to 30

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

FRH-0.2A, Step 4 Revision: 8 Page 38 of 51 CPNPP 2014 NRC RO Written Exam Worksheet 21 to 30

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

FRH-0.2A, Step 8 Revision: 8 Page 39 of 51 CPNPP 2014 NRC RO Written Exam Worksheet 21 to 30

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

FRH-0.2A, Attachment 2, Step 8 Bases Revision: 8 Page 40 of 51 CPNPP 2014 NRC RO Written Exam Worksheet 21 to 30

ES-401 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. Date: 3/6/2014 Tier 2 Change: 3 Group 1 K/A 003 K5.04 Level of Difficulty: 3 Importance Rating 3.2 Reactor Coolant Pump System: Knowledge of the operational implications of the following concepts as they apply to the RCPs: Effects of RCP shutdown on secondary parameters, such as steam pressure, steam flow, and feed flow Proposed Question: 28 Given the following conditions:

Unit 1 is steady at 30% power when Reactor Coolant Pump 1-02 trips and causes a transient in Steam Generator 1-02.

Which of the following describes how steam flow and water level in Steam Generator 1-02 initially respond to the trip of Reactor Coolant Pump 1-02 prior to manually tripping the reactor?

Initially, Steam Generator 1-02 steam flow __________ and level __________.

A. increases increases B. increases decreases C. decreases decreases D. decreases increases Proposed Answer: C Explanation:

A. Incorrect. Plausible if thought that pressure dropped in Steam Generator #2 which caused a swell and level increase.

B. Incorrect. Plausible because Steam Generator level will decrease, however, steam flow will also decrease as the Steam Generator cools and Steam Generator pressure decreases.

C. Correct. Because the Steam Generator with the tripped Reactor Coolant Pump stops steaming, steam flow will decrease and Steam Generator level will also decrease due to shrink.

D. Incorrect. Plausible because steam flow will decrease, however, Steam Generator level will also decrease due to shrink.

Technical Reference(s) ABN-101, Step 2.3.1 NOTE Attached w/ Revision: See Comments / Reference Page 41 of 51 CPNPP 2014 NRC RO Written Exam Worksheet 21 to 30

ES-401 RO Written Exam Worksheet Form ES-401-5 Proposed references to be provided during examination: None Learning Objective: COMPREHEND the normal, abnormal and emergency operation of the Reactor Coolant Pump System.

Question Source: Bank ILOT5789 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:

ABN-101, Step 2.3.1 NOTE Revision: 10 Page 42 of 51 CPNPP 2014 NRC RO Written Exam Worksheet 21 to 30

ES-401 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. Date: 5/14/2014 Tier 2 Change: 3 Group 1 K/A 026 K4.09 Level of Difficulty: 3 Importance Rating 3.2 Containment Spray System: Knowledge of the CSS design feature(s) and/or interlock(s) that provide for the following:

Prevention of path for escape of radioactivity from containment to the outside (interlock on RWST isolation after swapover).

Proposed Question: 29 Which of the following describes the direct interlock which automatically closes the Containment Spray Pump (CSP) 1-01 Recirculation Valve (1-FV-4772-1) with 1-HS-4772-1, CSP 1 RECIRC VLV in the AUTO position?

A. Upon receipt of a Containment Spray Actuation signal.

B. When CSP 1-01 discharge flow increases to greater than 800 gpm.

C. Upon receipt of a Safety Injection Actuation signal.

D. Opening Containment sump to CSP 1 & 3 Suction 1-HS-4782.

Proposed Answer: D Page 43 of 51 CPNPP 2014 NRC RO Written Exam Worksheet 21 to 30

ES-401 RO Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible if thought that a Containment Spray Actuation signal will isolate all recirc pathways to maximize flow to containment upon spray actuation.

B. Incorrect. Plausible because with 1-HS-4772-1 in AUTO the valve will open at 1090 gpm and close at 1210 gpm pump discharge flow unless the CNTMT sump suction or the heat exchanger outlet valves are open.

C. Incorrect. Plausible if thought that a Safety Injection Actuation signal will isolate all recirc pathways to maximize flow to containment should a spray action be imminent.

D. Correct. When the containment sump to CSP 1 & 3 suction isolation valve (1-HV-4782) is opened 1-FV-4772-1 will automatically close, when 1-HS-4772-1 is in the AUTO position to prevent radioactive contamination of the RWST.

Technical Reference(s) LO21.SYS.CT1, Page 12 Attached w/ Revision: See Comments / Reference Proposed references to be provided during examination: None Learning Objective: COMPREHEND the normal, abnormal and emergency operation of the Chemical and Volume Control system.

Question Source: Bank ILOT6515 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 Page 44 of 51 CPNPP 2014 NRC RO Written Exam Worksheet 21 to 30

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

LO21.SYS.CT1, Page 12 Revision: 05/02/11 To monitor for proper operation of the containment spray pumps, discharge pressure and flow indications are provided on the MCB (CB-02). Installed temperature elements provide operating temperatures which can be read on the plant computer. Temperature indications available include pump and motor bearing temperatures, motor stator winding temperatures and pump bearing cooling water temperatures. The containment spray pumps do not receive a start signal from the Blackout sequencer.

CONTAINMENT SPRAY PUMP RECIRC VALVES (U-HV-4772-1, 2-2, 3-1 AND 3-2 FIGURE 9)

After receipt of a Safety Injection signal, but prior to Containment Spray actuation ( Hi-3 setpoint), the Containment Spray pumps are running with the heat exchanger outlet valve closed. Each pump is provided with a recirculation valve which ensures a minimum flow through the pump to prevent pump damage. This recirculation line taps off of the pump discharge between the pump and the manual discharge isolation valve. A check valve downstream of the recirculation valve prevents backflow through the recirculation line. Each pump recirculation line ties into the common Containment Spray test line which returns flow back to the RWST.

The recirculation valves are motor operated and are supplied from their train related 480 vac bus. On loss of power the valves fail as is and have a handwheel for manual operation. The valves are located in the train related ECCS valve room safeguards building 790 elevation. Each valve is controlled from the main control board (CB-02) via a three position (close-auto-open) spring return to center hand switch. Close and open valve position indication lights are included on the hand switch. Other valve position indications include, a not closed indication on the plant computer and valve closed indication on a train related MLB (MLB-4A3/4B3).

In automatic, the valve is controlled by a flow bistable off of the pump discharge flow transmitter. At approximately 1090 gpm, the flow bistable trips, sending an open signal to the valve. The bistable resets at 120 gpm greater than the trip setpoint (1210 gpm). If the recirculation valve is not full open with a low flow signal present a discharge low flow alarm will come in on the MCB (ALB-2B).

The recirculation valve is interlocked with the heat exchanger outlet valve to ensure full flow is available from the pump. The recirculation valve goes closed once the outlet valve begins to open.

The recirculation valve is also interlocked with the recirculation sump suction valve, receiving a close signal once the sump suction valve begins to open. A recirculation valve fail to close alarm will come in if the valve is not fully closed with its associated pump running and the recirculation sump suction valve off of its closed seat. This ensures that a flow path does not exist from the sump to the RWST which would contaminate the RWST and reduce the sump inventory.

The recirculation flowpath is utilized to add pump heat to the RWST during cold weather when necessary to maintain tank temperature above the minimum temperature. If power is lost to the valve due to an overload a Motor Control Center motor operated valve overload alarm is received on the MCB (ALB-11B).

Page 45 of 51 CPNPP 2014 NRC RO Written Exam Worksheet 21 to 30

ES-401 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. Date: 3/4/2014 Tier 2 Change: 2 Group 1 K/A 005 K1.11 Level of Difficulty: 3 Importance Rating 3.5 Residual Heat Removal System: Knowledge of the physical connections and/or cause-effect relationships between the RHRS and the following systems: RWST Proposed Question: 30 Given the following conditions:

Unit 1 has experienced a Large Break Loss of Coolant Accident.

EOP-1.0A, Loss of Reactor or Secondary Coolant is in progress.

The following signals have been RESET:

Safety Injection Phase A Isolation Phase B Isolation Before any other signals are reset, the Refueling Water Storage Tank (RWST) level decreases to 33%.

Which of the following describes the response of the Residual Heat Removal (RHR) Pump Suction Valves?

RHR Containment RWST to RHR Sump Suction Valves Suction Valves 1/1-8811A & 1/1-8811B 1/1-8812A & 1/1-8812B A. Automatically open Automatically close B. Automatically open Must be manually closed C. Must be manually opened Must be manually closed D. Must be manually opened Automatically close Proposed Answer: B Page 46 of 51 CPNPP 2014 NRC RO Written Exam Worksheet 21 to 30

ES-401 RO Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible since the sump suction valves receive an automatic open signal, but the RWST suction valves must be manually closed.

B. Correct. The RHR suction valves receive an auto switchover signal. Auto switchover for RHR will occur on low-low RWST level coincident with SI. Auto switchover for RHR also requires that the operator complete the alignment by closing the RWST to RHR Suction Valves. This SI signal is reset independently, so it will occur even after SI is reset.

C. Incorrect. Plausible since the RWST suction valves must be manually closed, but the sump suction valves will automatically open even with the SI reset since a separate signal is used to input this coincidence.

D. Incorrect. Plausible since one set of valves receive an automatic signal and the other set of valves must be manually operated, but the sump suction valves automatically open and the RWST suction valves must be manually closed.

Technical Reference(s) EOS-1.3A, Step 3 Attached w/ Revision: See LO21.SYS.RH1, Pages 15 & 16 Comments / Reference Proposed references to be provided during examination: None Learning Objective: COMPREHEND the normal, abnormal and emergency operation of the Residual Heat Removal System.

Question Source: Bank ILOT6395 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 Page 47 of 51 CPNPP 2014 NRC RO Written Exam Worksheet 21 to 30

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

EOS-1.3A, Step 3 Revision: 8 Page 48 of 51 CPNPP 2014 NRC RO Written Exam Worksheet 21 to 30

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

EOS-1.3A, Step 3 Revision: 8 Page 49 of 51 CPNPP 2014 NRC RO Written Exam Worksheet 21 to 30

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

LO21.SYS.RH1, Page 16 Revision: 10/20/11 CONTAINMENT SUMP TO RHR PUMP SUCTION ISOLATION VALVES (U-8811A&B)

The Containment Sump to RHR Pump Suction Isolation Valves allow recovery of borated water that has spilled from the Reactor Coolant System onto the Containment floor during LOCA conditions. These valves (along with the Containment Sump to Containment Spray Pump Suction Isolation Valves) also allow recovery of water sprayed from the Containment Spray System into the Containment Building.

The Containment Sump to RHR Pump Suction Isolation Valves are normally closed and open automatically when the RWST reaches its LO-LO alarm at 33% level. The valves are located in independent train related lines to ensure a single failure does not cause a complete loss of Emergency Core Cooling System flow.

Bonnet Pressure Relief Valves (u-SI-0182 and u SI-0183) are provided on u-8811A and u-8811B respectively to preclude pressure locking of the Containment Sump to RHR Pump Suction Isolation Valves. Water trapped in the bonnet can thermally expand causing pressure to increase between the discs. Excessive pressure may preclude the motor from opening the valves. A relief is provided to prevent this occurrence. The relief valve lifts at 475 psig and relieves to the downstream side of the valve. The relief valve is classified as an active containment isolation valve.

The Containment Sump to RHR Pump Suction Isolation Valves are motor-operated valves which are operated from CB-04. Valve u-8811A receives power from uEB3-2 and valve u-8811B receives power from uEB4-2.

Opening the Train A valve using the Main Control Board switch requires the following interlocks be met:

RWST to RHR Pump Suction Valve (u-8812A) must be CLOSED, and One of the RHR Pump Hot Leg Recirculation Isolation Valve (u-8701A or u-8702A) must be CLOSED, and The Main Control Board switch in the OPEN position The interlocks for the Train B valve utilize their train B counterpart.

These interlocks aid in preventing cross-connecting the RWST and the Containment Sumps. Following a Safety Injection Actuation signal, these valves will automatically open once the RWST reaches its LO-LO alarm.

Page 50 of 51 CPNPP 2014 NRC RO Written Exam Worksheet 21 to 30

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

LO21.SYS.RH1, Pages 15 & 16 Revision: 10/20/11 REFUELING WATER STORAGE TANK TO RHR PUMP SUCTION VALVES (U-8812A&B)

The Refueling Water Storage Tank to RHR Pump Suction Valves allow the RHR System to transfer borated water from the RWST to the Reactor Coolant System. During normal at power operations, these valves are open as part of the RHR System standby status. These valves are closed when the RHR System is placed in the shutdown cooling mode. Once the unit had been shutdown and the reactor vessel head prepped for removal, these valves are reopened to allow filling of the Refueling Cavity.

The Refueling Water Storage Tank to RHR Pump Suction Valves are motor-operated valves which are operated from CB-04. Valve u-8812A receives power from uEB3-1 and valve u-8812B receives power from uEB4-1.

Opening the Train A valve using the Main Control Board switch requires the following interlocks be met:

Containment Sump to RHR Pump Suction Isolation Valve (u-8811A) must be CLOSED, and The Main Control Board switch in the OPEN position The interlocks for the Train B valve utilize its Train B counterpart Page 51 of 51 CPNPP 2014 NRC RO Written Exam Worksheet 21 to 30

ES-401 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. Date: 5/13/14 Tier 2 Change: 3 Group 1 K/A 006 A2.13 Level of Difficulty: 2 Importance Rating 3.9 Emergency Core Cooling System: Ability to (a) predict the impacts of the following malfunctions or operations on the ECCS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Inadvertent SIS actuation Proposed Question: 31 Given the following condition:

An inadvertent Safety Injection Actuation has occurred on Unit 1.

Which of the following is an adverse effect of allowing Safety Injection to continue while performing EOS-1.1A, Safety Injection Termination mitigates?

A. Emergency Core Cooling System Pump motor heating due to running for extended time periods at minimum flow.

B. Loss of Instrument Air to Containment will not allow the use of the normal Pressurizer Spray Valves to control Pressurizer Pressure.

C. Centrifugal Charging Pumps running in the Injection Mode will collapse the Pressurizer bubble and risk damaging the Pressurizer Safeties.

D. Reactor Coolant Pumps will be running without adequate pump seal cooling due to seal water return containment isolation.

Proposed Answer: C Explanation:

A. Incorrect. Plausible because it could be thought that ECCS Pumps running at low flows would be the most significant affect.

B. Incorrect. Plausible because it could be thought that using Pressurizer Spray Valves could prevent an overpressure condition, however, the Pressurizer would continue to fill and pressurize the RCS until inventory was controlled.

C. Correct. The high head Centrifugal Charging Pumps will continue to increase inventory resulting in high pressures up to the PORV setpoint if steps to reduce flow and restore Letdown as part of SI Termination are not performed.

D. Incorrect. Plausible because the RCPs would be running with Seal Injection but Seal Return flow would be via the Seal Water Return Relief Valve to the PRT, which could lead to the conclusion that the relief to the PRT would provide inadequate seal cooling.

Page 1 of 37 CPNPP 2014 NRC RO Written Exam Worksheet 31 to 40

ES-401 RO Written Exam Worksheet Form ES-401-5 Technical Reference(s) OPGD 3, Attachment 6 Attached w/ Revision: See Comments / Reference Proposed references to be provided during examination: None Learning Objective: Given a procedural step, or sequence of steps from EOS-1.1, Safety Injection Termination, STATE the purpose/basis for the step(s).

Question Source: Bank ILOT8273 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:

OPGD 3, Attachment 6 Revision: 06/13/13 1.1. Procedure Expediency The Unit Supervisors pace through ERGs / ABNs is always critical until the Reactor has been placed in a stable condition.

The Unit Supervisor should consider temporarily suspending expectations that may otherwise be in effect during ERG/ABN performance (e.g., Crew Briefings). any time delay in working through the procedures may contribute to the event severity. The following are examples of conditions associated with ERG/ABN performance.

Inadvertent SI: To preclude overfilling the pressurizer with the increased risk of damaging the pressurizer safeties, the ERG network should be worked expeditiously to the point of re-establishing letdown (terminating the pressurizer fill.)

Page 2 of 37 CPNPP 2014 NRC RO Written Exam Worksheet 31 to 40

ES-401 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. Date: 5/13/2014 Tier 2 Change: 3 Group 1 K/A 007 A4.09 Level of Difficulty: 3 Importance Rating 2.5 Pressurizer Relief/Quench Tank System: Ability to manually operate and/or monitor in the control room: Relationship between PZR level and changing levels of the PRT and bleed holdup tank Proposed Question: 32 Given the following conditions:

Unit 1 is at 100% power.

Reactor Coolant System Letdown flow is 132 gpm.

DC power is lost to the solenoid valve for 1-8160, Letdown Containment Isolation Valve.

Which of the following describes the effect on Pressurizer level and Pressurizer Relief Tank (PRT) level?

Pressurizer level will A. decrease and PRT level will remain the same.

B. remain on program and PRT level will increase.

C. remain on program and PRT level will remain the same.

D. decrease and PRT level will increase.

Proposed Answer: B Page 3 of 37 CPNPP 2014 NRC RO Written Exam Worksheet 31 to 40

ES-401 RO Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible because the VCT level will decrease with no Letdown Return Flow. Pressurizer level will remain the same (on program), but a misconception could be that RCS fluid is being lost and Pressurizer level must be adversely affected. PRT level will increase, but could be easily missed if the operator forgets that the letdown relief valve will relieve the entire letdown flow to the PRT.

B. Correct. Closing 1-8160 will cause the letdown relief valve 1-8117 to lift to the PRT causing level to increase. The capacity of the valve is enough to pass all letdown flow; therefore, pressurizer will remain on program. With letdown isolated, VCT level will decrease.

C. Incorrect. Plausible because the Pressurizer will remain on program and VCT level will decrease.

PRT level will increase, but could be easily missed if the operator forgets that the letdown relief valve will relieve the entire letdown flow to the PRT.

D. Incorrect. Plausible because the PRT level will increase. VCT level will decrease but with auto makeup it could be conceived that VCT level will remain the same, however, auto makeup cannot makeup at 132 gpm and VCT level will decrease. Pressurizer level will remain the same (on program), but a misconception could be that RCS fluid is being lost and Pressurizer level must be adversely affected.

Technical Reference(s) LO21.SYS.CS1, Pages 14 & 73 Attached w/ Revision: See ALM-0061A, 1-ALB-6A, Window 4.3 Comments / Reference Proposed references to be provided during examination: None Learning Objective: DESCRIBE the components of the Chemical and Volume Control system including interrelations with other systems to include interlocks and control loops.

Question Source: Bank ILOT8080 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 Page 4 of 37 CPNPP 2014 NRC RO Written Exam Worksheet 31 to 40

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

LO21.SYS.CS1 Page 14 Revision: 04/28/2011 Comments /

Reference:

LO21.SYS.CS1, Page 73 Revision: 04/28/2011 Comments /

Reference:

ALM-0061A, 1-ALB-6A, Window 4.3 Revision: 7 Page 5 of 37 CPNPP 2014 NRC RO Written Exam Worksheet 31 to 40

ES-401 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. Date: 5/13/2014 Tier 2 Change: 4 Group 1 K/A 008 K2.02 Level of Difficulty: 2 Importance Rating 3.0 Component Cooling Water System: Knowledge of bus power supplies to the following: CCW pump, including emergency backup Proposed Question: 33 Given the following conditions:

Unit 2 is in MODE 1.

An XST1 Transformer fault has just occurred.

All systems responded in accordance with design.

Given the above conditions, which of the following is correct for Component Cooling Water Pump (CCWP) 2-02?

CCWP 2-02 is powered from A. 2EA1 which is supplied by XST2.

B. 2EA1 which is supplied by Emergency Diesel Generator 2-01.

C. 2EA2 which is supplied by XST2.

D. 2EA2 which is supplied by the Emergency Diesel Generator 2-02.

Proposed Answer: C Explanation:

A. Incorrect. Plausible if thought that CCWP 2-02 was powered from Train A as 2EA1 would be powered from XST2.

B. Incorrect. Plausible if thought that CCWP 2-02 was powered from Train A and that 2EA1 would be supplied by the EDG following the transformer fault.

C. Correct. The Train B CCWP would be powered from 2EA2, which would be supplied by the alternate offsite source following a fault of the preferred power supply. Following the slow transfer to XST1, the bus would automatically load the CCWP 2-02 on the bus. The EDG would not start and supply the bus if the slow transfer was successful.

D. Incorrect. Plausible as the proper power supply is listed, however, following the slow transfer to XST1 the bus would automatically load the CCWP 2-02 on the bus. The EDG would not start and supply the bus if the slow transfer was successful.

Technical Reference(s) LO21.SYS.CC1, Page 14 Attached w/ Revision: See Page 6 of 37 CPNPP 2014 NRC RO Written Exam Worksheet 31 to 40

ES-401 RO Written Exam Worksheet Form ES-401-5 LO21.SYS.AC2, Page 12 Comments / Reference Proposed references to be provided during examination: None Learning Objective: COMPREHEND the normal, abnormal and emergency operation of the 6.9 KV and 480 V Electrical Distribution 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 X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 Comments /

Reference:

LO21.SYS.CC1, Page 14 Revision: 05/01/11 Page 7 of 37 CPNPP 2014 NRC RO Written Exam Worksheet 31 to 40

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

LO21.SYS.AC2, Page 12 Revision: 04/28/11 Page 8 of 37 CPNPP 2014 NRC RO Written Exam Worksheet 31 to 40

ES-401 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. Date: 5/14/2014 Tier 2 Change: 3 Group 1 K/A 010 K6.01 Level of Difficulty: 3 Importance Rating 2.7 Pressurizer Pressure Control System: Knowledge of the effect of a loss or malfunction of the following will have on the PZR PCS: Pressure Detection Systems Proposed Question: 34 Given the following conditions:

Unit 1 is proceeding to MODE 5 in accordance with IPO-005A, Plant Cooldown from Hot Standby to Cold Shutdown.

Pressurizer Pressure is 350 psig and stable.

RCS Temperature is 340°F and stable.

1-PT-403, HL PRESS (WR) fails high.

Which of the following would be the response of the Pressurizer Pressure Control System, specifically Power Operated Relief Valve (PORV) positions, following the instrument failure?

A. 1-PCV-455A OPEN, 1-PCV-456 OPEN B. 1-PCV-455A CLOSED, 1-PCV-456 OPEN C. 1-PCV-455A OPEN, 1-PCV-456 CLOSED D. 1-PCV-455A CLOSED, 1-PCV-456 CLOSED Proposed Answer: B Page 9 of 37 CPNPP 2014 NRC RO Written Exam Worksheet 31 to 40

ES-401 RO Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible because PORV PCV-456 will open due to the failure high of the pressure instrument., however, PORV PCV-455A would remain closed.

B. Correct. 1-PT-403 failing high will result in PORV-456 OPENING if LTOP system is armed. RCS Temperature is less than 350°F. PORV 455A will not open as 1-PT-0405 indicates normally.

C. Incorrect. Plausible because PORV PCV-455A would open if 1-PT-405 had failed, however, the failure only affects PORV 456..

D. Incorrect. Plausible because the spray valves open, the Reactor trips on low pressure and PORV PCV-455A will close, however, not in the order listed and PORV PCV-456 remains closed.

Technical Reference(s) ABN-715, Step 2.2 Attached w/ Revision: See LO21.SYS.PP2 Comments / Reference Proposed references to be provided during examination: None Learning Objective: EXPLAIN the instrumentation and controls of the Low Temperature Overpressure Protection System and PREDICT the system response.

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 10 of 37 CPNPP 2014 NRC RO Written Exam Worksheet 31 to 40

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

ABN-715, Step 2.2 Revision: 5 Page 11 of 37 CPNPP 2014 NRC RO Written Exam Worksheet 31 to 40

ES-401 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. Date: 3/3/2014 Tier 2 Change: 2 Group 1 K/A 012 G 2.1.32 Level of Difficulty: 2 Importance Rating 3.8 Reactor Protection System: Conduct of Operations: Ability to explain and apply system limits and precautions Proposed Question: 35 With Train A of Solid State Protection System (SSPS) being tested with its Input Error Inhibit Switch in INHIBIT and its Mode Selector Switch in TEST, Train A Protection can have...

A. ...an auto Safety Injection actuation and an auto Reactor Trip.

B. ...a manual Safety Injection actuation and a manual Reactor Trip.

C. ...a manual Reactor Trip and NO Safety Injection actuation.

D. ...NO Reactor Trip and NO Safety Injection actuation.

Proposed Answer: C Explanation:

A. Incorrect. Plausible if thought that placing the Input Error Inhibit Switch in INHIBIT would still allow an automatic Reactor Trip, however, the purpose of this switch is to prevent any external signals other than manual from being input into the Solid State Protection System. This condition includes an automatic Safety Injection (SI).

B. Incorrect. Plausible because a manual Reactor Trip is still available in this condition, however, with the Mode Selector Switch in TEST it prevents any Engineered Safety Feature (ESF) slave relay actuations such as SI.

C. Correct. As stated in the PRECAUTIONS of SOP-711A, Solid State Protection System, having the Mode Selector Switch in TEST while the Input Air Inhibit Switch is in INHIBIT blocks any ESF slave relay actuation such as Safety Injection. A manual Reactor Trip is still available in this condition.

D. Incorrect. Plausible because no SI actuation will occur because the Mode Selector Switch is in the TEST position, however, a manual Reactor Trip is still available in this condition.

Technical Reference(s) SOP-711A, Section 3.0 & Step 4.2.2 Attached w/ Revision: See OPT-447A, Steps 5.1.5 & 5.1.6 Comments / Reference Proposed references to be provided during examination: None Learning Objective: EXPLAIN the instrumentation and controls of the Solid State Protection System and PREDICT the system response.

Question Source: Bank ILOT5653 Modified Bank (Note changes or attach parent)

New Page 12 of 37 CPNPP 2014 NRC RO Written Exam Worksheet 31 to 40

ES-401 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 55.43 Page 13 of 37 CPNPP 2014 NRC RO Written Exam Worksheet 31 to 40

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

SOP-711A, Section 3.0 Revision: 9 Page 14 of 37 CPNPP 2014 NRC RO Written Exam Worksheet 31 to 40

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

SOP-711A, Step 4.2.2 Revision: 9 Comments /

Reference:

OPT-447A, Steps 5.1.5 & 5.1.6 Revision: 10 Page 15 of 37 CPNPP 2014 NRC RO Written Exam Worksheet 31 to 40

ES-401 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. Date: 5/13/2014 Tier 2 Change: 4 Group 1 K/A 013 K3.01 Level of Difficulty: 3 Importance Rating 4.4 Engineered Safety Features Actuation System: Knowledge of the effect that a loss or malfunction of the ESFAS will have on the following: Fuel Proposed Question: 36 Given the following conditions:

Unit 1 has experienced a 1 inch Small Break Loss of Coolant Accident.

A Loss of Offsite Power has occurred.

Bus 1EA1 has experienced an 86-1 lockout.

Station Service Water Pump 1-02 has tripped.

Reactor Coolant System (RCS) subcooling is currently 15°F and trending toward 0°F.

Given the current plant conditions, if the operators are unable to depressurize the RCS to less than 650 psig, what are the expected consequences?

A. As long as three of four Safety Injection Accumulators inject adequate core cooling will be maintained.

B. Natural Circulation will provide long term core cooling as long as an adequate Secondary Heat Sink is maintained.

C. As long as all Steam Generator safeties remain operable total heat removal capacity is great enough to prevent fuel damage.

D. Reactor Coolant System inventory loss will result in inadequate core cooling and fuel damage is expected.

Proposed Answer: D Page 16 of 37 CPNPP 2014 NRC RO Written Exam Worksheet 31 to 40

ES-401 RO Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible because if the RCS is depressurized to the point where the accumulators inject adequate inventory would exist until break flow removes enough inventory to uncover the core.

The stem states that the RCS cannot be depressurized by the operators to less than 650 psig which is above the accumulator injection pressure. If the RCS is not depressurized to the accumulator injection pressure additional inventory will not be provided by the accumulators.

B. Incorrect. Plausible because if a secondary heat sink is maintained, adequate heat removal will also occur as long as the loop remains filled to transport the heat from the core to the steam generators, however, without makeup the loops will eventually drain and heat removal via the steam generators will be ineffective no matter what level remains in the steam generators.

C. Incorrect. Plausible because heat removal through the safeties does provide cooling until RCS inventory is depleted to the point that coupling with the SGs no longer exists .

D. Correct. Given the degraded RCS and the inability to depressurize the RCS by the operators fuel damage is expected due to loss of RCS inventory.

Technical Reference(s) LO21.MCO.MI2, Pages 10, 15, & 16 Attached w/ Revision: See Comments / Reference Proposed references to be provided during examination: None Learning Objective: EVALUATE plant conditions to DETECT conditions which could lead to core damage from a lack of adequate core cooling and DETERMINE the appropriate mitigation strategies consistent with the Functional Restoration Guidelines.

Question Source: Bank Modified Bank ILOT5731 (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 Page 17 of 37 CPNPP 2014 NRC RO Written Exam Worksheet 31 to 40

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

LO21.MCO.MI2, Page 10 Revision: 04/30/04 Suppose that a 1-inch cold leg break occurs after the plant has established a very high power history (high decay heat levels are present). Suppose also that both the SG Atmospheric relief valves (ARVs) and the steam dump valves are inoperable, and only one train of ECCS is available. The plant begins to depressurize; a reactor trip, turbine trip, and SI occur. More subcooled liquid volume leaves the break than is being added. The pressurizer continues to empty and pressure steadily drops. As the pressurizer empties, the rate of pressure decrease suddenly rises since a relatively small quantity of vapor is produced from surge-line flashing.

Eventually, saturation conditions are reached in the hot legs and vessel outlet plenum and soon after in the entire primary. The break is small and is unable to remove all the decay heat. Thus, the break is insufficient to continue depressurizing the plant. Another heat sink must function to prevent the plant from reaching an over temperature condition. Using conservative assumptions, the SG safety valves are the only available heat sink. Assume that the lift pressure of the lowest safety valve is 1200 psia, which corresponds to a SG saturation temperature of about 567.2 °F.

The corresponding primary side temperature is about 567.8 °F. Since the RCS is saturated, this corresponds to a primary system pressure of about 1205 psia. The plant does not depressurize below this point as long as the safety valves are needed as a heat sink. Under the worst-case conditions, the plant may remain in this state for up to one day on a 1-inch break. Therefore, heat removal, limits plant depressurization. With primary pressure remaining high, volume flow out of the break exceeds SI flow (with plant pressure and temperature stabilized). This situation can occur when the combination of break volume removal and SG safety valve thermal energy removal is sufficient to terminate any plant pressure rise, but not sufficient to totally remove all decay heat, i.e., heat in exceeds heat out. Extended operation in this manner would obviously be undesirable, since the system mass inventory would eventually deplete. As it is, the vapor volume of the coolant system continuously rises and eventually (due to the decay heat drop-off) the reactor coolant pressure and temperature begins to decrease and SI flow increases. As long as SI flow is NOT terminated, adequate core cooling is maintained. In any case, the design of the ECCS must be such that, in the long run, it can inject more water into the RCS than is being lost out of the break.

Page 18 of 37 CPNPP 2014 NRC RO Written Exam Worksheet 31 to 40

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

LO21.MCO.MI2, Pages 15 & 16 Revision: 04/30/04 1-INCH BREAK WITH DEGRADED ECCS Assumptions Assumptions remain the same as in the 4 inch analysis. Basically, a 1 inch cold leg break occurs in a four loop plant at 100 percent power with no high head ECCS and no accumulators. The steam dumps remain available because off site power remains available. However, the RCPs trip during the accident.

Analysis (Figures 8, 9 and 10)

When the break occurs, RCS pressure inventory decrease. At 143 seconds, the low RCS pressure actuates reactor and turbine trips. The steam dumps actuate and reduce RCS temperature to no load.

At 376 seconds, the RCPs trip. The decreased core flow increases core fluid enthalpy and void fraction. As the RCPs coastdown, forced flow decreases. The increased core enthalpy increases natural circulation flow. At 550 seconds, the increased void fraction starts two phase natural circulation. 1000 seconds into the event, the RCS pressure stabilizes based on the heat generation and removal relationship.

At 1500 seconds, the operators place the steam dump in pressure control mode. In addition, sufficient void fraction occurs in the RCS loops that the SGs cannot condense all of the steam. So the cold side of the U tubes increases in void fraction. Core mixture level starts to decrease from the inventory loss.

As the core mixture level approaches the hot leg nozzles, the rate of level decrease slows because the drainage from the U tubes makes a significant difference.

At 4800 seconds, two phase natural circulation stops because of the inability of the SGs to condense the steam. At that time, reflux core cooling starts. In reflux, steam in the U tubes condenses. Steam from the vessel passes along the top of the hot leg, condenses in the U tubes, and returns as condensate back to the vessel along the bottom of the hot leg. Condensate in the cold side drains into the loop seal, which pulses water into the vessel downcomer. Although less efficient than natural circulation, reflux transfers heat to the SGs and cools the core.

At 7800 the core mixture level decreases below the flow holes in the guide tubes of the reactor upper internals. When the level uncovers the holes, the vessel steam enters the guide tubes, passes up the tubes, enters the head, and then passes backward through the head cooling holes in the core barrel flange. The steam then enters the cold leg and passes out the break. Similar to the clearing of the loop seal, this provides a direct vent from the reactor. The RCS pressure and core mixture levels begin to decrease. Reflux flow slows, and core mixture level starts another decrease at about 8200 seconds.

Core mixture level continuously decreases below the top of the fuel and down to the bottom of the fuel.

At 9300 seconds, the upper core uncovers, decreasing decay heat transferred to the fluid. The fuel and clad retain more of the heat, increasing their temperatures. Heat removal from the coolant now exceeds heat transferred into the coolant, decreasing RCS pressure. When RCS pressure decreases, RCS temperature decreases below SG temperature. This slows reflux flow, which was adding liquid to the vessel. So core mixture level and RCS pressure decrease faster.

Page 19 of 37 CPNPP 2014 NRC RO Written Exam Worksheet 31 to 40

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

LO21.MCO.MI2, Pages 15 & 16 Revision: 04/30/04 The fluid temperature basically follows RCS pressure once saturation occurs early in the accident. The fluid temperature breaks away from RCS pressure at 10,250 seconds, which is after the core uncovery starts and after the upper core uncovers at 9300 seconds. For the period between core uncovery and fluid temperature increase, a stabilized fluid temperature occurs because of the reflux flow. The reflux flow, limited as it is, cools the upper core fluid. However, at 10,250 seconds, the center of the core uncovers. The steam produced in the core increases in temperature as it passes over the upper core, producing superheated steam. The superheated steam boils the reflux water, preventing core cooling and increasing the core fluid temperature. At 11,000 seconds, the upper core fluid temperature exceeds 1200 °F. With no ECCS flow, core uncovery and the core fluid temperature increase continues. At 12,000 seconds, RCS pressure still remains higher than the low head ECCS shutoff head.

Original Question: CPNPP Exam Bank ILOT5731 For a small break LOCA (1 inch) where high head safety injection is not available, what is the expected plant condition with no operator action?

A. Low head safety injection will mitigate the LOCA and no core damage will result.

B. The core will remain covered but slight core damage will result.

C. The core will briefly uncover and slight core damage will result.

D. The core will completely uncover and severe core damage will result.

Answer: D Page 20 of 37 CPNPP 2014 NRC RO Written Exam Worksheet 31 to 40

ES-401 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. Date: 5/15/2014 Tier 2 Change: 4 Group 1 K/A 022 K4.03 Level of Difficulty: 3 Importance Rating 3.6 Containment Cooling System: Knowledge of the CCS design feature(s) and/or interlock(s) that provide for the following:

Automatic containment isolation Proposed Question: 37 Given the following conditions:

A Safety Injection occurred on Unit 1.

While performing EOP-0.0A, Reactor Trip or Safety Injection, Attachment 2, Safety Injection Actuation Alignment, the Balance of Plant operator verifies that Containment Ventilation Isolation (CVI) is complete for Unit 1 AND Unit 2.

Unit 2 is in MODE 6 with a Containment Purge in progress.

Containment Ventilation Isolation verification ensures which of the following design features is met?

A. Unit 1 Containment radiation release path is isolated AND a negative pressure boundary is maintained in the Safeguards and Auxiliary buildings.

B. Unit 1 Containment Phase A isolation actuated components have operated as expected to isolate Containment Ventilation.

C. Unit 2 Containment purge valves automatically closed AND a negative pressure boundary is maintained in the Safeguards and Auxiliary buildings.

D. Unit 2 Containment high radiation signal actuated components operated as expected to isolate Containment Ventilation.

Proposed Answer: A Page 21 of 37 CPNPP 2014 NRC RO Written Exam Worksheet 31 to 40

ES-401 RO Written Exam Worksheet Form ES-401-5 Explanation:

A. Correct. Safety Injection causes a Containment Phase A signal which results in a Containment Ventillation Isolation (CVI) signal being generated. This signal closes the Unit 1 Containment Pressure relief dampers automatically. The Containment Purge Dampers on Unit 2 do not receive an automatic closure signal and must be verified to ensure that the Primary Plant Exhaust Fans can maintain a negative pressure in the Auxilliary and Safeguards building.

B. Incorrect. Plausible because the Containment Pressure Relief dampers automatically isolate following a Safety Injection signal, as a result of the Containment Ventillation Isolation (CVI) signal being generated. Phase A does not directly close these dampers, therefore this answer is not correct..

C. Incorrect. Plausible because Unit 2 Containment Purge dampers would not be closed as a result of the Safety Injection, however the reason is not to isolate just the Unit 2 release path, but to maintain the negative pressure in the Safeguards and Auxilliary buildings. The misconception of why Unit 2 components are checked could lead to choosing this distractor.

D. Incorrect. Plausible because Containment Ventillation Isolation does result from High Radiation in Containment, however, this is not the reason or means of isolation in this condition. The CVI was a result of a Safety Injection, not a High Radiation condition.

Technical Reference(s) LO21.SYS.CL1, Pages 16 Attached w/ Revision: See EOP-0.0A, Attachment 10, Pages 29 & 38 Comments / Reference Proposed references to be provided during examination: None Learning Objective: DESCRIBE the components of the Containment Ventilation system including interrelations with other systems to include interlocks and control loops.

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 9 55.43 Comments /

Reference:

LO21.SYS.CL1, Page 16 Revision: 05/02/11 In the event of a Safety Injection, Containment ventilation isolation is verified on both units. The Safety Injection will trip all the Primary Plant Exhaust Fans and start the four (4) ESF Exhaust Fans. This configuration does not provide enough airflow to maintain a negative pressure in the Safeguards and Auxiliary Buildings if a Containment purge is in progress on the unaffected unit. A Containment Ventilation Isolation signal will cause the valves or isolation dampers on the affected unit to close.

Page 22 of 37 CPNPP 2014 NRC RO Written Exam Worksheet 31 to 40

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

EOP-0.0A, Att 10 Page 29 Revision: 8, PCN 9 Page 23 of 37 CPNPP 2014 NRC RO Written Exam Worksheet 31 to 40

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

EOP-0.0A, Att 10 Page 38 Revision: 8 PCN 9 Page 24 of 37 CPNPP 2014 NRC RO Written Exam Worksheet 31 to 40

ES-401 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. Date: 5/13/2014 Tier 2 Change: 4 Group 1 K/A 026 A1.06 Level of Difficulty: 4 Importance Rating 2.7 Containment Spray System: Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the CSS controls including: Containment spray pump cooling Proposed Question: 38 Given the following conditions:

A Large Break Loss of Coolant Accident has occurred on Unit 2.

Both Trains of Containment Spray have just actuated.

Train B Component Cooling Water Pump has tripped.

Which of the following identifies the condition of Train B Containment Spray Pump (CSP) cooling?

Cooling has been lost to the CSP________________.

CSP motor cooling is _______________.

A. Bearing coolers Degraded B. Bearing coolers Normal C. Seal coolers Degraded D. Seal coolers Normal Proposed Answer: C Page 25 of 37 CPNPP 2014 NRC RO Written Exam Worksheet 31 to 40

ES-401 RO Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible since CCW supplies cooling to the pump seal coolers, but SSW supplies cooling to the pump bearing coolers, therefore cooling has not been lost to the bearing coolers.

CSP motor cooling is degraded as the Safety Chiller which supplies chilled water for the room coolers will trip when the CCW Pump trips, thus the chilled water flowing to the room coolers will steadily increase in temperature resulting in degraded cooling to the motors which are cooled via the room coolers.

B. Incorrect. Plausible since CCW supplies cooling to the pump seal coolers, but SSW supplies cooling to the pump bearing coolers, therefore cooling has not been lost to the bearing coolers.

CSP motor cooling is degraded as the Safety Chiller which supplies chilled water for the room coolers will trip when the CCW Pump trips, thus the chilled water flowing to the room coolers will steadily increase in temperature resulting in degraded cooling to the motors which are cooled via the room coolers. CSP motor cooling remaining normal is plausible as a chain of events must be considered to determine that room cooling is being lost from the lost of the safety chillers.

C. Correct. CCW supplies cooling to the CSP seal coolers which has been lost with the trip of the CCW pump. CSP motor cooling is degraded as the Safety Chiller which supplies chilled water for the room coolers will trip when the CCW Pump trips, thus the chilled water flowing to the room coolers will steadily increase in temperature resulting in degraded cooling to the motors which are cooled via the room coolers.

D. Incorrect. Plausible since CCW supplies cooling to the CSP seal coolers which has been lost with the trip of the CCW pump. CSP motor cooling is degraded as the Safety Chiller which supplies chilled water for the room coolers will trip when the CCW Pump trips, thus the chilled water flowing to the room coolers will steadily increase in temperature resulting in degraded cooling to the motors which are cooled via the room coolers. CSP motor cooling remaining normal is plausible as a chain of events must be considered to determine that room cooling is being lost from the lost of the safety chillers.

Technical Reference(s) SOP-204B, Step 2.2.2 Attached w/ Revision: See LO21.SYS.CT1 page 11 Comments / Reference LO21.SYS.CH1 page 12 Proposed references to be provided during examination: None Learning Objective: DEMONSTRATE an understanding of the components of the Containment Spray system including interrelations with other systems to include interlocks and control loops.

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 55.43 Page 26 of 37 CPNPP 2014 NRC RO Written Exam Worksheet 31 to 40

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

SOP-204B, Step 2.2.2 Revision: 6 Page 27 of 37 CPNPP 2014 NRC RO Written Exam Worksheet 31 to 40

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

LO21.SYS.CH1 Page 12 Revision: 3/1/2011 Chiller Trips The chiller trips are divided into two categories: those that require a reset and those that do not require reset. The following trips require a reset after the signal clears:

LO oil pressure LO evaporator pressure HI condenser pressure HI motor temperature HI discharge temperature Motor overload HI oil temperature The trips that DO NOT require a reset and will allow an automatic restart of the compressor as soon as the signal clears include:

LO chilled water temperature LO CCW flow LO chilled water flow Comments /

Reference:

LO21.SYS.CT1 Page 11 Revision: 5/2/2011 CONTAINMENT SPRAY PUMPS (FIGURE 8)

Four Containment Spray Pumps are provided for each unit (two per train) and are of the horizontal, double-suction, centrifugal type. These pumps are designed to provide sufficient flow into Containment during accident conditions to both cool the Containment atmosphere and remove radioactive iodine. Each of the two pumps per train is 50% pumps and have a capacity of approximately 3000gpm at 260 psig discharge pressure. The pumps are designed to operate over a widely varied temperature ranging from 40 F (minimum RWST temperature) to 280 F (containment maximum design temperature).

Each of the pumps has separate bearing coolers (2) and seal coolers (2) mounted at the pump skid itself. The bearing coolers are supplied with cooling water from the related train of Station Service Water. Component Cooling Water from the same train supplies the seal coolers. ABN-501 provides alternate means of supplying cooling to the pump bearing coolers. Each set of train related pumps is located in a separate room in the safeguards building 773' elevation.

The pumps are driven by horizontal 700 hp electric motors powered from their respective 6.9 kv safeguards bus. The room/motors are kept cool by two room coolers supplied from their train related Safety Chill Water system. Both room coolers receive a start signal when either pump breaker closes.

The pump breaker closing also provides a computer input and MLB (MLB-4A-1/4B-1) input for pump running indication.

Page 28 of 37 CPNPP 2014 NRC RO Written Exam Worksheet 31 to 40

ES-401 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. Date: 3/4/2014 Tier 2 Change: 2 Group 1 K/A 039 K5.05 Level of Difficulty: 3 Importance Rating 2.7 Main and Reheat Steam System: Knowledge of the operational implications of the following concepts as they apply to the MRSS: Bases for RCS cooldown limits Proposed Question: 39 Given the following conditions:

Unit 1 has experienced a Reactor Trip.

A plant cooldown to MODE 5 is in progress using the Steam Dumps.

Which of the following describes the Reactor Coolant System cooldown rate limit and the bases for the limit?

A. 100°F/hr to minimize pressurizer fatigue cycle events.

B. 200°F/hr to minimize pressurizer fatigue cycle events.

C. 100°F/hr to provide margin to non-ductile failure of the reactor vessel.

D. 200°F/hr to provide margin to non-ductile failure of the reactor vessel.

Proposed Answer: C Explanation:

A. Incorrect. Plausible because the RCS cooldown limit is 100°F/hr, however, the bases for the limit is to provide margin for brittle failure of the reactor vessel not limit pressurizer cyclic fatigue.

B. Incorrect. Plausible because the Pressurizer cooldown limit is 200°F/hr, however, the bases for the limit is to provide margin for brittle failure of the reactor vessel not limit pressurizer cyclic fatigue.

C. Correct. The RCS cooldown limit is 100°F/hr, and the bases for the limit is to provide marging to brittle failure of the reactor vessel.

D. Incorrect. Plausible because the Pressurizer cooldown limit is 200°F/hr and the bases for the limit is to provide marging to brittle failure of the reactor vessel.

Page 29 of 37 CPNPP 2014 NRC RO Written Exam Worksheet 31 to 40

ES-401 RO Written Exam Worksheet Form ES-401-5 Technical Reference(s) Technical Specification LCO 3.4.3 Attached w/ Revision: See Technical Requirements Manual TR Comments / Reference 13.4.34 Proposed references to be provided during examination: None Learning Objective: DEMONSTRATE an understanding of the components of the Main Steam System including interrelations with other systems to include interlocks and control loops.

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

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 4 55.43 Page 30 of 37 CPNPP 2014 NRC RO Written Exam Worksheet 31 to 40

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

Technical Specification Bases LCO 3.4.3 Revision: 68 Page 31 of 37 CPNPP 2014 NRC RO Written Exam Worksheet 31 to 40

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

Pressure & Temperature Limits Report Revision: 3 Comments /

Reference:

Technical Requirements Manual TR 13.4.34 Revision: 84 Page 32 of 37 CPNPP 2014 NRC RO Written Exam Worksheet 31 to 40

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

Technical Requirements Manual TR 13.4.34 Bases Revision: 85 Page 33 of 37 CPNPP 2014 NRC RO Written Exam Worksheet 31 to 40

ES-401 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. Date: 3/26/2014 Tier 2 Change: 3 Group 1 K/A 059 A2.04 Level of Difficulty: 4 Importance Rating 2.9 Main Feedwater System: Ability to (a) predict the impacts of the following malfunctions or operations on the MFW; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Feeding a dry SG Proposed Question: 40 Given the following conditions:

Unit 1 has experienced a Loss of Heat Sink.

Bleed and Feed cooling was established when all Steam Generators (SG) had insufficient level.

Main Feedwater flow has been established.

Reactor Coolant System Hot Leg temperatures are all indicating 560°F and STABLE.

Core Exit Thermocouple temperatures are all indicating between 555°F and 565°F and STABLE.

ALL Steam Generator Wide Range levels are 0%.

Which of the following actions is required in accordance with FRH-0.1A, Response to Loss of Secondary Heat Sink?

Establish Main Feedwater flow to...

A. ...ONE Steam Generator at a rate not to exceed 100 gpm.

B. ...ALL Steam Generators not to exceed 100 gpm per SG.

C. ...ONE Steam Generator at maximum available feed flow.

D. ...ALL Steam Generators at maximum available feed flow.

Proposed Answer: A Page 34 of 37 CPNPP 2014 NRC RO Written Exam Worksheet 31 to 40

ES-401 RO Written Exam Worksheet Form ES-401-5 Explanation:

A. Correct. IAW Step 26, Table 1 of FRH-0.1A, Main Feedwater should be established to only one SG at a rate not to exceed 100 gpm.

B. Incorrect. Plausible because this is the proper feed rate and the misconception that reestablishing cooling in all loops is credible.

C. Incorrect. Plausible because IAW Step 26, Table 1 of FRH-0.1A, Main Feedwater would be established to only one SG at maximum rate if the RCS temperature was increasing, indicating a more critical situation.

D. Incorrect. Plausible because the maximum rate would be used if RCS temperature was increasing and the misconception that reestablishing cooling in all loops is credible.

Technical Reference(s) FRH-0.1A, Step 26.b, Table 1 Attached w/ Revision: See FRH-0.1A, Attachment 4, Step 26 Bases Comments / Reference Proposed references to be provided during examination: None Learning Objective: Given a procedural Step, NOTE, or CAUTION, DISCUSS the reason or basis for the Step, NOTE, or CAUTION in FRH-0.1, Response to a Loss of Secondary Heat Sink.

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 35 of 37 CPNPP 2014 NRC RO Written Exam Worksheet 31 to 40

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

FRH-0.1A, Step 26.b, Table 1 Revision: 8 Page 36 of 37 CPNPP 2014 NRC RO Written Exam Worksheet 31 to 40

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

FRH-0.1A, Attachment 4, Step 26 Bases Revision: 8 Page 37 of 37 CPNPP 2014 NRC RO Written Exam Worksheet 31 to 40

ES-401 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. Date: 3/4/2014 Tier 2 Change: 2 Group 1 K/A 061 K6.02 Level of Difficulty: 3 Importance Rating 2.6 Auxiliary/Emergency Feedwater System: Knowledge of the effect that a loss or malfunction of the following will have on the AFW components: Pumps Proposed Question: 41 Given the following conditions:

Unit 1 is in a condition requiring Turbine Driven Auxiliary Feedwater Pump (TDAFWP) operation.

The TDAFWP has tripped due to an overspeed condition.

The cause of the overspeed trip has been corrected.

The Safeguards Building NEO has been directed to reset 1-HV-2452 AFWPT 1-01 TRIP AND THROT VLV, in accordance with ABN-305, Turbine Driven Auxiliary Feedwater Pump Malfunction.

The actions of ABN-305 have been completed prior to the NEO resetting the valve.

Which of the following describes the response of the TDAFWP as the NEO opens 1-HV-2452 in accordance with ABN-305?

The TDAFW Pump Turbine will A. remain at approximately 0 rpm and speed will have to be increased by the Reactor Operator.

B. increase speed to approximately 2000 rpm and hold speed.

C. remain at approximately 0 rpm until both steam supply valves are re-opened.

D. increase speed to approximately 4075 rpm and hold speed.

Proposed Answer: B Page 1 of 52 CPNPP 2014 NRC RO Written Exam Worksheet 41 to 50

ES-401 RO Written Exam Worksheet Form ES-401-5 Explanation: Refer to reference ABN-305, Step 4.3.2 RNO to follow the logic for the correct answer.

A. Incorrect. Plausible if thought the TDAFWP would remain at 0 speed until the RO takes further action. However, the pump will immediately go to a minimum speed of 2000 rpm at minimum demand.

B. Correct. Following guidance of ABN-305, Step 4.3.2 RNO, the reactor operator lowers 1-SK-2452A AFWPT SPD CTRL to 0% output. This action limits the speed of the turbine to 2000 RPM when the NEO opens 1-HS-2452, AFWPT Trip and Throttle Valve.

C. Incorrect. Plausible if thought the TDAFWP would remain at 0 speed until the RO takes further action or until both steam admission valves are open. However, the pump will run at full speed on either steam admission valve being open and the pump will immediately go to a minimum speed of 2000 rpm at minimum demand.

D. Incorrect. Plausible if a loss of power to either Safeguards Bus were to occur the TDAFWP would have a full speed signal applied, however, because the actions of ABN-305 have been performed the turbine speed would only increase to ~2000 rpm.

Technical Reference(s) ABN-305, Step 4.3.2 RNO Attached w/ Revision: See SOP-304A, Step 5.1.2.F Comments / Reference ABN-602, Step 2.3.3 NOTE LO21.SYS.AF1, Page 26 Proposed references to be provided during examination: None Learning Objective: ANALYZE the response to Turbine Driven Auxiliary Feedwater Pump Malfunction in accordance with ABN-305, Auxiliary Feedwater System Malfunction.

Question Source: Bank ILOT0082 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 52 CPNPP 2014 NRC RO Written Exam Worksheet 41 to 50

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

ABN-305, Step 4.3.2 RNO Revision: 7 Page 3 of 52 CPNPP 2014 NRC RO Written Exam Worksheet 41 to 50

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

ABN-602, Step 2.3.3 NOTE Revision: 8 Page 4 of 52 CPNPP 2014 NRC RO Written Exam Worksheet 41 to 50

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

SOP-304A, Step 5.1.2.F Revision: 17 Page 5 of 52 CPNPP 2014 NRC RO Written Exam Worksheet 41 to 50

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

LO21.SYS.AF1, Page 26 Revision: 05/11/11 TDAFWP OVERSPEED TRIP Annunciator window 4.6, TD AFWP OVRSPD TRIP, on ALB-8B provides indication of a trip of the TDAFW Pump. When the TDAFW pump turbine trips, the AFWPT TRIP light on HS-2452F will be ON. The MDAFW pumps are started per SOP-304A(B) as necessary to maintain SG levels and the AFWPT speed controller on CB-09 is lowered to 0%, this will limit the initial speed to 2000 RPM when the PEO opens the T&TV. When directed, the AO will reset the overspeed trip linkage and slowly open the T&TV using the job aid similar to ABN-305, ATT. 1 posted on the wall in the TDAFWP room near the T&TV. When resetting the overspeed trip, it is important to note that the tappet nut and the emergency head lever have beveled mating surfaces. For the overspeed trip to reset, the tappet nut must drop into place to maintain the emergency head lever trip mechanism upright. With the trip mechanism upright, the T&TV can be repositioned which allows the emergency connecting rod to remain in the reset "standby" condition.

As the AO opens the T&TV, the governor valve stem should be monitored for movement to ensure that the valve stem is operating smoothly to maintain turbine speed. As the T&TV is opened the turbine will slowly accelerate to approximately 2000 rpm at which point the governor should maintain speed. The T&TV can then be fully opened. The AFWPT speed controller is set to zero and the T&TV is opened slowly in order to minimize the potential for turbine overspeed. After verifying that the T&TV indicates open at HS-2452G/H and the valve red lights are lit, the Control Room operator will adjust speed controller SK-2452A, AFWPT SPD CTRL on CB-09 to the desired output.

Page 6 of 52 CPNPP 2014 NRC RO Written Exam Worksheet 41 to 50

ES-401 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. Date: 3/26/2014 Tier 2 Change: 3 Group 1 K/A 062 A4.01 Level of Difficulty: 3 Importance Rating 3.3 AC Electrical Distribution System: Ability to manually operate and/or monitor in the control room: All breakers (including available switchyard)

Proposed Question: 42 The Balance of Plant Operator is transferring power from 2UT to 2ST during a normal plant shutdown in accordance with SOP-603B, 6900 V Switchgear.

CS-2A1-2, INCOMING BKR 2A1-2 is taken to the CLOSE position.

The following indications are observed WITHOUT the transfer occurring:

Red position indicating light for breaker 2A1-1 (from 2UT) is LIT.

Green position indicating light for breaker 2A1-2 (from 2ST) is LIT.

2A1 bus voltage is 6.9 KV.

Synchroscope is at the 12 o'clock position.

Synchroscope lights are DARK.

Zero running volts are indicated on RUNNING VOLT V-RUN.

Which of the following prevented the bus transfer?

A. 2ST has NO power and thus the transfer is blocked.

B. The anti-parallel relay did NOT open 2UT supply breaker 2A1-1 to allow 2ST supply breaker 2A1-2 to close.

C. SS-2A1-2, BKR 2A1-2 SYNCHROSCOPE is in the OFF position.

D. The spring charging motor for 2UT supply breaker 2A1-1 is in the OFF position.

Proposed Answer: C Page 7 of 52 CPNPP 2014 NRC RO Written Exam Worksheet 41 to 50

ES-401 RO Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible because with zero running volts indicated, a misconception that the source has no power is possible.

B. Incorrect. Plausible because the anti-parallel circuit prevents the two power sources from being paralleled for more than one second, however, the breaker from 2ST never had the opportunity to close because the synchroscope was not energized.

C. Correct. If the synchroscope is at the 12 oclock position then the synchroscope lights should be at maximum brightness. If the lights are dark then the synchroscope has not been energized.

D. Incorrect. Plausible because the status of the charging motor can affect the breakers ability to close, however, with a green light indication on the breaker it can be assumed that the spring charging motor is in the ON position. Even if it were in the OFF position there is sufficient charge to cycle the breaker at least once.

Technical Reference(s) LO21.SYS.AC2, Pages 30 & 61 Attached w/ Revision: See SOP-603, Step 5.1.1 Comments / Reference SOP-603, Attachment 1 Proposed references to be provided during examination: None Learning Objective: EXPLAIN the instrumentation and controls of the 6.9 KV and 480 V Electrical Distribution System and PREDICT the system response.

Question Source: Bank ILOT4006 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:

LO21.SYS.AC2, Page 61 Revision: 04/28/11 In the synchronization section, the synchroscope displays the phase relationship between the two sources while paralleling. The synchroscope is energized by placing the appropriate Synchroscope Switch for the breaker to be closed to ON. Running frequency and voltage meters display to the operator the on-line sources frequency and voltage (normally bus frequency and voltage). Incoming frequency and voltage meters display to the operator the oncoming sources frequency and voltage (normally DG frequency and voltage). Running and Incoming light indication below the synchroscope indicate phase differential (voltage) between the sources being paralleled. With the sources 180° out of phase, these lights would not be lit. DG control switches to raise and lower DG speed and voltage are provided to allow the operator to adjust parameters to allow the DG to pickup load when being paralleled.

Page 8 of 52 CPNPP 2014 NRC RO Written Exam Worksheet 41 to 50

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

LO21.SYS.AC2, Page 30 Revision: 04/28/11 Anti-Parallel Interlock All of the 6.9KV buses have this feature. As soon as the alternate (or normal) power supply is paralleled and the breaker closed, the normal (or alternate) power supply breaker will open, preventing the two power sources from being paralleled for more than 1 second. This interlock does not affect the diesel breaker.

Comments /

Reference:

SOP-603, Step 5.1.1 Revision: 10 Page 9 of 52 CPNPP 2014 NRC RO Written Exam Worksheet 41 to 50

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

SOP-603, Attachment 1 Revision: 10 Page 10 of 52 CPNPP 2014 NRC RO Written Exam Worksheet 41 to 50

ES-401 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. Date: 5/14/2014 Tier 2 Change: 3 Group 1 K/A 063 K2.01 Level of Difficulty: 3 Importance Rating 2.9 DC Electrical Distribution System: Knowledge of bus power supplies to the following: Major DC loads Proposed Question: 43 Which of the following lists the Unit 2 DC Bus which supplies power to the following loads?

Main Turbine Emergency Main Turbine Electro-Hydraulic Lube Oil Pump 2-01 Control Unit A. 2D1 2D2 B. 2D2 2D1 C. 2ED1 2ED2 D. 2ED2 2ED1 Proposed Answer: B Explanation:

A. Incorrect. Plausible because it could be thought that 2D1 supplies the large DC pumps and that 2D2 supplies the turbine and generator controllers.

B. Correct. 2D2 is the larger voltage bus which supplies the large DC pumps and 2D1 supplies power to the turbine and generator controllers.

C. Incorrect. Plausible because it could be thought that 2ED1 supplies the large DC pumps and that 2ED2 supplies the turbine and generator controllers.

D. Incorrect. Plausible because it could be thought that 2ED2 supplies the large DC pumps and that 2ED1 supplies the turbine and generator controllers.

Technical Reference(s) LO21.SYS.DC1 Attached w/ Revision: See Comments / Reference Proposed references to be provided during examination: None Learning Objective: COMPREHEND the normal, abnormal and emergency operation of the DC Electrical Distribution System.

Page 11 of 52 CPNPP 2014 NRC RO Written Exam Worksheet 41 to 50

ES-401 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 8 55.43 Page 12 of 52 CPNPP 2014 NRC RO Written Exam Worksheet 41 to 50

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

LO21.SYS.DC1 Revision: 05/05/11 125 VDC SAFEGUARDS DISTRIBUTION MCC uEB3-1 MCC uEB1-1 MCC uEB1-1 MCC uEB3-1 TRAIN A SAFEGUARDS BT uED3 BC uED1-2 BC uED1-1 BT uED1 BC uED3-1 BC uED3-2 125V D.C.

uED1 uED3 uEC3 uEC3 uEC3 uEC3 uEC3 IV uPC1 IV uEC1 IV uEC3 IV uPC3 10 KVA 10 KVA 10 KVA 10 KVA DIST DIST DIST DIST PNL PNL PNL PNL uED1-1 XED1-1 uED1-2 IV uEC1/3 uED3-1 10 KVA uPC1 uEC1 uEC uPC 5 3 MCC uEB2-1 MCC uEB4-1 TRAIN B MCC uEB2-1 MCC uEB4-1 SAFEGUARDS BC uED2-1 BC uED2-2 BT uED2 125V D.C. BC uED4-1 BC uED4-2 BT uED4 uED2 uED4 uEC4 uEC4 uEC4 uEC4 uEC4 IV uPC2 IV uEC2 IV uEC4 IV uPC4 10 KVA 10 KVA 10 KVA 10 KVA DIST DIST DIST DIST PNL PNL PNL PNL uED2-1 XED2-1 uED2-2 IV uEC2/4 uED4-1 10 KVA uPC2 uEC2 uEC6 uPC 4

OP51.SYS.DC1.FG02 03-31-09 Page 13 of 52 CPNPP 2014 NRC RO Written Exam Worksheet 41 to 50

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

LO21.SYS.DC1 Revision: 05/05/11 Page 14 of 52 CPNPP 2014 NRC RO Written Exam Worksheet 41 to 50

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

LO21.SYS.DC1 Revision: 05/05/11 Page 15 of 52 CPNPP 2014 NRC RO Written Exam Worksheet 41 to 50

ES-401 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. Date: 5/14/2014 Tier 2 Change: 3 Group 1 K/A 064 K6.08 Level of Difficulty: 3 Importance Rating 3.2 Emergency Diesel Generator System: Knowledge of the effect that a loss or malfunction of the following will have on the EDG system: Fuel oil storage tanks Proposed Question: 44 Given the following conditions:

Unit 1 has completed a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> run of emergency Diesel Generator (DG) 1-02.

Diesel Generator 1-02 Panel, Window 3.3 - LOW LEVEL FUEL STORAGE TANK is LIT.

Level in the DG 1-02 Fuel Oil Storage Tank (FOST) is determined to be 82,000 gallons.

Which of the following is the potential operational impact of the DG FOST level?

The FOST level is A. less than the amount required by OWI-104-26, Control Room Diesel Generator 1-02 Operating Logs and DG 1-02 may not be capable of satisfying the requirement to supply power following a Design Basis Accident for a minimum period of 6 days.

B. less than the amount required by OWI-104-26, Control Room Diesel Generator 1-02 Operating Logs and DG 1-02 may not be capable of satisfying the requirement to supply power following a Design Basis Accident for a minimum period of 7 days.

C. greater than the amount required by OWI-104-26, Control Room Diesel Generator 1-02 Operating Logs and DG 1-02 is capable of satisfying the requirement to supply power following a Design Basis Accident for a minimum period of 6 days.

D. greater than the amount required by OWI-104-26, Control Room Diesel Generator 1-02 Operating Logs and DG 1-02 is capable of satisfying the requirement to supply power following a Design Basis Accident for a minimum period of 7 days.

Proposed Answer: B Page 16 of 52 CPNPP 2014 NRC RO Written Exam Worksheet 41 to 50

ES-401 RO Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible because the level of 82,000 gallons is less than the required amount of 86,000 gallons, but is greater than the amount allowed for up to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> that ensures a 6 day supply.

B. Correct. 82,000 gallons is less than the required amount of 86,000 gallons. 86,000 gallons of fuel oil in the FOST is required to ensure that the DG can supply all required loads following a DBA for a period of 7 days.

C. Incorrect. Plausible because 82,000 gallons is greater than the 74,600 gallons that ensures a 6 day supply, allowed for up to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> when the 7 day supply is unavailable, but 82,000 gallons is still below the required amount.

D. Incorrect. Plausible because 82,000 gallons is greater than the 74,600 gallons that ensures a 6 day supply, allowed for up to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> when the 7 day supply is unavailable, but 82,000 gallons is still below the required amount.

Technical Reference(s) Technical Specification LCO 3.8.3.A Attached w/ Revision: See OWI-104-26 Comments / Reference Technical Specification LCO 3.8.3 Bases Proposed references to be provided during examination: None Learning Objective: APPLY the administrative requirements of the Spent Fuel Pool Cooling and Cleanup system including Technical Specifications, TRM and ODCM.

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 Page 17 of 52 CPNPP 2014 NRC RO Written Exam Worksheet 41 to 50

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

Technical Specification LCO 3.8.3.A, SR 3.8.3.1 Amendment: 161 Page 18 of 52 CPNPP 2014 NRC RO Written Exam Worksheet 41 to 50

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

OWI-104-26 Revision: 11 Page 19 of 52 CPNPP 2014 NRC RO Written Exam Worksheet 41 to 50

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

Technical Specification LCO 3.8.3 Bases Revision: 68 Page 20 of 52 CPNPP 2014 NRC RO Written Exam Worksheet 41 to 50

ES-401 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. Date: 5/15/2014 Tier 2 Change: 3 Group 1 K/A 073 A2.02 Level of Difficulty: 2 Importance Rating 2.7 Process Radiation Monitoring System: Ability to (a) predict the impacts of the following malfunctions or operations on the PRM system; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Detector failure Proposed Question: 45 Given the following condition:

The PC-11 is alarming due to OPERATE FAILURE - CHANNEL NO PULSES RECEIVED on X-RE-5251A, Auxiliary Building Low Volume Waste Monitor.

Which of the following describes the Auxiliary Building Drains response and the actions taken in accordance with ALM-3200, Digital Radiation Monitoring System?

Auxiliary Building Drains will A. remain in the normal alignment and would still initiate automatic actions on high radiation. Notify I&C to investigate loss of counts alarm on PC-11.

B. automatically isolate and the Auxiliary Building Sump Pumps will stop. Verify the automatic actions occurred and stop any waste producing activities.

C. remain in the normal alignment and no automatic actions would occur on high radiation. Locally align Auxiliary Building Drains to the Co-Current Waste System.

D. automatically divert to the Co-Current Waste System. Verify the automatic action occurred and implement the requirements of the Offsite Dose Calculation Manual.

Proposed Answer: D Page 21 of 52 CPNPP 2014 NRC RO Written Exam Worksheet 41 to 50

ES-401 RO Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible because it could be thought that an OPERATE FAILURE would not result in automatic initiation and the high radiation feature is still available. Also the action to have I&C investigate is somewhat correct, however, this monitor in OPERATE FAILURE will initiate the automatic actions and is a required instrument per the Offsite Dose Calculation Manual (ODCM).

B. Incorrect. Plausible because it could be thought that an OPERATE FAILURE would result in these automatic actions and the specified actions make sense to prevent sumps from backing up, however, the loss of pulses is an OPERATE FAILURE condition that results in automatic actions that must be verified. The detector is out of service and the ODCM actions apply.

C. Incorrect. Plausible because the response would be correct for some monitors, however, this monitor in OPERATE FAILURE will initiate the automatic actions and is a required instrument per the ODCM.

D. Correct. The loss of pulses is an OPERATE FAILURE condition that results in automatic actions that must be verified. The detector is out of service and the ODCM actions apply.

Technical Reference(s) ALM-3200, Pages 38, 83, & 102 Attached w/ Revision: See Comments / Reference Proposed references to be provided during examination: None Learning Objective: COMPREHEND the normal, abnormal and emergency operations of the Liquid Waste Systems.

Question Source: Bank ILOT8281 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 13 55.43 Page 22 of 52 CPNPP 2014 NRC RO Written Exam Worksheet 41 to 50

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

ALM-3200, Page 38 Revision: 4 Page 23 of 52 CPNPP 2014 NRC RO Written Exam Worksheet 41 to 50

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

ALM-3200, Page 83 Revision: 4 Page 24 of 52 CPNPP 2014 NRC RO Written Exam Worksheet 41 to 50

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

ALM-3200, Page 102 Revision: 4 Page 25 of 52 CPNPP 2014 NRC RO Written Exam Worksheet 41 to 50

ES-401 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. Date: 3/27/2014 Tier 2 Change: 3 Group 1 K/A 076 K3.07 Level of Difficulty: 3 Importance Rating 3.7 Service Water System: Knowledge of the effect that a loss or malfunction of the SWS will have on the following: ESF loads Proposed Question: 46 Given the following conditions:

Train A is the protected train on both units.

A fault has occurred on XST2.

All equipment responded in accordance with design, with the following noted exception:

The Train B Station Service Water (SSW) Pump on the affected bus failed to start on the Blackout Sequencer.

Which component is operating without cooling until the affected SSW Pump can be started?

A. Emergency Diesel Generator 1-02.

B. Centrifugal Charging Pump 1-02.

C. Emergency Diesel Generator 2-02.

D. Centrifugal Charging Pump 2-02.

Proposed Answer: B Explanation:

A. Incorrect. Plausible because the fault on XST2 will cause a loss of the preferred offsite power supply to 1EA2 but the slow transfer to the alternate offsite power supply will occur prior to the EDG getting a start signal. Therefore, the EDG 1-02 does not start and is thus not running without cooling.

B. Correct. The fault on XST2 will cause a loss of the preferred offsite power supply to 1EA2 which will result in the Blackout Sequencer operating. The Blackout Sequencer will start CCP 1-02 and should also start SSWP 1-02. Since the malfunction is a failure of the SSWP 1-02 to start, CCP 1-02 which is cooled by SSW is running without cooling water.

C. Incorrect. Plausible because the loss of XST2 will affect both units. However, the effect on Unit 2 is the loss of the alternate power supply to 2EA2 and thus the Blackout Sequencer does not operate for Unit 2 and there is no affected train of equipment.

D. Incorrect. Plausible because the loss of XST2 will affect both units. However, the effect on Unit 2 is the loss of the alternate power supply to 2EA2 and thus the Blackout Sequencer does not operate for Unit 2 and there is no affected train of equipment.

Page 26 of 52 CPNPP 2014 NRC RO Written Exam Worksheet 41 to 50

ES-401 RO Written Exam Worksheet Form ES-401-5 Technical Reference(s) ABN-601, Steps 2.1.b & 3.2 Attached w/ Revision: See Comments / Reference Proposed references to be provided during examination: None Learning Objective: ANALYZE the response to Station Service Water Pump Trip in accordance with ABN-501, Station Service Water System Malfunction.

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:

ABN-601, Step 2.1.b Revision: 12 Page 27 of 52 CPNPP 2014 NRC RO Written Exam Worksheet 41 to 50

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

ABN-601, Step 3.2 Revision: 12 Page 28 of 52 CPNPP 2014 NRC RO Written Exam Worksheet 41 to 50

ES-401 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. Date: 3/5/2014 Tier 2 Change: 2 Group 1 K/A 078 K1.02 Level of Difficulty: 2 Importance Rating 2.7 Instrument Air System: Knowledge of the physical connections and/or cause-effect relationships between the IAS and the following systems: Service air Proposed Question: 47 Given the following conditions:

Unit 1 is in MODE 6 during a Refueling Outage.

Work is in progress inside the Steam Generators.

Core offload is in progress.

The Wet Cask Pit is at reduced level for fuel inspection equipment repair.

A loss of Instrument Air has occurred on Unit 1.

Based on the given conditions, which of the following is the impact on the Unit 1 Service Air System?

Service Air...

A. ...is unavailable for use inside the Control Room Envelope Boundary.

B. ...must be aligned to the Wet Cask Pit and Spent Fuel Pool X-01 gates.

C. ...quality must be evaluated prior to aligning to Unit 1 Instrument Air.

D. ...is unavailable for use inside the Unit 1 Containment Building.

Proposed Answer: D Explanation:

A. Incorrect. Plausible because it could be thought that loss of instrument air would cause a loss of ability to supply service air inside the CRE boundary, however the controls for service air use inside the CRE boundary are administrative.

B. Incorrect. Plausible because instrument air supplies the gate seals but compressed air bottles are provided for the gate seals.

C. Incorrect. Plausible because the misconception could exist that the service air compressor could be used as a temporary supply to the instrument air per SOP-301A .

D. Correct. 1-HS-3486, CNTMT SERV AIR ISOL VLV will fail close isolating service air to Containment.

Page 29 of 52 CPNPP 2014 NRC RO Written Exam Worksheet 41 to 50

ES-401 RO Written Exam Worksheet Form ES-401-5 Technical Reference(s) ABN-301, Steps 2.3.2 RNO & 2.3.3 RNO Attached w/ Revision: See ABN-301, Attachment 1 Comments / Reference SOP-509A SOP-508 Proposed references to be provided during examination: None Learning Objective: COMPREHEND the normal, abnormal and emergency operation of the Instrument Air 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 X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 9 55.43 Page 30 of 52 CPNPP 2014 NRC RO Written Exam Worksheet 41 to 50

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

ABN-301, Step 2.3.2 RNO Revision: 12 Page 31 of 52 CPNPP 2014 NRC RO Written Exam Worksheet 41 to 50

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

ABN-301, Step 2.3.3 RNO Revision: 12 Page 32 of 52 CPNPP 2014 NRC RO Written Exam Worksheet 41 to 50

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

ABN-301, Attachment 1 Revision: 12 Page 33 of 52 CPNPP 2014 NRC RO Written Exam Worksheet 41 to 50

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

SOP-509A Revision: 22 Page 34 of 52 CPNPP 2014 NRC RO Written Exam Worksheet 41 to 50

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

SOP-508 Revision: 4 Page 35 of 52 CPNPP 2014 NRC RO Written Exam Worksheet 41 to 50

ES-401 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. Date: 3/3/2014 Tier 2 Change: 2 Group 1 K/A 103 A3.01 Level of Difficulty: 3 Importance Rating 3.9 Containment System: Ability to monitor automatic operation of the Containment system, including: Containment isolation Proposed Question: 48 Given the following conditions:

Unit 1 is responding to a Large Break Loss of Coolant Accident in accordance with EOP-0.0A, Reactor Trip or Safety Injection.

While verifying Containment Isolation Phase A, the Balance of Plant Operator determines that automatic and manual actuation of Containment Isolation Phase A will NOT function.

Which of the following valves must be closed as part of Containment Isolation Phase A?

A. 1/1-8153, XS LTDN ISOL VLV.

1/1-8154, XS LTDN ISOL VLV.

B. 1/1-LCV-0459, U1 LTDN ISOL VLV 0459.

1/1-LCV-0460, U1 LTDN ISOL VLV 0460.

C. 1-HV-6082, CH WTR RET ISOL VLV.

1-HV-6084, CH WTR SPLY ISOL VLV.

D. 1-HV-4650, VENT CHLR CCW SPLY & RET VLV.

1-HV-4631, PSC CCW SPLY & RET VLV.

Proposed Answer: C Explanation:

A. Incorrect. Plausible because 1/1-8152 & 1/1-8160, LTDN ISOL VLVs are part of a Phase A Containment Isolation Signal, however, the Excess Letdown Isolation Valves are located inside Containment and are isolated via the Seal Water Return lines.

B. Incorrect. Plausible because these valves will close, however, the actuation signal is a Safety Injection slave relay.

C. Correct. The Chill Water Return and Supply Isolation Valves are isolated on a Phase A Containment Isolation Signal.

D. Incorrect. Plausible because these valves do receive a close signal, however, it is generated from a Safety Injection Actuation Signal.

Page 36 of 52 CPNPP 2014 NRC RO Written Exam Worksheet 41 to 50

ES-401 RO Written Exam Worksheet Form ES-401-5 Technical Reference(s) EOP-0.0A, Attachments 2, 4, 7, & 10 Attached w/ Revision: See Comments / Reference Proposed references to be provided during examination: None Learning Objective: EXPLAIN the instrumentation and controls of the Containment System and PREDICT the system response.

Question Source: Bank ILOT7344 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 9 55.43 Comments /

Reference:

EOP-0.0A, Attachment 2 Revision: 8 Page 37 of 52 CPNPP 2014 NRC RO Written Exam Worksheet 41 to 50

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

EOP-0.0A, Attachment 2 Revision: 8 Page 38 of 52 CPNPP 2014 NRC RO Written Exam Worksheet 41 to 50

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

EOP-0.0A, Attachment 4 Revision: 8 Page 39 of 52 CPNPP 2014 NRC RO Written Exam Worksheet 41 to 50

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

EOP-0.0A, Attachment 7 Revision: 8 Page 40 of 52 CPNPP 2014 NRC RO Written Exam Worksheet 41 to 50

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

EOP-0.0A, Attachment 10 Revision: 8 Page 41 of 52 CPNPP 2014 NRC RO Written Exam Worksheet 41 to 50

ES-401 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. Date: 5/13/2014 Tier 2 Change: 4 Group 1 K/A 003 A1.02 Level of Difficulty: 2 Importance Rating 2.9 Reactor Coolant Pump System: Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the RCPS controls including: RCP pump and motor bearing temperatures Proposed Question: 49 Given the following conditions on Reactor Coolant Pump (RCP) 1-02:

Lower seal water bearing (pump radial) temperature is 221ºF.

Shaft vibration is 14 mils and steady.

Motor bearing temperature is 207ºF.

Number one (#1) seal water leakoff temperature is 215ºF.

RCP 1-02 must be stopped in accordance with ABN-101, Reactor Coolant Pump Trip/Malfunction due to high...

A. ...lower seal water bearing temperature.

B. ...pump shaft vibration.

C. ...motor bearing temperature.

D. ...seal water leakoff temperature.

Proposed Answer: C Explanation:

A. Incorrect. Plausible because there is a temperature limit on the lower seal water bearing, however, that value is 225°F.

B. Incorrect. Plausible because there is a limit on pump shaft vibration, however, that value is less than 20 mils and between 15 mils and 20 mils with increasing amplitude of 1 mil per hour.

C. Correct. With a motor bearing temperature greater than 195°F RCP 1-02 must be stopped.

D. Incorrect. Plausible because seal leak off temperature is a monitored parameter for a loss of seal injection and/or thermal barrier cooling water, however, only abnormal seal leak off flow requires an RCP to be stopped.

Technical Reference(s) ABN-101, Steps 3.3.2, 6.3.1, & Section 9.1 Attached w/ Revision: See ABN-101, Attachment 1 Comments / Reference Proposed references to be provided during examination: None Page 42 of 52 CPNPP 2014 NRC RO Written Exam Worksheet 41 to 50

ES-401 RO Written Exam Worksheet Form ES-401-5 Learning Objective: COMPREHEND the normal, abnormal and emergency operation of the Reactor Coolant System.

Question Source: Bank ILOT2044 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 3 55.43 Comments /

Reference:

ABN-101, Step 3.3.2 Revision: 10 Page 43 of 52 CPNPP 2014 NRC RO Written Exam Worksheet 41 to 50

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

ABN-101, Step 6.3.1 Revision: 10 Page 44 of 52 CPNPP 2014 NRC RO Written Exam Worksheet 41 to 50

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

ABN-101, Section 9.1 Revision: 10 Page 45 of 52 CPNPP 2014 NRC RO Written Exam Worksheet 41 to 50

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

ABN-101, Attachment 1 Revision: 10 Page 46 of 52 CPNPP 2014 NRC RO Written Exam Worksheet 41 to 50

ES-401 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. Date: 5/18/2014 Tier 2 Change: 3 Group 1 K/A 004 A3.15 Level of Difficulty: 3 Importance Rating 3.5 Chemical and Volume Control System: Ability to monitor automatic operation of the CVCS, including: PZR pressure and temperature Proposed Question: 50 Given the following conditions:

The Unit 1 Reactor Coolant System is currently at 125ºF with the Pressurizer solid.

Reactor Coolant System (RCS) pressure is 340 psig.

Train A Residual Heat Removal (RHR) is in service, with Letdown flow via HCV-0128, RHR LTDN FLO CTRL, (fully open) and PCV-131, LTDN HX OUT PRESS CTRL (in AUTO).

RHR discharge pressure is 510 psig.

If RHR Pump 1-01 were to trip, what would be the expected RCS pressure response?

Reactor Coolant System pressure would go to A. PCV-131 AUTO setpoint.

B. RHR pump suction relief setpoint.

C. RHR pump discharge relief setpoint.

D. PORV LTOP setpoint.

Proposed Answer: D Page 47 of 52 CPNPP 2014 NRC RO Written Exam Worksheet 41 to 50

ES-401 RO Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible if thought that the automatic setpoint on PCV-131 is lower than RCS pressure, however, the setpoint is RCS pressure plus RHR pump differential pressure.

B. Incorrect. Plausible because there are relief valves on both the suction and discharge side of the RHR pump, however, the suction side relief lifts at 450 psig while the discharge side relief setpoint is 600 psig. With anticipated pressure increase between 100 and 150 psig and current RCS pressure at 340 psig one could anticipate lifting of the suction side but not the discharge side relief.

C. Incorrect. Plausible because there are relief valves on both the suction and discharge side of the RHR pump, however, the suction side relief lifts at 450 psig while the discharge side relief setpoint is 600 psig. With anticipated pressure increase between 100 and 150 psig and current RCS pressure at 340 psig one could anticipate lifting of the suction side but not the discharge side relief.

D. Correct. With RCS temperature at 125°F and RCS pressure at 340 psig, tripping of the RHR Pump would cause an expected rise in RCS pressure between 100 and 150 psig. This would result in lifting of the PORV since the LTOP setpoint when less than 150°F is 375 psig per the TDM.

Technical Reference(s) IPO-005A, Section 3.3 Attached w/ Revision: See TDM-301A, PORV LTOP Setpoints Comments / Reference LO21.SYS.RH1, Pages 21 & 22 LO21.SYS.RH1, Figure Proposed references to be provided during examination: None Learning Objective: DISCUSS the actions for placing the plant in a solid condition in accordance with IPO-005, Plant Cooldown from Hot Standby to Cold Shutdown.

Question Source: Bank ILOT5914 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 Page 48 of 52 CPNPP 2014 NRC RO Written Exam Worksheet 41 to 50

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

IPO-005A, Section 3.3 Revision: 25 Page 49 of 52 CPNPP 2014 NRC RO Written Exam Worksheet 41 to 50

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

TDM-301A, PORV LTOP Setpoints Revision: 10 Page 50 of 52 CPNPP 2014 NRC RO Written Exam Worksheet 41 to 50

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

LO21.SYS.RH1, Page 21 Revision: 10/20/11 RHR PUMP SUCTION RELIEF VALVES (U-8708A&B)

The RHR System is designed for 600psig, thus when the RCS Hot Leg Recirculation Isolation Valves are open, the RHR System requires overpressure protection. The RHR Pump Suction Relief Valves have a design capacity of 900 gpm and discharge to the Pressurizer Relief Tank.

The relief setpoint is 450 psig and relief capacity requirements are calculated based on 2 analyzed situations:

Comments /

Reference:

LO21.SYS.RH1, Page 22 Revision: 10/20/11 RHR COLD LEG INJECTION RELIEF VALVE (U-8856A&B)

The RHR Cold Leg Injection Relief Valves provide overpressure protection from the effects of slow thermal expansion of fluid trapped within the lines they protect. This condition could result from stopping flow during LOCA conditions when the RHR system would contain radioactive materials that would generate heat as they decay. The leg of piping bounded by the discharge check valve (downstream of the Flow Control Valve), u-8716A/B and u-8809A/B would be trapped, and as the temperature increases, the pressure would increase. This pressure increase could rupture piping if not relieved.

The relief valves also protect against the leakage of the Reactor Coolant System water past the check valves located upstream of the RHR System injection point. These valves are set to open when pressure reaches 600psig and have the capacity to relieve 20 gpm to the Recycle Holdup Tank.

Page 51 of 52 CPNPP 2014 NRC RO Written Exam Worksheet 41 to 50

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

LO21.SYS.RH1, Figure Revision: 10/20/11 Page 52 of 52 CPNPP 2014 NRC RO Written Exam Worksheet 41 to 50

ES-401 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. Date: 5/16/2014 Tier 2 Change: 3 Group 1 K/A 012 A2.02 Level of Difficulty: 3 Importance Rating 3.6 Reactor Protection System: Ability to (a) predict the impacts of the following malfunctions or operations on the RPS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Loss of instrument power Proposed Question: 51 Given the following conditions on Unit 1:

Solid State Protection System (SSPS) Train A is in the Slave Relay Testing alignment.

One of the 48 VDC power supplies to the Train B SSPS just failed.

Which of the following identifies the impact on the Reactor Protection System and what action should be taken to mitigate the situation?

A. A Reactor trip should occur based on a General Warning Condition in both SSPS trains.

Ensure a Reactor Trip occurs and enter EOP-0.0A, Reactor Trip or Safety Injection.

B. Train B SSPS continues to have 48 VDC power from its other 48 VDC power supply.

Complete testing in the Train A SSPS while repairing the Train B power supply.

C. A General Warning Alarm for Train B SSPS will annunciate and Train B will be inoperable.

Stop all testing in the Train A SSPS until repairs are completed on the failed Train B SSPS power supply.

D. A single channel trip will be initiated in each trip function on Train B SSPS due to the power loss.

Stop all testing, have I&C BYPASS the individual channels on Train B SSPS and TRIP the bypassed channels if not corrected within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

Proposed Answer: A Page 1 of 44 CPNPP 2014 NRC RO Written Exam Worksheet 51 to 60

ES-401 RO Written Exam Worksheet Form ES-401-5 Explanation:

A. Correct. The Train A alignment causes the Train A Solid State Protection System to be in a degraded condition with a General Warning Alarm. If both trains have a General Warning Condition a Reactor trip is generated and the actions would be to ensure the trip occurs and to perform EOP-0.0A Reactor Trip or Safety Injection actions.

B. Incorrect. Plausible because it could be thought that this was the reason for having two 48 VDC power supplies and would have no effect which would make the action appropriate, however, the loss of the power supply creates a second General Warning Alarm condition resulting in a Reactor Trip.

C. Incorrect. Plausible because it could be thought that Train A was not in a degraded condition and though this could cause a General Warning there would be no trip signal generated.

D. Incorrect. Plausible because it could be thought that this would just cause individual trips on each trip parameter on that channel and as long as another channel was not in trip then there would be no other affect, however, a General Warning Alarm for Train B will occur which will complete the logic to trip the Reactor.

Technical Reference(s) ALM-0064A, 1-ALB-6D, Window 2.5 Attached w/ Revision: See SOP-711A, Step 5.4.2 NOTE & CAUTION Comments / Reference Proposed references to be provided during examination: None Learning Objective: EXPLAIN the instrumentation and controls of the Solid State Protection System and PREDICT the system response.

Question Source: Bank ILOT8216 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 Page 2 of 44 CPNPP 2014 NRC RO Written Exam Worksheet 51 to 60

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

ALM-0064A, 1-ALB-6D, Window 2.5 Revision: 6 Page 3 of 44 CPNPP 2014 NRC RO Written Exam Worksheet 51 to 60

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

SOP-711A, Step 5.4.2 NOTE & CAUTION Revision: 9 Page 4 of 44 CPNPP 2014 NRC RO Written Exam Worksheet 51 to 60

ES-401 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. Date: 5/22/2014 Tier 2 Change: 5 Group 1 K/A 008 K1.04 Level of Difficulty: 3 Importance Rating 3.3 Component Cooling Water System (CCWS): Knowledge of the physical connections and/or cause-effect relationships between the CCWS and the following systems: RCS, in order to determine source(s) of RCS leakage into the CCWS Proposed Question: 52 Given the following conditions:

Unit 1 is in a normal Middle-of-life (MOL) full power alignment.

Unit 2 is in MODE 3.

Unit 1 Component Cooling Water (CCW) surge tank level is 80% and rising.

1RE-4510 (CCW168), UNIT 1 COMPONENT COOLING WATER NON-SFGD RADIATION DETECTOR, goes into ALERT.

1-FI-132, LTDN FLO is 132 gpm.

Unit 1 Reactor Coolant System Average Temperature (Tave) is 585.4ºF.

Unit 1Turbine load is 1265 MWe.

Unit 1 Reactor Coolant Pump seal leakoff is 2.5 gpm per pump.

Which of the following Unit 1/Common components is the source of the CCW surge tank level increase?

A. RCP Thermal Barrier Heat Exchanger.

B. Letdown Heat Exchanger.

C. Seal Water Return Heat Exchanger.

D. Spent Fuel Pool Heat Exchanger X-02.

Proposed Answer: A Page 5 of 44 CPNPP 2014 NRC RO Written Exam Worksheet 51 to 60

ES-401 RO Written Exam Worksheet Form ES-401-5 Explanation:

A. Correct. With leakage from a RCP thermal barrier heat exchanger to the CCW system, indications would include radiation monitor alarms in the non-safeguards CCW loop and rising CCW surge tank level. As leakage would be out of the RCS automatic makeup would replace the leakage flow with minimal impact on RCS parameters as long as the leak remained small.

B. Incorrect. Plausible because the CCW non-safeguards loop cools the CVCS via the Letdown Heat Exchanger which would leak into the CCW system if a leak were to develop in the Letdown Heat Exchanger. However, as letdown flow is normal, no leakage out of the Letdown Heat Exchanger into CCW is indicated.

C. Incorrect. Plausible because the CCW non-safeguards loop cools the CVCS via the Seal Water Return Heat Exchanger. The CCW system would leak into CVCS if a leak were to develop in the Seal Water Return Heat Exchanger. However, as RCS temperature is normal, no leakage into CVCS from CCW is indicated as this would appear as a dilution event and RCS temperature would be increasing.

D. Incorrect. Plausible because the Unit 1 CCW non-safeguards loop can be aligned to cool Spent Fuel Pool Heat Exchanger X-02, however during normal alignments Unit 1 CCW cools the X-01 Heat Exchanger and Unit 2 cools the X-02 Heat Exchanger.

Technical Reference(s) LO21.SYS.CC1 Attached w/ Revision: See ABN-502 Comments / Reference Proposed references to be provided during examination: None Learning Objective: Given a procedural step, or sequence of steps from ECA-2.1, Uncontrolled Depressurization of All Steam Generators, STATE the purpose/basis for the step(s).

Question Source: Bank Modified Bank ILOT5655 (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 Page 6 of 44 CPNPP 2014 NRC RO Written Exam Worksheet 51 to 60

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

LO21.SYS.CC1 Revision: 5/01/2011 Page 7 of 44 CPNPP 2014 NRC RO Written Exam Worksheet 51 to 60

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

ABN-502 , Step 4 Revision: 6 Comments /

Reference:

ABN-502, Step 2b Revision: 6 Page 8 of 44 CPNPP 2014 NRC RO Written Exam Worksheet 51 to 60

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

ABN-502, Step 4 Revision: 6 Page 9 of 44 CPNPP 2014 NRC RO Written Exam Worksheet 51 to 60

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

ABN-502, Step 5a Revision: 6 Page 10 of 44 CPNPP 2014 NRC RO Written Exam Worksheet 51 to 60

ES-401 RO Written Exam Worksheet Form ES-401-5 Original Question: ILOT5655 Given the following conditions:

Unit 1 is in a normal full power lineup.

1RE-4510 (CCW168), Unit 1 COMPONENT COOLING WATER NON-SFGD RADIATION DETECTOR, goes into ALERT.

The CCW surge tank level is observed to be increasing slowly.

Which of the following components could be the source of this inleakage?

A. Letdown heat exchanger B. Excess letdown heat exchanger C. Seal water heat exchanger D. Boron Recycle Evaporator vent condenser Answer: A Page 11 of 44 CPNPP 2014 NRC RO Written Exam Worksheet 51 to 60

ES-401 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. Date: 3/3/14 Tier 2 Change: 2 Group 1 K/A 064 G 2.4.34 Level of Difficulty: 2 Importance Rating 4.2 Emergency Diesel Generator System: Emergency Procedures/Plan: Knowledge of RO tasks performed outside the main control room during an emergency and the resultant operational effects Proposed Question: 53 Given the following conditions on Unit 1:

The Unit 1 Control Room was evacuated due to smoke entering the area.

ABN-905A, Loss of Control Room Habitability, Attachment 11, Local Start of Diesel Generators is being performed.

A procedure CAUTION in ABN-905A warns against running the Auxiliary Oil Pump for longer than one minute in the HAND Mode.

Which of the following describes the potential impact on operations that the CAUTION is trying to prevent?

A. Extended operation in hand could result in the Discharge Oil Pressure Regulators sticking in the diverted position and insufficient oil may be supplied to the Diesel Generator on start.

B. Extended operation in hand could result in the Engine Lube Oil Pump Suction Relief sticking open and result in insufficient oil being supplied to the Diesel Generator on start.

C. The Lube Oil Strainer is not rated for extended periods of operation with the higher differential pressure when the oil is being diverted to the sump and damage to the strainer internals could occur.

D. Extended operation in hand could result in flooding the Turbocharger with oil and resultant ignition of the oil when the Diesel Generator is started which will damage the Turbocharger Bearings.

Proposed Answer: D Page 12 of 44 CPNPP 2014 NRC RO Written Exam Worksheet 51 to 60

ES-401 RO Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible because problems have occurred with the oil pressure regulators sticking, however, not due to this operation. The concern is flooding the Turbocharger with oil and then igniting the oil when the Diesel starts causing exhaust pressure surges that damage the Turbocharger Bearings.

B. Incorrect. Plausible because this is a possible failure mode when the engine is operated in reverse, however, this is not the concern in this case.

C. Incorrect. Plausible because this is a possible failure mode, however, this is not the concern in this case.

D. Correct. The concern is flooding the Turbocharger with oil and then igniting the oil when the Diesel starts causing exhaust pressure surges that damage the Turbocharger Bearings.

Technical Reference(s) ABN-905A, Attachment 11 Attached w/ Revision: See LO21.SYS.ED1, Pages 41 & 44 Comments / Reference Proposed references to be provided during examination: None Learning Objective: ANALYZE the response to a Loss Of Control Room Habitability in accordance with ABN-905, Loss Of Control Room Habitability.

Question Source: Bank ILOT8247 Modified Bank (Note changes or attach parent)

New Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Page 13 of 44 CPNPP 2014 NRC RO Written Exam Worksheet 51 to 60

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

ABN-905A, Attachment 11 Revision: 9 Comments /

Reference:

LO21.SYS.ED1, Page 41 Revision: 05/02/11 Diesel Generator Engine Lube Oil Pump Suction Relief Valve uDO-0155 (uDO-0255) is located by the lube oil sump tank and lifts at > 70 psig to relieve excess pressure back to the lube oil sump. The relief is provided to protect against a sudden pressure buildup in the suction in the event of a reverse rotation of the diesel engine. During engine operation, the lube oil pressure will be approximately 55 to 60 psig.

Page 14 of 44 CPNPP 2014 NRC RO Written Exam Worksheet 51 to 60

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

LO21.SYS.ED1, Page 44 Revision: 05/02/11 Diesel Generator Lube Oil Duplex Filter Differential Pressure Indicator u-PI-3411-2B (u-PI-3412-2B) on the engine control panel (0 to 60 psid) is used to evaluate the condition of the in-service filter. The maximum operating limit is 20 psid across a vessel.

Diesel Generator Lube Oil Duplex Filter Inlet And Outlet Pressure Indicators u-PI-3411-2H and u-PI-3411-2I (u-PI-3412-2H and u-PI-3412-2I), respectively, provide local indication (0 to 160 psig) at a gageboard on the engine skid.

Each filter has two pressure sensing taps, coming directly off of the vessels, for indication of individual vessel inlet and outlet pressure (0 to 160 psig). These pressure instruments, u-PI-3411-2D, -2E, -2F and -

2G (u-PI-3412-2D, -2E, -2F and -2G) are mounted on a gageboard on the engine skid.

Diesel Generator Lube Oil Filter Differential Pressure Switch u-PS-3411-2A (u-PS-3412-2A) senses the differential pressure across the duplex filters and causes the HIGH P LUBE FILTER alarm on the engine control panel if it exceeds 20 psid (PS-17C).

Page 15 of 44 CPNPP 2014 NRC RO Written Exam Worksheet 51 to 60

ES-401 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. Date: 3/3/14 Tier 2 Change: 2 Group 1 K/A 078 A3.01 Level of Difficulty: 3 Importance Rating 3.1 Instrument Air System: Ability to monitor automatic operation of the IAS, including: Air pressure Proposed Question: 54 Given the following conditions:

Instrument Air (IA) Compressor 1-01 is operating as the LEAD compressor.

IA Compressor 1-02 is in an AUTO-START condition as the BACKUP compressor.

IA Compressor X-01 is in STANDBY and aligned to Unit 1 through Air Dryer X-01.

The following sequence of events occur:

At 1415, 1-ALB-01, Window 2.4 - CNTMT INSTR AIR HDR PRESS LO, alarms as pressure drops to 84 psig.

At 1416, 1-ALB-01, Window 3.3 - INSTR AIR HDR PRESS LO, alarms as pressure drops to 85 psig.

All other Unit 1 Control Room alarms related to the IA System remain clear.

At 1420, a stuck-open relief valve on Air Dryer 1-01 reseats.

At 1422, both Instrument Air alarms (1-ALB-01, Windows - 2.4 and 3.3) clear.

At 1423, Instrument Air header pressure is 93 psig and slowly rising.

At 1423, assuming NO additional operator actions and with IA Compressor 1-01 running and loaded, which of the following is the status of IA Compressors 1-02 and X-01 in accordance with SOP-509A, Instrument Air System?

IA Compressor 1-02 is __________ and IA Compressor X-01 is __________.

A. running and loaded; running and loaded B. running and loaded; shutdown C. running and unloaded; running and unloaded D. running and unloaded; shutdown Proposed Answer: A Page 16 of 44 CPNPP 2014 NRC RO Written Exam Worksheet 51 to 60

ES-401 RO Written Exam Worksheet Form ES-401-5 Explanation:

A. Correct. Given the conditions listed, the BACKUP and STANDBY IA Compressors will both be running and loaded. Air pressure must reach 115 psig to place the compressors in an unloaded condition and then they will run for 20 minutes then shutdown to an Auto-Start condition.

B. Incorrect. Plausible because the LEAD compressor is running and loaded and given that the low pressure alarms are clear it could be thought that the STANDBY compressor would shutdown, however air pressure must reach 115 psig for both compressors to unload and then they must run unloaded for 20 minutes before shutting down.

C. Incorrect. Plausible because the BACKUP and STANDBY IA Compressors will be running, however, they will also be loaded until air pressure reaches 115 psig at which point they would unload.

D. Incorrect. Plausible because it could be thought that the BACKUP compressor will unload when the low pressure alarms clear and that the STANDBY compressor will shut down when the low pressure alarms clear.

Technical Reference(s) SOP-509A, Step 5.2.1.J NOTE Attached w/ Revision: See ALM-0011A, 1-ALB-01, Windows 2.4 & 3.3 Comments / Reference SOP-509A, Step 5.4.1.H NOTE Proposed references to be provided during examination: None Learning Objective: COMPREHEND the normal, abnormal and emergency operation of the Instrument Air System.

Question Source: Bank ILOT8178 Modified Bank (Note changes or attach parent)

New Question History: Last NRC Exam 2012 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 Page 17 of 44 CPNPP 2014 NRC RO Written Exam Worksheet 51 to 60

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

SOP-509A, Step 5.2.1.J NOTE Revision: 22 Page 18 of 44 CPNPP 2014 NRC RO Written Exam Worksheet 51 to 60

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

SOP-509A, Step 5.4.1.H NOTE Revision: 22 Page 19 of 44 CPNPP 2014 NRC RO Written Exam Worksheet 51 to 60

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

ALM-0011A, 1-ALB-01, Windows 2.4 & 3.3 Revision: 10 Page 20 of 44 CPNPP 2014 NRC RO Written Exam Worksheet 51 to 60

ES-401 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. Date: 5/16/2014 Tier 2 Change: 3 Group 1 K/A 006 K4.16 Level of Difficulty: 2 Importance Rating 3.2 Emergency Core Cooling System: Knowledge of the ECCS design feature(s) and/or interlock(s) that provide for the following:

Interlocks between RHR valves and RCS Proposed Question: 55 Which of the following should PREVENT the Reactor Operator from manually opening 1/1-8701A, RHRP1 HL RECIRC ISOL VLV?

A. 1/1-8804A, RHRP 1 TO CCP SUCT VLV, is open.

B. 1/1-8809A, RHR TO CL 1&2 INJ ISOL VLV, is open.

C. 1/1-8811A, CNTMT SMP TO RHRP 1 SUCT ISOL VLV, is closed.

D. 1/1-8812A, RWST TO RHRP 1 SUCT VLV, is closed.

Proposed Answer: A Explanation:

A. Correct. 1/1-8701A is interlocked with 1/1-8804A and prevents 1/1-8701A from opening when 1/1-8804A is in the OPEN position. These interlocks prevent a suction source from the RCS hot leg to the RHR pump from interfering with other RHR pump alignments.

B. Incorrect. Plausible if thought that 1/1-8701A is interlocked with 1/1-8809A, however, there are no interlocks associated with these valves and they remain open in all modes of plant operation with the exception of ECCS hot leg injection.

C. Incorrect. Plausible because 1/1-8701A is interlocked with 1/1-8811A, however, 1/1-8811A must be in the OPEN vice CLOSED position to prevent 1/1-8701A from opening.

D. Incorrect. Plausible because 1/1-8701A is interlocked with 1/1-8812A, however, 1/1-8812A must be in the OPEN vice CLOSED position to prevent 1/1-8701A from opening.

Technical Reference(s) LO21.SYS.RH1, Pages 15 & 18 Attached w/ Revision: See Comments / Reference Proposed references to be provided during examination: None Learning Objective: DEMONSTRATE an understanding of the components of the Residual Heat Removal system including interrelations with other systems to include interlocks and control loops.

Question Source: Bank ILOT7140 Modified Bank (Note changes or attach parent)

New Page 21 of 44 CPNPP 2014 NRC RO Written Exam Worksheet 51 to 60

ES-401 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 55.43 Comments /

Reference:

LO21.SYS.RH1, Page 15 Revision: 10/20/11 In order for the Reactor Operator to open the Train A RHR Pump Hot Leg Recirculation Isolation Valves (u-8701A and u-8702A), the following interlocks must be met:

the Containment Sump to RHR Pump Suction Isolation Valve (u-8811A) must be CLOSED, and the RWST to RHR Pump Suction Valve (u-8812A) must be CLOSED, and the RHR Pump to CCP/SIP Suction Valve (u-8804A) must be CLOSED, and detected Reactor Coolant System pressure from pressure transmitter PT-405 (Train A) must be less than 364 psig, and the valve handswitch on CB-04 placed in its OPEN position.

The interlocks for the Train B valves utilize their train B counterpart. PT-403 would be used instead of PT-405. The above interlocks are bypassed when control is transferred to the Remote Shutdown Panel for u-8701A and u-8701B.

Comments /

Reference:

LO21.SYS.RH1, Page 18 Revision: 10/20/11 RHR TO COLD LEG INJECTION ISOLATION VALVE (U-8809A&B)

The RHR to Cold Leg Injection Isolation Valves provide a means to isolate the discharge of the RHR System from the Reactor Coolant System. These valves remain open in all modes of plant operation with the exception of ECCS Hot Leg Injection Mode.

Page 22 of 44 CPNPP 2014 NRC RO Written Exam Worksheet 51 to 60

ES-401 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. Date: 5/16/14 Tier 2 Change: 3 Group 2 K/A 001 K2.01 Level of Difficulty: 2 Importance Rating 3.5 Control Rod Drive System: Knowledge of bus power supplies to the following: One-line diagram of power supply to MG sets Proposed Question: 56 One method to remove power from the control rods during an Anticipated Transient without Trip is to remove power from the Control Rod Drive Motor Generators.

The power flow path to the Unit 1 Control Rod Drive Motor Generators 1-01/1-02 is from 480V buses...

A. ...1B1/1B2 - Motor Breakers - Motor Generators 1-01/1-02.

B. ...1B2/1B3 - Motor Breakers - Motor Generators 1-01/1-02.

C. ...1B3/1B4 - Motor Breakers - Motor Generators 1-01/1-02.

D. ...1B4/1B1 - Motor Breakers - Motor Generators 1-01/1-02.

Proposed Answer: C Explanation:

A. Incorrect. Plausible if a misconception of which non safety 480 VAC Buses are the power supplies as the remaining portion of the one-line diagram is correct.

B. Incorrect. Plausible if a misconception of which non safety 480 VAC Buses are the power supplies as the remaining portion of the one-line diagram is correct.

C. Correct. The electrical one-line diagram to the Unit 1 CRD MGs is 1B3/1B4 - Motor Breaker -

Motor Generators.

D. Incorrect. Plausible if a misconception of which non safety 480 VAC Buses are the power supplies as the remaining portion of the one-line diagram is correct.

Technical Reference(s) LO21.SYS.CR1, Page 14 Attached w/ Revision: See LO21.SYS.CR1, Figure 5 Comments / Reference Proposed references to be provided during examination: None Learning Objective: COMPREHEND the normal, abnormal and emergency operation of the Rod Control System.

Question Source: Bank ILOT6013 Modified Bank (Note changes or attach parent)

New Page 23 of 44 CPNPP 2014 NRC RO Written Exam Worksheet 51 to 60

ES-401 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 6 55.43 Comments /

Reference:

LO21.SYS.CR1, Page 14 Revision: 05/02/11 ROD CONTROL GENERAL DESCRIPTION The Rod Control System is housed in seven cabinets (located 832 safeguard building) with main power supplied through two paralleled, full capacity, rod drive motor generators (M-Gs). Power is fed from uB3 and uB4 (480 vac) to the M-Gs.

Comments /

Reference:

LO21.SYS.CR1, Figure 5 Revision: 05/02/11 Page 24 of 44 CPNPP 2014 NRC RO Written Exam Worksheet 51 to 60

ES-401 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. Date: 5/16/14 Tier 2 Change: 3 Group 2 K/A 002 K4.05 Level of Difficulty: 3 Importance Rating 3.8 Reactor Coolant System: Knowledge of RCS design feature(s) and/or interlock(s) that provide for the following: Detection of RCS leakage Proposed Question: 57 Which of the following design features is used to assist in identifying the source of Reactor Coolant System leakage from the Reactor Vessel Head when the Reactor Vessel Flange Leakoff Temperature High alarms in the Control Room?

Flow into the A. Reactor Coolant Drain Tank.

B. Containment Sump.

C. Pressurizer Relief Tank.

D. Volume Control Tank.

Proposed Answer: A Explanation:

A. Correct. When conditions permit, Containment entry is made and Reactor Vessel Flange Leakoff is directed to the Reactor Coolant Drain Tank (RCDT).

B. Incorrect. Plausible because the Reactor Coolant Pump number 3 seal is directed to the Containment Sump while the number 2 seal is directed to the RCDT.

C. Incorrect. Plausible because numerous components drain to the PRT such as RHR suction reliefs and CVCS Letdown and seal return reliefs, however, Reactor Vessel Flange Leakoff is directed to the RCDT.

D. Incorrect. Plausible because the Reactor Coolant Pump Number 1 Seal is directed to the Volume Control Tank, however, the Number 2 Seal is directed to the RCDT and the Number 3 Seal is directed to the Containment Sump.

Technical Reference(s) ALM-0053A, 1-ALB-5C, Window 1.1 Attached w/ Revision: See LO21.SYS.RC1, Pages 11 & 14 Comments / Reference LO21.SYS.RC1, Figure 17 LO21.SYS.CS1, Pages 9 & 10 Proposed references to be provided during examination: None Page 25 of 44 CPNPP 2014 NRC RO Written Exam Worksheet 51 to 60

ES-401 RO Written Exam Worksheet Form ES-401-5 Learning Objective: EXPLAIN the instrumentation and controls of the Reactor Vessel, Internals and Core Components System and PREDICT the system response.

ANALYZE the response to Excessive Reactor Coolant Leakage in accordance with ABN-103, Excessive Reactor Coolant Leakage.

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

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 3 55.43 Page 26 of 44 CPNPP 2014 NRC RO Written Exam Worksheet 51 to 60

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

ALM-0053A, 1-ALB-5C, Window 1.1 Revision: 7 Page 27 of 44 CPNPP 2014 NRC RO Written Exam Worksheet 51 to 60

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

LO21.SYS.RC1, Page 11 Revision: 04/28/11 Reactor Vessel Flange Seal For both Units, two self-energizing O-ring gaskets, constructed of silver plated Ni-Cr-Fe alloy, form the pressure boundary seal between the closure head and reactor vessel flanges. Each closure head flange has two machined-in 1/2 inch wide by 1/4-inch deep grooves and 32 screw taps in the closure head flange.

Sixteen (16) retainer clips and Allen head screws hold each O-ring in place. For O-ring leak detection, a tapped space outside the inner O-ring and a space outside the outer O-ring drain to one-inch piping connections on the lower reactor vessel flange. Piping from these drain connections extends through the missile barrier walls to allow operator interaction during plant operation.

Outside the missile barrier, each pipe from these connections reduces to 3/4-inch and contains a manual isolation valve (uRC-8069A, B). Manual isolation valves provide local isolation in case of leakage from the respective O-ring. A common line joins the inner and outer O-ring leakoff connections. This common line contains a normally closed, manual 3/4-inch tell-tale valve and a 3/4-inch to -inch reducer.

Valve u-8032 is downstream of the reducer, draining to the Reactor Coolant Drain Tank (RCDT). It is air-operated and manually controlled from Main Control Board CB05 to provide remote isolation capability for a leaking O-ring. This valve will fail open upon a loss of power or instrument air. A bottom-mounted, strap-on RTD (u-TE-0401) provides reactor vessel flange leak-off temperature indication at CB05.

During normal operation, inner O-ring isolation valve uRC-8069B is open. Should the inner O-ring leak, a high temperature alarm actuates at 140F on CB05, informing the operator of the leak. The operator monitors reactor vessel flange leak-off temperature and closes u-8032, isolating the leak. Procedure directs shutting manual isolation valve uRC-8069B and opening uRC-8069A, transferring RCS pressure boundary maintenance to the outer O-ring. Opening valve u-8032 then transfers leak detection to the outer O-ring.

Comments /

Reference:

LO21.SYS.RC1, Page 14 Revision: 04/28/11 Number two seal is a face rubbing seal consisting of a carbon insert shrunk into a stainless steel ring.

The carbon insert rubs against a rotating, chrome carbide coated, surface on a stainless steel runner.

Leakage from the number two seal joins with the outer dam leakage of the number three seal and drains to the reactor coolant drain tank.

Number three seal is a face rubbing seal consisting of a carbon insert, with two concentric sealing faces or dams, shrunk into a stainless steel seal ring. These dams rub against a rotating, chrome carbide coated, surface on a stainless steel runner. Clean water is injected between the dams on the seal ring at a pressure greater than the number two seal leak-off cavity. Part of the injected water flows past the outer dam where it joins the leakage from the number two seal and passes out of the pump, through the number two seal leak-off connection. The remainder of the injected water flows past the inner dam and drains through the number three seal leak-off piping to a floor drain sump inside containment.

Page 28 of 44 CPNPP 2014 NRC RO Written Exam Worksheet 51 to 60

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

LO21.SYS.RC1, Figure 17 Revision: 04/28/11 Page 29 of 44 CPNPP 2014 NRC RO Written Exam Worksheet 51 to 60

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

LO21.SYS.CS1, Pages 9 & 10 Revision: 04/28/11 through the #1 seal. The majority of the #1 seal leakoff flow (approximately 3 gpm per pump) is routed to a header with the seal leakoff from the other three reactor coolant pumps. The seal return flow, along with water from the excess letdown heat exchanger, is directed from the containment to the safeguards building, through the seal water return filter, the seal water heat exchanger, and then to the suction of the charging pumps. An alternate path can be aligned such that seal return flow out of the seal water heat exchanger is directed to the volume control tank through a spray nozzle.

A very small portion of the #1 seal leakoff on each reactor coolant pump (approximately 3 gallons per hour) leaks through the #2 seal and is routed to the reactor coolant drain tank in the containment building. A #3 seal provides a final barrier to prevent leakage of reactor coolant to the containment atmosphere. The #3 seal consists of two sealing faces called dams. Reactor makeup water from a standpipe is injected between the dams at a pressure greater than that of the #2 seal leakoff. Part of this clean injection water flows past the inner dam and is directed to the containment sump. The remainder of the injected water flows past the outer dam where it joins the leakage from the #2 seal and goes to the reactor coolant drain tank.

Page 30 of 44 CPNPP 2014 NRC RO Written Exam Worksheet 51 to 60

ES-401 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. Date: 5/16/2014 Tier 2 Change: 4 Group 2 K/A 011 K5.12 Level of Difficulty: 3 Importance Rating 2.7 Pressurizer Level Control System: Knowledge of the operational implications of the following concepts as they apply to the PZR LCS: Criteria and purpose of PZR level program Proposed Question: 58 Given the following conditions:

Unit 1 is being operated at 40% power with Pressurizer level at program.

RCS temperature is within 0.1ºF of TREF.

Assuming all control systems are maintained on program, which of the following describes how Pressurizer level will change as Reactor Power is increased from 40% to 60% and the purpose of the level change?

Level increases from approximately A. 25% to approximately 36% to maintain a relatively constant mass in the RCS.

B. 25% to approximately 36% to maintain a relatively constant volume in the RCS.

C. 39% to approximately 46% to maintain a relatively constant mass in the RCS.

D. 39% to approximately 46% to maintain a relatively constant volume in the RCS.

Proposed Answer: C Page 31 of 44 CPNPP 2014 NRC RO Written Exam Worksheet 51 to 60

ES-401 RO Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible if math error made.

B. Incorrect. Plausible if math error made.

C. Correct. Programmed level is between 25% and 60% for a Tave of no-load to full temperature, which is equivalent to 0% to 100% power. At 40% power, level should be 39% and at 60% power level should be 46%. The program is designed to allow a constant RCS mass as the RCS heats up and cools down.

D. Incorrect. Plausible since these are the correct values of Pressurizer level, but the program is designed to maintain a constant mass, not a constant volume.

Technical Reference(s) LO21.SYS.PP1, Page 5 Attached w/ Revision: See Comments / Reference Proposed references to be provided during examination: None Learning Objective: EXPLAIN the instrumentation and controls of the Pressurizer Pressure and Level Control System and PREDICT the system response.

Question Source: Bank ILOT6269 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:

LO21.SYS.PP1, Page 5 Revision: 05/05/11 Average Reactor Coolant System temperature (TAVG) increases from 557°F at 0% reactor power to 585.4 °F (589.2°F) at 100% reactor power. Pressurizer level is programmed to change as a function of the TAVG change. This allows the water in the RCS to expand as temperature increases from 0 - 100%

power, raising pressurizer level from 25% to 60% without having to drain water from the RCS. In the same manner, pressurizer level is allowed to decrease during power reduction as the RCS water cools without the need to add water to make up for the contraction. The RCS volume is allowed to change as a result of temperature changes, while the mass of the RCS water remains constant. This reduces transient response time and the amount of water required to be processed during normal operations.

Page 32 of 44 CPNPP 2014 NRC RO Written Exam Worksheet 51 to 60

ES-401 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. Date: 5/19/2014 Tier 2 Change: 4 Group 2 K/A 015 K6.02 Level of Difficulty: 2 Importance Rating 2.6 Nuclear Instrumentation System: Knowledge of the effect of a loss or malfunction of the following will have on the NIS:

Discriminator/compensation circuits Proposed Question: 59 Given the following conditions:

Unit 2 was manually tripped from 20% power.

Intermediate Range (IR) Channel 2-NI-35 indicates off-scale low.

IR Channel 2-NI-36 indicates 4 x 10-11 amps.

Source Range (SR) Channel 2-NI-31 indicates 1.1 x 104 cps.

SR Channel 2-NI-32 indicates 2.1 x 104 cps.

What are the effects exhibited on the Nuclear Instrumentation system response?

A. 2-NI-35 is over-compensated and the SR channels energized automatically.

B. 2-NI-35 is under-compensated and the SR channels required manual energization.

C. 2-NI-36 is over-compensated and the SR channels energized automatically.

D. 2-NI-36 is under-compensated and the SR channels required manual energization.

Proposed Answer: A Page 33 of 44 CPNPP 2014 NRC RO Written Exam Worksheet 51 to 60

ES-401 RO Written Exam Worksheet Form ES-401-5 Explanation:

A. Correct. With both SR channels indicating in the 104 cps decade, both IR channels should indicate in the 10-11 amps decade. As NI-35 is off-scale low the detector is over-compensated and thus reducing the signal to outside of the detectable range. As the IR channel would have still passed below 10-10 amps prior to the NI-36, the SR channels would have automatically reenergized.

B. Incorrect. Plausible because a misconception could exist as to the effect that compensation has on the IR channels and it could be believed that under-compensation lead to the off-scale low response of NI-35. As one channel was not working properly it could be believed that manual action to reenergize both SR channels is required.

C. Incorrect. Plausible because if unaware of the proper overlap between the SR and IR channels and suffering a misconception of the effect of compensation on the IR channels, it could be believed that NI-36 is ready erroneously instead of NI-35. As the IR channel would have still passed below 10-10 amps after NI-35, the SR channels would have automatically reenergized.

D. Incorrect. Plausible because if unaware of the proper overlap between the SR and IR channels and suffering a misconception of the effect of compensation on the IR channels, it could be believed that NI-36 is ready erroneously instead of NI-35. As one channel was not working properly it could be believed that manual action to reenergize both SR channels is required.

Technical Reference(s) EOS-0.1B Attached w/ Revision: See LO21.SYS.EC1 Comments / Reference ABN-701, Step 3.3 Proposed references to be provided during examination: None Learning Objective: EXPLAIN the instrumentation and controls of the Excore Instrumentation System and PREDICT the system response.

Question Source: Bank ILOT7128 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 Page 34 of 44 CPNPP 2014 NRC RO Written Exam Worksheet 51 to 60

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

EOS-0.1B Revision: 8 Page 35 of 44 CPNPP 2014 NRC RO Written Exam Worksheet 51 to 60

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

LO21.SYS.EC1 Revision: 5/01/2011 Comments /

Reference:

LO21.SYS.EC1 Revision: 5/01/2011 Page 36 of 44 CPNPP 2014 NRC RO Written Exam Worksheet 51 to 60

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

ABN-701, Step 3.3 Revision: 11 Page 37 of 44 CPNPP 2014 NRC RO Written Exam Worksheet 51 to 60

ES-401 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. Date: 5/17/2014 Tier 2 Change: 3 Group 2 K/A 016 K3.04 Level of Difficulty: 4 Importance Rating 2.6 Non-Nuclear Instrumentation System: Knowledge of the effect that a loss or malfunction of the NNIS will have on the following: MFW system Proposed Question: 60 Given the following conditions:

Unit 1 is at 60% power.

Instrument and Control (I&C) is performing a calibration on Steam Generator Narrow Range Level Channel, Loop 1, Protection Set II, CH 0519.

During restoration the I&C Technician places the channel in TRIP.

1-TSLB-3, Window 2.2 - SG 1 LVL HI-HI LB-519A is LIT.

1-TSLB-5, Window 2.4 - SG 1 LVL LO-LO LB-519B is LIT.

While Channel 519 is in trip a loss of 1PC1, 118 VAC INSTRUMENT DISTRIBUTION PANEL (CHAN I) 1PC1 occurs.

The Unit 1 Reactor will trip on A. Steam Generator Low-Low Level.

The Main Feedwater Pump turbines will NOT trip.

B. Steam Generator Low-Low Level.

The Main Feedwater Pump turbines will trip.

C. Turbine Trip.

The Main Feedwater Pump turbines will NOT trip.

D. Turbine Trip.

The Main Feedwater Pump turbines will trip.

Proposed Answer: A Page 38 of 44 CPNPP 2014 NRC RO Written Exam Worksheet 51 to 60

ES-401 RO Written Exam Worksheet Form ES-401-5 Explanation:

A. Correct. With TSLB-5 Window - 2.4 LIT, a loss of Panel 1PC1 will result in a 2 of 4 LVL LO-LO coincidence on one Steam Generator which will cause a Reactor Trip. A Main Feedwater Isolation will occur on the Reactor Trip (P-4) in conjunction with Low Tavg of 564ºF, however, the Main Feedwater Pump turbines will not trip as they would if a P-14 signal was assumed to cause the Main Feedwater Isolation.

B. Incorrect. Plausible because with TSLB-5 Window - 2.4 LIT a loss of Panel 1PC1 will result in a 2 of 4 LVL LO-LO coincidence on one Steam Generator which will cause a Reactor Trip. A Main Feedwater Isolation will occur on the Reactor Trip (P-4) in conjunction with Low Tavg of 564ºF, however, the Main Feedwater Pump turbines will not trip as they would if a P-14 signal was assumed to cause the Main Feedwater Isolation.

C. Incorrect. Plausible because one Steam Generator LVL HI-HI TSLB is LIT. A failure of Panel PC1 could be thought to meet the 2 of 3 LVL HI-HI coincidence on one Steam Generator which will cause a Turbine Trip and Main Feedwater Isolation. However, the LVL HI-HI is 2 of 3 and the channel fed by Panel PC1 is not one of the 3 channels as it is used for normal control function.

Thus the Turbine Trip will not be the initiating trip for the Reactor Trip, but the Turbine will trip on the Reactor Trip. A Main Feedwater Isolation will occur on the Reactor Trip (P-4) in conjunction with Low Tavg of 564ºF, however, the Main Feedwater Pump turbines will not trip as they would if a P-14 signal was assumed to cause the Main Feedwater Isolation.

D. Incorrect. Plausible because one Steam Generator LVL HI-HI TSLB is LIT. A failure of Panel PC1 could be thought to meet the 2 of 3 LVL HI-HI coincidence on one Steam Generator which will cause a Turbine Trip and Main Feedwater Isolation. However, the LVL HI-HI is 2 of 3 and the channel fed by Panel PC1 is not one of the 3 channels as it is used for normal control function.

Thus the Turbine Trip will not be the initiating trip for the Reactor Trip, but the Turbine will trip on the Reactor Trip. A Main Feedwater Isolation will occur on the Reactor Trip (P-4) in conjunction with Low Tavg of 564ºF, however, the Main Feedwater Pump turbines will not trip as they would if a P-14 signal was assumed to cause the Main Feedwater Isolation.

Technical Reference(s) INC-7296A, Section 9 Attached w/ Revision: See ABN-603, Attachment 1 Comments / Reference Technical Specification Table 3.3.1-1 Technical Specification Table 3.3.2-1 LO21.SYS.ES1 Proposed references to be provided during examination: None Learning Objective: COMPREHEND the normal, abnormal and emergency operation of the Main Feedwater 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 7 55.43 Page 39 of 44 CPNPP 2014 NRC RO Written Exam Worksheet 51 to 60

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

INC-7296A, Section 9 Revision: 7 Page 40 of 44 CPNPP 2014 NRC RO Written Exam Worksheet 51 to 60

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

ABN-603, Attachment 1 Revision: 8 Page 41 of 44 CPNPP 2014 NRC RO Written Exam Worksheet 51 to 60

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

Technical Specification Table 3.3.1-1, Item 14 Amendment: 161 Page 42 of 44 CPNPP 2014 NRC RO Written Exam Worksheet 51 to 60

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

Technical Specification Table 3.3.2-1, Item 5.b Amendment: 161 Page 43 of 44 CPNPP 2014 NRC RO Written Exam Worksheet 51 to 60

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

LO21.SYS.ES1, Pages 75 & 76 Revision: 5-04-2011 Page 44 of 44 CPNPP 2014 NRC RO Written Exam Worksheet 51 to 60

ES-401 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. Date: 3/3/2014 Tier 2 Change: 2 Group 2 K/A 033 G 2.4.31 Level of Difficulty: 2 Importance Rating 4.2 Spent Fuel Pool Cooling System: Emergency Procedures/Plan: Knowledge of annunciator alarms, indications, or response procedures Proposed Question: 61 Given the following conditions:

X-LS-4849A-1 and X-LS-4849A-2, Spent Fuel Pool level switches have both failed low.

Annunciator 1-ALB-6B, Window 4.4 - SFPCS TRBL is in alarm.

The Nuclear Equipment Operator reports that Window 1.1 - SFP CS PUMP 1 TRIP is in alarm on the Spent Fuel Pool Panel.

Which of the following identifies the effect on Spent Fuel Pool Cooling Water Pump X-01?

Spent Fuel Pool Cooling Water Pump X-01 A. ...can only be started at the Spent Fuel Pool Panel.

Once started, all pump interlocks are restored.

B. ...can only be started at the breaker.

Once started, all pump interlocks are restored.

C. ...can only be started at the Spent Fuel Pool Panel.

In this condition, the pump will NOT be load shed on a Safety Injection or Blackout Sequencer Signal.

D. ...can only be started at the breaker.

In this condition, the pump will NOT be load shed on a Safety Injection or Blackout Sequencer Signal.

Proposed Answer: D Explanation:

A. Incorrect. Plausible if thought the SFP Pump could be started with a low level condition.

B. Incorrect. Plausible because the SFP Pump can be started locally, however, in this condition all pump interlocks are bypassed.

C. Incorrect. Plausible because the SFP Pump will not load shed on a safety injection Signal or Blackout Sequencer Signal; however, with a low level condition the pump can only be started locally.

D. Correct. Given the conditions listed, operation of the SFP Pump is as stated.

Page 1 of 34 CPNPP 2014 NRC RO Written Exam Worksheet 61 to 70

ES-401 RO Written Exam Worksheet Form ES-401-5 Technical Reference(s) ABN-909, Step 4.3.1 Attached w/ Revision: See ALM-0701, Window 1.1 Comments / Reference Proposed references to be provided during examination: None Learning Objective: COMPREHEND the normal, abnormal and emergency operation of the Spent Fuel Pool Cooling and Cleanup System.

Question Source: Bank ILOT8409 Modified Bank (Note changes or attach parent)

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

Reference:

ABN-909, Step 4.3.1 Revision: 8 Page 2 of 34 CPNPP 2014 NRC RO Written Exam Worksheet 61 to 70

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

ALM-0701, Window 1.1 Revision: 5 Page 3 of 34 CPNPP 2014 NRC RO Written Exam Worksheet 61 to 70

ES-401 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. Date: 5/19/2014 Tier 2 Change: 3 Group 2 K/A 035 K1.12 Level of Difficulty: 2 Importance Rating 3.7 Steam Generator System: Knowledge of the physical connections and/or cause-effect relationships between the SGS and the following systems: RPS Proposed Question: 62 Given the following conditions:

Unit 2 is operating at 100% power.

2-LT-554, SG 4 LVL (NR) CHAN I fails to 100% coincident with the following annunciators:

2-ALB-8A, Window 4.8 - SG 4 STM & FW FLO MISMATCH.

2-ALB-8A, Window 4.12 - SG 4 LVL DEV.

Assuming NO operation action, which of the following identifies the Unit 2 actuation setpoint and expected plant response?

A. The Turbine will trip at 84% level in Steam Generator 2-04 causing a Reactor Trip.

B. The Turbine will trip at 81.5% level in Steam Generator 2-04 causing a Reactor Trip.

C. The Reactor will trip at 38% level in Steam Generator 2-04 causing a Turbine Trip.

D. The Reactor will trip at 35.4% level in Steam Generator 2-04 causing a Turbine Trip.

Proposed Answer: D Explanation:

A. Incorrect. Plausible because if the channel failed low the SG level would rise and the turbine would trip at 84% SG level on Unit 1 causing a reactor trip.

B. Incorrect. Plausible because if the channel failed low on Unit 2 the SG level would rise and the turbine would trip at 81.5% SG level causing a reactor trip.

C. Incorrect. Plausible because channel 554 is the controlling channel which when it fails high causes the feedwater control valve to close and at 38% level in Unit 1 SG 1-04 a reactor trip would be generated which would then cause a turbine trip.

D. Correct. Channel 554 is the controlling channel which when it fails high causes the feedwater control valve to close and at 35.4% level on a Unit 2 SG a reactor trip would be generated which would then cause a turbine trip.

Technical Reference(s) ABN-710, Sections 2.1 & 2.2 Attached w/ Revision: See Page 4 of 34 CPNPP 2014 NRC RO Written Exam Worksheet 61 to 70

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments / Reference Proposed references to be provided during examination: None Learning Objective: ANALYZE the response to a Steam Generator Level Instrument Malfunction in accordance with ABN-710 Steam Generator Level Instrument Malfunction.

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 5 of 34 CPNPP 2014 NRC RO Written Exam Worksheet 61 to 70

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

ABN-710, Section 2.1 Revision: 10 Page 6 of 34 CPNPP 2014 NRC RO Written Exam Worksheet 61 to 70

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

ABN-710, Sections 2.2 Revision: 10 Page 7 of 34 CPNPP 2014 NRC RO Written Exam Worksheet 61 to 70

ES-401 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. Date: 5/16/2014 Tier 2 Change: 4 Group 2 K/A 041 A1.02 Level of Difficulty: 3 Importance Rating 3.1 Steam Dump/Turbine Bypass Control System: Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the SDS controls including: Steam pressure Proposed Question: 63 Given the following conditions:

Unit 1 was operating at 20% power when a manual Reactor Trip was initiated for a planned shutdown.

1-PK-507, STM DMP PRESS CTRL potentiometer is at zero and in AUTO.

Which of the following statements describes the Steam Dump System response if handswitch 43/1-SD, STM DMP MODE SELECT is placed in the Steam Pressure Mode?

A. The HI-1 and HI-2 bistables will actuate opening all Steam Dump Valves.

When Steam Header pressure reaches 1057 psig, all Steam Dump Valves will be closed by the P-12 interlock.

B. The HI-1 and HI-2 bistables will actuate opening all Steam Dump Valves.

When Steam Header pressure reaches 1057 psig, all but three Steam Dump Valves will close and temperature will be controlled by the remaining three Steam Dump Valves in automatic.

C. A proportional error signal will open all Steam Dump Valves.

When Steam Header pressure reaches 1057 psig, all but three Steam Dump Valves will close and temperature will be controlled by the remaining three Steam Dump Valves in automatic.

D. A proportional error signal will open all Steam Dump Valves.

When Steam Header pressure reaches 1057 psig, all Steam Dump Valves will be closed by the P-12 interlock.

Proposed Answer: D Page 8 of 34 CPNPP 2014 NRC RO Written Exam Worksheet 61 to 70

ES-401 RO Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible because on a plant trip the HI-1 and HI-2 bistables input to the steam dump controller based on differential temperature between actual Tave and no-load Tave (557°F) at 1092 psig steam pressure, however once placed in the Steam Pressure Mode the HI-1 and HI-2 bistables are no longer in the dump control circuit. With setpoint at zero, the controller will be calling for the steam dumps to open to maintain 200 psig so all the Steam Dump Valves will open until RCS Tave reaches 553°F (1057 psig steam pressure) at which time all Steam Dump Valves will close.

B. Incorrect. Plausible because on a plant trip the HI-1 and HI-2 bistables input to the steam dump controller based on differential temperature between actual Tave and no-load Tave (557°F) at 1092 psig steam pressure, however once placed in the Steam Pressure Mode the HI-1 and HI-2 bistables are no longer in the steam dump control circuit. With setpoint at zero, the controller will be calling for the steam dumps to open to maintain 200 psig so all the Steam Dump Valves will open until RCS Tave reaches 553°F (1057 psig steam pressure) at which time all Steam Dump Valves will close, however, three dump valves (cooldown valves) may be bypassed to allow the cooldown valves to be open below 553°F.

C. Incorrect. Plausible because 1-PK-507 compares controller setpoint to steam header pressure and creates an error signal to position the Steam Dump Valves. With setpoint at zero the controller will be calling for the steam dumps to open to maintain 200 psig so all the dump valves will open until RCS Tave reaches 553°F (1057 psig steam pressure) at which time all dump valves will close, however three dump valves (cooldown valves) may be bypassed to allow the cooldown valves to be open below 553°F.

D. Correct. 1-PK-507 compares controller setpoint to steam header pressure and creates an error signal to position the Steam Dump Valves. With setpoint at zero, the controller will be calling for the steam dumps to open to maintain 200 psig so all the Steam Dump Valves will open until RCS Tave reaches 553°F (1057 psig steam pressure) at which time all Steam Dump Valves will close.

Technical Reference(s) LO21.SYS.SD1, Figure 3 Attached w/ Revision: See Steam Tables Comments / Reference Proposed references to be provided during examination: Steam Tables Learning Objective: DEMONSTRATE an understanding of the components of the Steam Dump System including interrelations with other systems to include interlocks and control loops.

Question Source: Bank Modified Bank ILOT5907 (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 Page 9 of 34 CPNPP 2014 NRC RO Written Exam Worksheet 61 to 70

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

LO21.SYS.SD1, Figure 3 Revision: 11/30/10 Page 10 of 34 CPNPP 2014 NRC RO Written Exam Worksheet 61 to 70

ES-401 RO Written Exam Worksheet Form ES-401-5 Original Question: ILOT5907 The plant was tripped from 20% Reactor Power. The steam dump system pressure controller potentiometer is set to zero and in AUTO. Which ONE of the following statements describes steam dump system response if the select switch is placed in the steam pressure mode?

A. The HI-1 and HI-2 bistables will actuate causing all steam dump valves to open. When RCS Tave reaches 553 degrees, all steam dump valves will be closed by the P-12 interlock.

B. A proportional error signal will open all steam dumps. When RCS Tave reaches 553 degrees, all but three valves close. PV-2369A, B, C will be controlled by the proportional controller.

C. A proportional error signal will open all steam dumps. RCS Tave reaches 553 degrees, all steam dump valves will be closed by the P-12 interlock.

D. The HI-1 and HI-2 bistables will actuate causing all steam dump valves to open. When Tave reaches NO-LOAD condition, valves will modulate to maintain NO-LOAD Tave.

Answer: C Page 11 of 34 CPNPP 2014 NRC RO Written Exam Worksheet 61 to 70

ES-401 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. Date: 3/3/2014 Tier 2 Change: 2 Group 2 K/A 045 A4.06 Level of Difficulty: 4 Importance Rating 2.8 Main Turbine Generator System: Ability to manually operate and/or monitor in the control room: Turbine stop valves Proposed Question: 64 Given the following conditions:

Unit 1 is responding to an inadvertent closure of Turbine High Pressure (HP) Stop Valve 1 in accordance with ABN-401, Main Turbine Malfunction.

Engineering has recommended that HP Stop Valve 1 be opened from the Control Room using the HP Valve Test Display in accordance with OPT-217A, Turbine Overspeed Protection System Test.

Which of the following describes the expected sequence of component operation when using the HP Valve Test Display to open HP Stop Valve 1?

A. CLOSE HP Control Valve 1.

TRIP HP Stop Valve 1.

OPEN HP Stop Valve 1.

OPEN HP Control Valve 1.

B. TRIP HP Stop Valve 1.

CLOSE HP Control Valve 1.

OPEN HP Stop Valve 1.

OPEN HP Control Valve 1.

C. CLOSE HP Control Valve 1.

TRIP HP Stop Valve 1.

OPEN HP Control Valve 1.

OPEN HP Stop Valve 1.

D. TRIP HP Stop Valve 1.

CLOSE HP Control Valve 1.

OPEN HP Control Valve 1.

OPEN HP Stop Valve 1.

Proposed Answer: A Page 12 of 34 CPNPP 2014 NRC RO Written Exam Worksheet 61 to 70

ES-401 RO Written Exam Worksheet Form ES-401-5 Explanation:

A. Correct. IAW OPT-217A, the correct sequence of valve operation is to close the HP CTRL VLV, trip the HP STOP VLV, then open the HP STOP VLV and open the HP CTRL VLV.

B. Incorrect. Plausible because with HP Stop Valve 1 closed it could be thought that tripping the stop valve first would prepare the stop valve for re-opening prior to closing the control valve.

C. Incorrect. Plausible because it could be thought that with the stop valve failing closed that the sequence of the control valve opening prior to opening the stop valve would ensure operation of the stop valve with a higher differential pressure.

D. Incorrect. Plausible because with HP Stop Valve 1 closed it could be thought that tripping the stop valve first would prepare the stop valve for re-opening prior to closing the control valve and that with the stop valve failing closed that the sequence of the control valve opening prior to opening the stop valve would ensure operation of the stop valve with a higher differential pressure.

Technical Reference(s) ABN-401, Step 9.3.14 Attached w/ Revision: See OPT-217A, Step 8.1.I NOTE Comments / Reference Proposed references to be provided during examination: None Learning Objective: COMPREHEND the normal, abnormal and emergency operation of Main Turbine and its support 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 Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Page 13 of 34 CPNPP 2014 NRC RO Written Exam Worksheet 61 to 70

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

ABN-401, Step 9.3.14 Revision: 12 Page 14 of 34 CPNPP 2014 NRC RO Written Exam Worksheet 61 to 70

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

OPT-217A, Step 8.1.I NOTE Revision: 17 Page 15 of 34 CPNPP 2014 NRC RO Written Exam Worksheet 61 to 70

ES-401 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. Date: 5/16/2014 Tier 2 Change: 4 Group 2 K/A 086 K4.01 Level of Difficulty: 4 Importance Rating 3.1 Fire Protection System: Knowledge of design feature(s) and/or interlock(s) which provide for the following: Adequate supply of water for FPS Proposed Question: 65 Given the following condition:

Both Units are in MODE 1.

In accordance with SOP-904, Fire Protection Main Water Supply and Fire Pumps System, which of the following is the MINIMUM acceptable level in each Fire Water Storage Tank?

A. 75%.

B. 80%.

C. 85%.

D. 90%.

Proposed Answer: C Explanation:

A. Incorrect. Plausible as level is close to acceptable level.

B. Incorrect. Plausible as level is close to acceptable level.

C. Correct. Minimum acceptable level is 82% and 85% is greater then 82%.

D. Incorrect. Plausible as level is close to acceptable level.

Technical Reference(s) SOP-904, Section 4.1 Attached w/ Revision: See OWI-104-18 Comments / Reference Proposed references to be provided during examination: None Learning Objective: COMPREHEND the normal, abnormal and emergency operations of the Fire Protection System.

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

New X Page 16 of 34 CPNPP 2014 NRC RO Written Exam Worksheet 61 to 70

ES-401 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 10 55.43 Comments /

Reference:

SOP-904, Section 4.1 Revision: 15 Page 17 of 34 CPNPP 2014 NRC RO Written Exam Worksheet 61 to 70

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

OWI-104-18 Revision: 60 Page 18 of 34 CPNPP 2014 NRC RO Written Exam Worksheet 61 to 70

ES-401 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. Date: 5/22/2013 Tier 3 Change: 4 Category 1 K/A G 2.1.26 Level of Difficulty: 2 Importance Rating 3.4 Conduct of Operations: Knowledge of industrial safety procedures (such as rotating equipment, electrical, high temperature, high pressure, caustic, chlorine, oxygen and hydrogen)

Proposed Question: 66 Which of the following describes the MINIMUM clothing and flash protection boundary requirements when racking a 480 V Switchgear load breaker in accordance with STA-124, Electrical Safe Work Practices?

A flash suit and hood rated at greater than or equal to A. 8 cal/cm2 with a 10 ft boundary.

B. 8 cal/cm2 with a 20 ft boundary.

C. 50 cal/cm2 with a 10 ft boundary.

D. 50 cal/cm2 with a 20 ft boundary.

Proposed Answer: C or D Explanation:

2 2 A. Incorrect. Plausible as the 8 cal/cm suit would be used for 480 V MCC breakers but a 50 cal/cm flash suit and hood is required for racking 480 V switchgear breakers. 10 feet is the appropriate minimum flash protection boundary.

2 2 B. Incorrect. Plausible as the 8 cal/cm suit would be used for 480 V MCC breakers but a 50 cal/cm flash suit and hood is required for racking 480 V switchgear breakers, but 20 feet is the appropriate minimum flash protection boundary for 6.9 kV switchgear breakers.

2 C. Correct. In accordance with STA-124 and SOP-604A a 50 cal/cm flash suit and hood with a 10 foot flash boundary is required per STA-124 Hazard/Risk matrix.

2 D. Correct. The 50 cal/cm flash suit and hood is required for racking 480 V switchgear breakers. 20 feet is the appropriate minimum flash protection boundary for 6.9 kV switchgear breakers, however due to the lack of the word MINIMUM prior to flash protection boundary in the stem of the question it could be reasoned that if the boundary is 10 feet then 20 feet is inclusive of the 10 foot boundary.

Technical Reference(s) STA-124, Att. 8.A Attached w/ Revision: See SOP-604A, Page 113. Comments / Reference Proposed references to be provided during examination: None Page 19 of 34 CPNPP 2014 NRC RO Written Exam Worksheet 61 to 70

ES-401 RO Written Exam Worksheet Form ES-401-5 Learning Objective: INTERPRET and ENSURE compliance with plant administrative and operational procedures, guidelines, and policies.

Page 20 of 34 CPNPP 2014 NRC RO Written Exam Worksheet 61 to 70

ES-401 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 10 55.43 Page 21 of 34 CPNPP 2014 NRC RO Written Exam Worksheet 61 to 70

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

STA-124 Revision: 2 Page 22 of 34 CPNPP 2014 NRC RO Written Exam Worksheet 61 to 70

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

SOP-604A Revision: 12 Page 23 of 34 CPNPP 2014 NRC RO Written Exam Worksheet 61 to 70

ES-401 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. Date: 5/22/2014 Tier 3 Change: 3 Category 1 K/A G 2.1.3 Level of Difficulty: 2 Importance Rating 3.7 Conduct of Operations: Knowledge of shift or short-term relief turnover practices Proposed Question: 67 Given the following conditions:

At 1315, the Unit 2 Reactor Operator must leave the Control Room for a short period of time to get an annual audiometric test.

In order for the short term relief to be in compliance with OWI-107, Operations Department Turnover and Briefing Instructions the Reactor Operator should return to the AT THE CONTROLS AREA no later than _____ and at a MINIMUM permission should be granted by the ________.

A. ...1345; Unit 2 Unit Supervisor.

B. ...1345; Shift Manager.

C. ...1415; Unit 2 Unit Supervisor.

D. ...1415; Shift Manager.

Proposed Answer: C Explanation:

A. Incorrect. Plausible because it could be thought 30 minutes is the short term relief time limit; however OWI-107 defines short term relief as < 60 minutes. In accordance with OWI-107, the respective Unit Supervisor is the permission authority for operators on the unit.

B. Incorrect. Plausible because it could be thought 30 minutes is the short term relief time limit; however OWI-107 defines short term relief as < 60 minutes. In accordance with OWI-107, the Shift Manager is the permission authority for the Unit Supervisors.

C. Correct. Short term relief is no longer than 60 minutes in accordance with OWI-107. In accordance with OWI-107, the respective Unit Supervisor is the permission authority for operators on the unit.

D. Incorrect Short term relief is no longer than 60 minutes in accordance with OWI-107. In accordance with OWI-107, the Shift Manager is the permission authority for the Unit Supervisors.

Page 24 of 34 CPNPP 2014 NRC RO Written Exam Worksheet 61 to 70

ES-401 RO Written Exam Worksheet Form ES-401-5 Technical Reference(s) OWI-107, Step 6.1.4 Attached w/ Revision: See Comments / Reference Proposed references to be provided during examination: None Learning Objective: CONDUCT shift relief and turnover in accordance with station procedures; VERIFYING that an adequate number of qualified personnel are available for turnover and ENSURING that all personnel are properly relieved.

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:

OWI-107, Step 6.1.4 Revision: 8 Page 25 of 34 CPNPP 2014 NRC RO Written Exam Worksheet 61 to 70

ES-401 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. Date: 5/22/2014 Tier 3 Change: 3 Category 1 K/A G 2.1.4 Level of Difficulty: 3 Importance Rating 3.3 Conduct of Operations: Knowledge of individual licensed operator responsibilities related to shift staffing, such as medical requirements no-solo operation, maintenance of active license status, 10CFR55, etc.

Proposed Question: 68 Given the following conditions:

A Nuclear Regulatory Commission (NRC) Licensed Reactor Operator (RO) has just been convicted of a felony.

An appeal has been filed on their behalf.

Which of the following is the MAXIMUM time allowed for written notification to the NRC?

A. If the conviction is upheld following appeal, the RO must notify the NRC in writing within 60 days of the date the conviction was upheld.

B. The RO is responsible for notifying the NRC in writing within 60 days of the date of the conviction.

C. The RO is responsible for notifying the NRC in writing within 30 days of the date of the conviction.

D. If the conviction is upheld following appeal, the RO must notify the NRC in writing within 30 days of the date the conviction was upheld.

Proposed Answer: C Explanation:

A. Incorrect. Plausible because 60 days is a common reporting time for non-routine reports; however the RO must notify the NRC in writing within 30 days of the conviction and there is no allowance for any appeal time.

B. Incorrect. Plausible because 60 days is a common reporting time for non-routine reports; however the RO must notify the NRC in writing within 30 days of the conviction.

C. Correct. As required per STA-501, Attachment 8.B and 10CFR55.53(g).

D. Incorrect. Plausible because the RO must notify the NRC in writing within 30 days of the conviction; however there is no allowance for any appeal time.

Technical Reference(s) STA-501, Attachment 8.B Attached w/ Revision: See Comments / Reference Proposed references to be provided during examination: None Page 26 of 34 CPNPP 2014 NRC RO Written Exam Worksheet 61 to 70

ES-401 RO Written Exam Worksheet Form ES-401-5 Page 27 of 34 CPNPP 2014 NRC RO Written Exam Worksheet 61 to 70

ES-401 RO Written Exam Worksheet Form ES-401-5 Learning Objective: Given an event related to system operation/status, CLASSIFY which events must be reported to external agencies, with respect to written reports.

Question Source: Bank ILOT4548 Modified Bank (Note changes or attach parent)

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

Reference:

STA-501, Attachment 8.B Revision: 17 Page 28 of 34 CPNPP 2014 NRC RO Written Exam Worksheet 61 to 70

ES-401 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. Date: 5/19/2014 Tier 3 Change: 3 Category 2 K/A G 2.2.39 Level of Difficulty: 3 Importance Rating 3.9 Equipment Control: Knowledge of less than or equal to one hour Technical Specification action statements for systems Proposed Question: 69 Unit 1 is in MODE 6 and the Shift Chemist reports that the boron concentration from the last sample of the refueling canal is less than required by Technical Specification LCO 3.9.1, Boron Concentration.

Which of the following describes the action requirements of Technical Specification LCO 3.9.1, Boron Concentration?

A. Within one hour verify all dilution paths are isolated.

(1CS-8455 or valves 1CS-8560, 1-FCV-111B, 1CS-8441, and 1CS-8453.)

B. Immediately suspend all operations involving CORE ALTERATIONS and positive reactivity changes and initiate boration of the Reactor Coolant System.

C. Within one hour initiate boration until the boron concentration is within limits specified in the COLR.

D. Immediately suspend all movement of fuel assemblies in the Refueling Canal and restore the boron concentration to within its limit within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

Proposed Answer: B Explanation:

A. Incorrect. Plausible because positive reactivity additions must be immediately suspended and isolating dilution flow paths would assist in performing this action, however, this action is required to be performed immediately.

B. Correct. As outlined in Technical Specification LCO 3.9.1.

C. Incorrect. Plausible because Boration must be performed to restore boron concentration to greater than the limit specified in the COLR, however, it must be immediately restored to within limits.

D. Incorrect. Plausible because movement of fuel assemblies must be suspended, however, boron concentration must be immediately restored to within limits.

Technical Reference(s) Technical Specification LCO 3.9.1 Attached w/ Revision: See Comments / Reference Proposed references to be provided during examination: None Page 29 of 34 CPNPP 2014 NRC RO Written Exam Worksheet 61 to 70

ES-401 RO Written Exam Worksheet Form ES-401-5 Learning Objective: APPLY the administrative requirements of the Spent Fuel Pool Cooling and Cleanup system including Technical Specifications, TRM and ODCM.

Question Source: Bank ILOT5517 Modified Bank (Note changes or attach parent)

New Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Page 30 of 34 CPNPP 2014 NRC RO Written Exam Worksheet 61 to 70

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

Technical Specification LCO 3.9.1 Amendment 161 Page 31 of 34 CPNPP 2014 NRC RO Written Exam Worksheet 61 to 70

ES-401 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. Date: 3/27/2014 Tier 3 Change: 3 Category 2 K/A G 2.2.37 Level of Difficulty: 3 Importance Rating 3.6 Equipment Control: Ability to determine operability and/or availability of safety related equipment Proposed Question: 70 Which of the following describes equipment that is available but NOT operable?

A. Safety Injection Accumulator parameters have been verified in accordance with IPO-001A, Plant Heatup from Cold Shutdown to Hot Standby.

B. Main Steam Isolation Valve closure times have been verified in accordance with IPO-002A, Startup from Hot Standby.

C. Auxiliary Feedwater Flow Control and Isolation Valve positions have been verified in accordance with IPO-003A, Power Operations.

D. Safety Injection train alignment has been verified in accordance with IPO-010A, RCS Reduced Inventory Operations.

Proposed Answer: D Page 32 of 34 CPNPP 2014 NRC RO Written Exam Worksheet 61 to 70

ES-401 RO Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible because a misconception could exist that the actions taken in IPO-001A for the SI accumulators are to place the accumulators in an Available status, but the actions are actually to ensure operability of the accumulators.

B. Incorrect. Plausible because a misconception could exist that the actions taken in IPO-002A for the MSIVs are to place the accumulators in an Available status, but the actions are actually to ensure operability of the MSIVs.

C. Incorrect. Plausible because a misconception could exist that the actions taken in IPO-001A for the AFW flow control and isolation valves are to place the AFW system in an Available status, but the actions are actually to ensure operability of the AFW system.

D. Correct. The SI train alignment actions in IPO-010A are to ensure that the SI train is Available, but the pump breaker is racked out ensuring that the SI train is not operable.

Technical Reference(s) IPO-010A, Attachment 1 Attached w/ Revision: See IPO-001A, Step 5.3.15, IPO-002A, Step Comments / Reference 5.1.3, IPO-003A, IPO-003A, Step 5.1.7 Proposed references to be provided during examination: None Learning Objective: APPLY the administrative requirements of the Reactor Coolant System including Technical Specifications, TRM and ODCM.

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 Page 33 of 34 CPNPP 2014 NRC RO Written Exam Worksheet 61 to 70

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

IPO-010A, Attachment 1 Revision: 18 Page 34 of 34 CPNPP 2014 NRC RO Written Exam Worksheet 61 to 70

ES-401 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. Date: 5/19/2014 Tier 3 Change: 1 Category 3 K/A G 2.3.5 Level of Difficulty: 2 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: 71 Given the following conditions:

A Portable Frisker is being used to perform a whole body frisk.

Background radiation is at 100 counts per minute.

Which of the following is the MINIMUM count rate at which an individual is considered to be contaminated in accordance with STA-653, Contamination Control Program?

A. 175 counts per minute B. 225 counts per minute C. 275 counts per minute D. 325 counts per minute Proposed Answer: B Explanation:

A. Incorrect. Plausible if actual level of 100 cpm above background cannot be recalled.

B. Correct. With a background radiation of 100 cpm + a detected radiation level of 100 cpm above background = 200 cpm. Therefore the minimum choice that would indicate that the worker is contaminated would be 225 cpm.

C. Incorrect. Plausible if actual level of 100 cpm above background cannot be recalled.

D. Incorrect. Plausible if actual level of 100 cpm above background cannot be recalled.

Technical Reference(s) STA-653, Step 6.6.2 & Attachment 3 Attached w/ Revision: See Comments / Reference Proposed references to be provided during examination: None Learning Objective: EXPLAIN how to monitor personnel and personal items for contamination, including the use of friskers and personnel contamination monitors.

Question Source: Bank ILOT8316 Modified Bank (Note changes or attach parent)

New Page 1 of 22 CPNPP 2014 NRC RO Written Exam Worksheet 71 to 75

ES-401 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 11 55.43 Comments /

Reference:

STA-653, Step 6.6.2 Revision: 16 Page 2 of 22 CPNPP 2014 NRC RO Written Exam Worksheet 71 to 75

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

STA-653, Attachment 3 Revision: 16 Page 3 of 22 CPNPP 2014 NRC RO Written Exam Worksheet 71 to 75

ES-401 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. Date: 5/19/2014 Tier 3 Change: 3 Category 3 K/A G 2.3.14 Level of Difficulty: 3 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: 72 Given the following condition:

Unit 2 is in MODE 6 and Containment Fan Coolers are being alternated in accordance with SOP-801B, Containment Ventilation System.

In accordance with SOP-801B, Containment Ventilation System, which of the following describes the CAUTION applicable to alternating the running Containment Fan Coolers?

A. Alternating Fan Coolers can cause changes in indicated radiation levels due to noble gases in stagnant pockets of air.

B. Alternating Fan Coolers cannot be performed with a Containment Purge or Containment Vent in progress.

C. Core alterations and movement of irradiated fuel assemblies in containment must be stopped prior to alternating fan coolers.

D. Monitor to ensure air flows into Containment from the Equipment Hatch and not from Containment out of the Equipment Hatch.

Proposed Answer: A Page 4 of 22 CPNPP 2014 NRC RO Written Exam Worksheet 71 to 75

ES-401 RO Written Exam Worksheet Form ES-401-5 Explanation:

A. Correct. The CAUTION is concerned with stagnant pockets of air with noble gases that will be introduced into the Containment atmosphere and the potential to cause automatic Containment Isolation.

B. Incorrect. Plausible because SOP-801A Section 5.1.3 has instructions for additional actions which must be performed if a Containment Purge or Vent is in progress. As such, alternating the fan coolers is allowed by taking the additional actions and therefore the statement that it cannot be performed is incorrect.

C. Incorrect. Plausible because SOP-801A Section 5.1.3 has instructions for additional actions which must be performed if a Containment Purge or Vent is in progress. These actions include that CVI cannot be disabled if core alterations or movement of irradiated fuel assemblies in containment is in progress. However, other alternatives for continuing exist. As such, alternating the fan coolers is allowed when core alterations or movement of irradiated fuel assemblies in containment is in progress by taking the additional actions and therefore the statement that it cannot be performed is incorrect.

D. Incorrect. Plausible because this is a concern in ventilation changes that change air flow out of or into Containment, however, these fans recirculate air and alternating these fans does not change the air flow through the Equipment Hatch.

Technical Reference(s) SOP-801B, Step 5.1.3 CAUTION; 5.1.3.B Attached w/ Revision: See LO21.SYS.CL1, Page 8 Comments / Reference Proposed references to be provided during examination: None Learning Objective: DESCRIBE the basis for the precautions and limitations, and given major procedure steps relative to the Containment Ventilation system, PLACE them in the proper sequence for SOP-801, Containment Ventilation System.

Question Source: Bank ILOT8246 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 9 55.43 Page 5 of 22 CPNPP 2014 NRC RO Written Exam Worksheet 71 to 75

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

SOP-801B, Step 5.1.3 CAUTION; 5.1.3.B Revision: 7 Page 6 of 22 CPNPP 2014 NRC RO Written Exam Worksheet 71 to 75

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

LO21.SYS.CL1, Page 6 Revision: 05/02/11 Containment Air Cooling and Recirculation System The Containment Air Cooling and Recirculation System is designed to circulate cool air throughout the Containment structure, equipment rooms, and cubicles. The system functions to remove all heat released to the Containment by the reactor, steam generators and related equipment during all normal plant operations and following a loss of offsite power. Heat is dissipated to the Ventilation Chilled Water System in order to maintain ambient temperature 120°F, thereby protecting equipment and structures. The system provides a supply of tempered air to the Control Rod Drive Mechanism Ventilation System and the Reactor Coolant Pipe Penetration Cooling System. The system consists of four (4) 33% capacity cooling units and fans, each capable of handling 65,000 cfm of air.

Each cooling unit, referred to as Containment Fan Coolers, consists of eight (8) cooling coils stacked two high in a rectangular array. The structure also supports a vane axial, direct drive fan.

This arrangement allows warm air to be drawn through the coils by the fan, cooled, and delivered to the main supply plenum located on the 884 elevation, which is connected to ductwork that supplies air throughout the Containment Building. Cool air is supplied to the open areas and the equipment rooms to ensure mixing of the Containment atmosphere. No return ductwork is installed. The warm air rises through openings and gratings in floors and cubicle walls. Two cooling units are located on the 860' level of the Containment, each with a fan mounted above it to blow cooled air up to the main supply plenum. Two units are located on the 905' level of the Containment with fans mounted below to supply cooled air down to the main supply plenum.

Pneumatic dampers are installed at the outlet of each fan to prevent back flow through the standby unit.

Page 7 of 22 CPNPP 2014 NRC RO Written Exam Worksheet 71 to 75

ES-401 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. Date: 5/19/2014 Tier 3 Change: 3 Category 3 K/A G 2.3.12 Level of Difficulty: 3 Importance Rating 3.2 Radiation Control: Knowledge of radiological safety principles pertaining to licensed operator duties, such as containment entry requirements, fuel handling responsibilities, access to locked high radiation areas, aligning filters, etc.

Proposed Question: 73 Given the following conditions:

A room containing highly contaminated resin has been posted in the Fuel Building.

General Dose Rates are 1500 mR/hr.

Dose Rates at 1 foot from the contaminated resin are as high as 20 R/hr.

Which of the following is the type of radiological area and who is the LOWEST approval authority required for entry in accordance with STA-660, Control of High Radiation Areas?

A. Very High Radiation Area (VHRA).

Plant Manager.

B. Locked High Radiation Area (LHRA).

Plant Manager.

C. Very High Radiation Area (VHRA).

Radiation Protection Manager.

D. Locked High Radiation Area (LHRA).

Radiation Protection Manager.

Proposed Answer: D Page 8 of 22 CPNPP 2014 NRC RO Written Exam Worksheet 71 to 75

ES-401 RO Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible if thought that greater than 1000mr/hr is VHRA as the lowest approval authority for entry into a VHRA is the Plant Manager in accordance with STA-660.

B. Incorrect. Plausible as the area meets requirements for a LHRA. However, the lowest approval authority for entry into a LHRA with a dose rate of 10 R/hr or greater is the Radiation Protection Manger not the Plant Manager in accordance with STA-660.

C. Incorrect. Plausible if thought that greater than 1000mr/hr is VHRA and the lowest approval authority for the stated conditions is the Radiation Protection Manage in accordance with STA-660.

D. Correct. The area meets requirements for posting as a LHRA and the lowest approval authority for the stated conditions is the Radiation Protection Manage in accordance with STA-660.

Technical Reference(s) STA-660, Section 4.4 & 4.11 Attached w/ Revision: See STA-660, Step 6.2.9 & 6.3.6.6 Comments / Reference STA-660-1 Proposed references to be provided during examination: None Learning Objective: INTERPRET and ENSURE compliance with plant administrative and operational procedures, guidelines, and policies.

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 12 55.43 Page 9 of 22 CPNPP 2014 NRC RO Written Exam Worksheet 71 to 75

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

STA-660, Section 4.4 & 4.11 Revision: 15 Page 10 of 22 CPNPP 2014 NRC RO Written Exam Worksheet 71 to 75

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

STA-660, Step 6.2.9 Revision: 15 Page 11 of 22 CPNPP 2014 NRC RO Written Exam Worksheet 71 to 75

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

STA-660, Step 6.3.6.6 Revision: 15 Page 12 of 22 CPNPP 2014 NRC RO Written Exam Worksheet 71 to 75

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

STA-660-1 Revision: 14 Page 13 of 22 CPNPP 2014 NRC RO Written Exam Worksheet 71 to 75

ES-401 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. Date: 3/6/2014 Tier 3 Change: 3 Category 4 K/A G 2.4.6 Level of Difficulty: 2 Importance Rating 3.7 Emergency Procedures/Plan: Knowledge of EOP mitigation strategies Proposed Question: 74 Given the following conditions:

Unit 1 is responding to a ruptured Steam Generator in accordance with EOP-3.0A, Steam Generator Tube Rupture.

The ruptured Steam Generator has been identified and isolated.

The Reactor Coolant System cooldown to target Core Exit Thermocouple temperature has been completed and required subcooling has been established.

Which of the following describes the next major action categories to be accomplished in EOP-3.0A, Steam Generator Tube Rupture to mitigate the tube rupture?

A. Terminate Safety Injection, and then Prepare to Cooldown to Cold Shutdown.

B. Terminate Safety Injection, and then Depressurize the Reactor Coolant System.

C. Depressurize the Reactor Coolant System, and then Terminate Safety Injection.

D. Depressurize the Reactor Coolant System, and then Prepare to Cooldown to Cold Shutdown.

Proposed Answer: C Explanation:

A. Incorrect. Plausible because these are the two major action categories, in order, that are performed following the depressurization of the RCS. However, Depressurization of the RCS must be performed prior to attempting to Terminate Safety Injection.

B. Incorrect. Plausible because these are the two major action categories that are performed following the completion of the RCS cooldown. However, they are listed in reverse order.

C. Correct. IAW EOP-3.0A, depressurizing the RCS and terminating Safety Injection are the next two major action categories to be performed following completion of the RCS Cooldown.

D. Incorrect. Plausible because Depressurize the RCS is the next major action category which must be performed. However, the preparation for the Cooldown to Cold Shutdown cannot be completed until Safety injection termination has taken place.

Page 14 of 22 CPNPP 2014 NRC RO Written Exam Worksheet 71 to 75

ES-401 RO Written Exam Worksheet Form ES-401-5 Technical Reference(s) EOP-3.0A, Flowchart Attached w/ Revision: See Comments / Reference Proposed references to be provided during examination: None Learning Objective: DISCUSS the analysis of a SGTR including the operator expected actions and associated times.

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:

EOP-3.0A, Flowchart Revision: 8 Page 15 of 22 CPNPP 2014 NRC RO Written Exam Worksheet 71 to 75

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

EOP-3.0A, Attachment 6, Step 19 Bases Revision: 8 Page 16 of 22 CPNPP 2014 NRC RO Written Exam Worksheet 71 to 75

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

EOP-3.0A, Attachment 6, Step 19 Bases Revision: 8 Comments /

Reference:

EOP-3.0A Bases, Step 23 Revision: 8 Page 17 of 22 CPNPP 2014 NRC RO Written Exam Worksheet 71 to 75

ES-401 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. Date: 3/3/2014 Tier 3 Change: 2 Category 4 K/A G 2.4.32 Level of Difficulty: 2 Importance Rating 3.6 Emergency Procedures/Plan: Knowledge of operator response to loss of all annunciators Proposed Question: 75 Given the following conditions:

Unit 2 is responding to a loss of all Control Room Annunciators in accordance with ABN-740B, Control Room Annunciator System and Status Light Malfunction.

Unit load is stable at 100%.

All work on Unit 2 has been stopped.

Which of the following surveillances must be initiated while the annunciators are out of service?

OPT-303, Reactor Coolant System Water Inventory, ...

A. ...OPT-112B, Accident Monitoring Instrumentation Check, and OPT-302, Calculating Power Tilt Ratio.

B. ...OPT-112B, Accident Monitoring Instrumentation Check, and OPT-309, Unit Calorimetric.

C. ...OPT-309, Unit Calorimetric, and OPT-403, Axial Flux Difference.

D. ...OPT-302, Calculating Power Tilt Ratio, and OPT-403, Axial Flux Difference.

Proposed Answer: D Page 18 of 22 CPNPP 2014 NRC RO Written Exam Worksheet 71 to 75

ES-401 RO Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible because OPT-302, QPTR is required, however OPT-112B, Accident Monitoring Instrumentation Checks is not required. It could be thought that OPT-112B performance is required because it is required by ABN-906, Plant Process Computer System Malfunction.

B. Incorrect. Plausible because OPT-112B and OPT-309 performance is required but it is required by ABN-906, Plant Process Computer System Malfunction.

C. Incorrect. Plausible because OPT-403, AFD is required, however OPT-309, Unit Calorimetric is not required. It could be thought that OPT-309 performance is required because it is required by ABN-906, Plant Process Computer System Malfunction.

D. Correct. IAW ABN-740B, Attachment 1, OPT-303, OPT-302 and OPT-403 are required during a loss of all annunciators.

Technical Reference(s) ABN-740B, Attachment 1 Attached w/ Revision: See ABN-906, Steps 2.3.2 RNO & 2.3.3 RNO Comments / Reference Proposed references to be provided during examination: None Learning Objective: INTERPRET and ENSURE compliance with plant administrative and operational procedures, guidelines, and policies.

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 19 of 22 CPNPP 2014 NRC RO Written Exam Worksheet 71 to 75

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

ABN-740B, Attachment 1 Revision: 1 Page 20 of 22 CPNPP 2014 NRC RO Written Exam Worksheet 71 to 75

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

ABN-906, Step 2.3.2 RNO Revision: 7 Page 21 of 22 CPNPP 2014 NRC RO Written Exam Worksheet 71 to 75

ES-401 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

ABN-906, Step 2.3.3 RNO Revision: 7 Page 22 of 22 CPNPP 2014 NRC RO Written Exam Worksheet 71 to 75

ES-401 SRO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. Date: 2/27/2014 Tier 1 Change: 2 Group 1 K/A 009 EA2.02 Level of Difficulty: 3 Importance Rating 3.8 Small Break LOCA: Ability to determine and interpret the following as they apply to the small break LOCA: Possible leak paths Proposed Question: 76 Given the following conditions:

A Small Break Loss of Coolant Accident (LOCA) is in progress on Unit 2.

A cooldown is in progress in accordance with EOS-1.2B, Post LOCA Cooldown and Depressurization.

Containment pressure is 3.5 psig and slowly rising.

One Control Rod failed to insert on the Reactor Trip.

Reactor Coolant Pumps are stopped.

Reactor Coolant System (RCS) pressure is 1585 psig and slowly lowering.

Subcooled Margin is 10°F and becoming less subcooled.

Pressurizer level is 94% and rising.

The RVLIS 49 above the flange light is DARK and all other RVLIS lights are LIT.

The Shift Technical Advisor reports the Inventory Critical Safety Function is YELLOW.

Which of the following describes the location of the Reactor Coolant System leak and the expected mitigation actions?

A. Pressurizer Steam Space.

Remain in EOS-1.2B, Post LOCA Cooldown and Depressurization and continue cooldown to less than 200°F.

B. Reactor Vessel Upper Head.

Remain in EOS-1.2B, Post LOCA Cooldown and Depressurization and continue cooldown to less than 200°F.

C. Pressurizer Steam Space.

Transition to FRI-0.3B, Response to Voids in Reactor Vessel, and prepare to start a Reactor Coolant Pump.

D. Reactor Vessel Upper Head.

Transition to FRI-0.3B, Response to Voids in Reactor Vessel, and prepare to start a Reactor Coolant Pump.

Proposed Answer: A Page 1 of 27 CPNPP 2014 NRC SRO Written Exam Worksheet 76 to 80

ES-401 SRO Written Exam Worksheet Form ES-401-5 Explanation:

A. Correct. The rising Pressurizer level with voiding in the vessel head indicates a Pressurizer steam space leak exists and the correct action is to continue the cooldown to less than 200°F.

B. Incorrect. Plausible because the upper RVLIS light being DARK and a Control Rod failure to insert could be indication of a head leak, however, the Pressurizer level would not be rising. Remaining in EOS-1.2B and cooling down to less than 200°F is the correct action.

C. Incorrect. Plausible because the leak is in the Pressurizer steam space but transition to FRI-0.3B is not the correct action. The YELLOW path exists due to voids and Pressurizer level which could be thought the correct action to take, however, cooldown to less than 200°F is the correct action.

In accordance with ODA-407, In the prioritization scheme of the ERGs, the ORGs (including applicable foldout pages) have priority over YELLOW path FRG(s).

D. Incorrect. Plausible because the upper RVLIS light being DARK and a Control Rod failure to insert could be indication of a head leak however the Pressurizer level would not be rising. The YELLOW path exists due to voids and Pressurizer level which could be thought the correct action to take, however, cooldown to less than 200°F is the correct action. In accordance with ODA-407, In the prioritization scheme of the ERGs, the ORGs (including applicable foldout pages) have priority over YELLOW path FRG(s).

Technical Reference(s) EOS-1.2B, Flowchart Attached w/ Revision: See FRI-0.3B, CSFST Comments / Reference FRI-0.3B, Step 1 CAUTION Proposed references to be provided during examination: None Learning Objective: Given a set of plant conditions, IDENTIFY the proper transitions through/out of EOS-1.2B, Post LOCA Cooldown and Depressurization.

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 2 of 27 CPNPP 2014 NRC SRO Written Exam Worksheet 76 to 80

ES-401 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

EOS-1.2B, Flowchart Revision: 8 Page 3 of 27 CPNPP 2014 NRC SRO Written Exam Worksheet 76 to 80

ES-401 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

FRI-0.3B, CSFST Revision: 8 Page 4 of 27 CPNPP 2014 NRC SRO Written Exam Worksheet 76 to 80

ES-401 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

FRI-0.3B, Step 1 CAUTION Revision: 8 Page 5 of 27 CPNPP 2014 NRC SRO Written Exam Worksheet 76 to 80

ES-401 SRO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. Date: 2/27/2014 Tier 1 Change: 2 Group 1 K/A E04 EA2.01 Level of Difficulty: 2 Importance Rating 4.3 LOCA Outside Containment: Ability to determine and interpret the following as they apply to the (LOCA Outside Containment): Facility conditions and selection of appropriate procedures during abnormal and emergency operations.

Proposed Question: 77 Given the following conditions:

A Unit 1 Reactor Trip and Safety Injection have occurred.

The Safeguards Building area radiation monitors are in RED alarm.

All Containment Building parameters are normal.

ECA-1.2A, LOCA Outside Containment is in progress.

After closing 1/1-8835, Safety Injection to Cold Leg 1

• 4 Injection Isolation Valve, Reactor Coolant System pressure is 1850 psig and rising with Emergency Core Cooling System flow lowering.

Which of the following describes the status of the Loss of Coolant Accident and the required procedure transition?

A. The LOCA is isolated.

Transition to EOP-1.0A, Loss of Reactor or Secondary Coolant.

B. The LOCA is isolated.

Transition to ECA-1.1A, Loss of Emergency Coolant Recirculation.

C. The LOCA is NOT isolated.

Transition to EOP-1.0A, Loss of Reactor or Secondary Coolant.

D. The LOCA is NOT isolated.

Transition to ECA-1.1A, Loss of Emergency Coolant Recirculation.

Proposed Answer: A Page 6 of 27 CPNPP 2014 NRC SRO Written Exam Worksheet 76 to 80

ES-401 SRO Written Exam Worksheet Form ES-401-5 Explanation:

A. Correct. When 1/1-8835 is closed in Step 2 of ECA-1.2A and RCS pressure rises, this leads to a transition to EOP-1.0A at Step 3 of ECA-1.2A.

B. Incorrect. Plausible because the LOCA is isolated, however, a transition to ECA-1.1A is only required if RCS pressure is lowering (ECA-1.2A, Step 3 RNO).

C. Incorrect. Plausible because transition to EOP-1.0A is required, however, RCS pressure is rising so the LOCA is isolated.

D. Incorrect. Plausible if thought that transition to ECA-1.1A is required even when RCS pressure is rising due to the lack of inventory in the Containment.

Technical Reference(s) ECA-1.2A, Steps 2 & 3 Attached w/ Revision: See Comments / Reference Proposed references to be provided during examination: None Learning Objective: IDENTIFY the proper transitions out of ECA-1.2, LOCA Outside Containment.

Question Source: Bank ILOT5963 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 7 of 27 CPNPP 2014 NRC SRO Written Exam Worksheet 76 to 80

ES-401 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

ECA-1.2A, Step 2 Revision: 8 Page 8 of 27 CPNPP 2014 NRC SRO Written Exam Worksheet 76 to 80

ES-401 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

ECA-1.2A, Step 3 Revision: 8 Page 9 of 27 CPNPP 2014 NRC SRO Written Exam Worksheet 76 to 80

ES-401 SRO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. Date: 3/24/2014 Tier 1 Change: 4 Group 1 K/A 038 EA2.09 Level of Difficulty: 4 Importance Rating 4.2 Steam Generator Tube Rupture: Ability to determine and interpret the following as they apply to a SGTR: Existence of natural circulation, using plant parameters Proposed Question: 78 Given the following conditions:

Unit 1 is responding to a Steam Generator Tube Rupture on Steam Generator 1-01 in accordance with EOS-3.1A, Post-SGTR Cooldown Using Backfill.

All Reactor Coolant Pumps (RCP) were tripped on a loss of subcooling while performing EOP-0.0A, Reactor Trip or Safety Injection.

EOP-0.0A, Reactor Trip or Safety Injection, Attachment 9, Post Event System Realignment is completed with systems reset or restored.

The following indications exist:

Steam Generator 1-01 pressure is 785 psig and stable.

Steam Generator 1-01 level is 70% and stable.

Steam Generators 1-02, 1-03, & 1-04 pressures are 435 psig and lowering.

Steam Generators 1-02, 1-03 & 1-04 are 55% and stable.

Loop 1 Reactor Coolant Cold Leg temperature is 475°F and lowering.

Loops 2, 3, and 4 Reactor Coolant Cold Leg temperatures are 456°F and stable.

Loop 1 Reactor Coolant Hot Leg temperature is 490°F and lowering.

Loops 2, 3, and 4 Reactor Coolant Hot Leg temperatures are 476°F and lowering.

Reactor Coolant System pressure is 885 psig and lowering.

Highest reading Core Exit Thermocouple is 480°F and lowering.

Which of the following states the current natural circulation status and the action to be taken in accordance with EOS-3.1A, Post-SGTR Cooldown Using Backfill?

A. Natural Circulation exists.

Start a Reactor Coolant Pump.

B. Natural Circulation exists.

Do NOT start a Reactor Coolant Pump.

C. Natural Circulation does NOT exist.

Start a Reactor Coolant Pump.

D. Natural Circulation does NOT exist.

Do NOT start a Reactor Coolant Pump.

Page 10 of 27 CPNPP 2014 NRC SRO Written Exam Worksheet 76 to 80

ES-401 SRO Written Exam Worksheet Form ES-401-5 Proposed Answer: B Explanation:

A. Incorrect. Plausible because IAW EOS-3.1A, Attachment 3, all indications support Natural Circulation flow. However, EOS-3.1A requires 60°F of subcooling in order to restart an RCP and only 42°F subcooling exists.

B. Correct. IAW EOS-3.1A, Attachment 3, all indications support natural circulation flow and subcooling is 42°F which is not sufficient for RCP restart.

C. Incorrect. Plausible if incorrectly included Loop 1 values in the determination of Natural Circulation.

Additionally, it could be thought that the subcooling to determine existence of natural circulation also satisfies RCP restart criteria.

D. Incorrect. Plausible if incorrectly included Loop 1 values in the determination of Natural Circulation.

However, conditions do not support RCP restart.

Technical Reference(s) EOS-3.1A, Attachment 3 Attached w/ Revision: See EOS-3.1A, Steps 1 Comments / Reference Proposed references to be provided during examination: Steam Tables Learning Objective: Given a procedural Step, NOTE, or CAUTION, DISCUSS the reason or basis for the Step, NOTE, or CAUTION in EOP-3.0A, Steam Generator Tube Rupture.

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 11 of 27 CPNPP 2014 NRC SRO Written Exam Worksheet 76 to 80

ES-401 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

EOS-3.1A, Attachment 3 Revision: 8 Page 12 of 27 CPNPP 2014 NRC SRO Written Exam Worksheet 76 to 80

ES-401 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

EOS-3.1A Step 1 Revision: 8 Page 13 of 27 CPNPP 2014 NRC SRO Written Exam Worksheet 76 to 80

ES-401 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

EOS-3.1A, Step 1 Revision: 8 Page 14 of 27 CPNPP 2014 NRC SRO Written Exam Worksheet 76 to 80

ES-401 SRO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. Date: 3/27/2014 Tier 1 Change: 3 Group 1 K/A 040 G 2.1.7 Level of Difficulty: 4 Importance Rating 4.7 Steam Line Rupture: Conduct of Operations: Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation Proposed Question: 79 Given the following conditions:

A seismic event has occurred at Comanche Peak.

Unit 1 has experienced a Reactor Trip and Safety Injection.

A complete Main Steamline Isolation was verified in EOP-0.0A, Reactor Trip or Safety Injection.

Transition has been made to EOP-2.0A, Faulted Steam Generator Isolation.

The following parameters are observed:

Containment pressure is 15 psig and rising.

Reactor Coolant System (RCS) average temperature is 475°F and lowering.

RCS pressure is 1620 psig and lowering.

Pressurizer level is 32% and lowering.

Steam Generator (SG) 1-01 pressure is 15 psig.

SG 1-02, 1-03 & 1-04 pressures are 540 psig and lowering.

SG 1-01 has been isolated.

AFW flow to SGs 1-02, 1-03 and 1-04 is 180 gpm per SG.

SG 1-01 level is 20% wide range and rising.

SG 1-02, 1-03 & 1-04 levels are 55% narrow range and stable.

Which of the following is the first procedure transition required in accordance with EOP-2.0A, Faulted Steam Generator Isolation?

Transition to...

A. ...ECA-2.1A, Uncontrolled Depressurization of All Steam Generators.

B. ...EOP-3.0A, Steam Generator Tube Rupture.

C. ...EOS-1.1A, Safety Injection Termination.

D. ...EOP-1.0A, Loss of Reactor or Secondary Coolant.

Proposed Answer: B Page 15 of 27 CPNPP 2014 NRC SRO Written Exam Worksheet 76 to 80

ES-401 SRO Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible because the plant indications show pressure lowering in all intact steam generators, however the pressure is not decreasing in an uncontrolled manner based on the intact steam generators all being at the same pressure and at an expected pressure following blowdown of a single steam generator.

B. Correct. The rising level in the faulted steam generator and the lowering pressurizer level, RCS pressure and RCS temperature indicate that the faulted steam generator is also ruptured.

C. Incorrect. Plausible because subcooling is 100°F and a secondary heat sink exists however with RCS pressure lowering and pressurizer level less than 34% SI cannot be terminated.

D. Incorrect. Plausible because there would be no radiation monitor feedback for the faulted ruptured steam generator in this condition and the operator would have to diagnose that the faulted steam generator is not attaining dryout condition to avoid bypassing the transition to EOP-3.0A.

Technical Reference(s) EOP-2.0A, Steps 8 & 9 Attached w/ Revision: See EOP-2.0A, Attachment 3, Bases Comments / Reference Proposed references to be provided during examination: None Learning Objective: IDENTIFY the symptoms for the entry conditions of EOP-2.0, Faulted Steam Generator Isolation.

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 16 of 27 CPNPP 2014 NRC SRO Written Exam Worksheet 76 to 80

ES-401 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

EOP-2.0A, Step 7 Bases Revision: 8 Page 17 of 27 CPNPP 2014 NRC SRO Written Exam Worksheet 76 to 80

ES-401 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

EOP-2.0A, Step 2 Bases Revision: 8 Page 18 of 27 CPNPP 2014 NRC SRO Written Exam Worksheet 76 to 80

ES-401 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

EOP-2.0A, Step 2 Bases continued Revision: 8 Page 19 of 27 CPNPP 2014 NRC SRO Written Exam Worksheet 76 to 80

ES-401 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

EOP-2.0A, Step 8 Revision: 8 Page 20 of 27 CPNPP 2014 NRC SRO Written Exam Worksheet 76 to 80

ES-401 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

EOP-2.0A, Step 9 Revision: 8 Page 21 of 27 CPNPP 2014 NRC SRO Written Exam Worksheet 76 to 80

ES-401 SRO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. Date: 5/19/2014 Tier 1 Change: 3 Group 1 K/A 054 AA2.03 Level of Difficulty: 4 Importance Rating 4.2 Loss of Main Feedwater: Ability to determine and interpret the following as they apply to the Loss of Main Feedwater:

Conditions and reasons for AFW pump startup Proposed Question: 80 Which of the following would result in the start of the Unit 2 Motor Driven Auxiliary Feedwater (AFW) Pumps and the Technical Specification Bases for the MINIMUM number of Steam Generators required?

A. Narrow Range Level in ONE Steam Generator lowering to 37% following a Reactor Trip.

Ensure that at least ONE Steam Generator is available with water to act as a heat sink for the Reactor.

B. Trip of Main Feedwater Pump 2A at 18% Reactor Power.

Ensure that at least ONE Steam Generator is available with water to act as a heat sink for the Reactor.

C. Narrow Range Level in ONE Steam Generator lowering to 37% following a Reactor Trip.

Ensure that at least TWO Steam Generators are available with water to act as a heat sink for the Reactor.

D. Trip of Main Feedwater Pump 2A at 18% Reactor Power.

Ensure that at least TWO Steam Generators are available with water to act as a heat sink for the Reactor.

Proposed Answer: B Page 22 of 27 CPNPP 2014 NRC SRO Written Exam Worksheet 76 to 80

ES-401 SRO Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible at a single SG lowering to 37% on Unit 1 would start both Motor Driven AFW Pumps, however, on Unit 2 the level would need to lower to 35.4%. According to Technical Specification (TS) LCO 3.3.2 Bases, the reason for the AFW Pump start is to ensure at least one SG is available to act as a heat sink.

B. Correct. At 18% power only one Main Feedwater Pump would be in operation, however, the other Main Feedwater Pump would either be in a tripped condition or the trip oil pressure switches isolated to indicate a trip to the RPS. Thus, a trip of the operating Main Feedwater Pump would start both motor driven AFW Pumps. According to TS LCO 3.3.2 Bases, the reason for the AFW pump start is to ensure at least one SG is available to act as a heat sink.

C. Incorrect. Plausible at a signal SG lowering to 37% on Unit 1 would start both Motor Driven AFW Pumps, however on Unit 2 the level would need to lower to 35.4%. According to TS LCO 3.3.2 Bases, the reason as stated is incorrect in that the start signal only ensures at least ONE SG is available and not TWO. The distractor is plausible as Unit 1 requires two SG for SGTR recovery in accordance with the FSAR.

D. Incorrect. Plausible because the signal for starting the Motor Driven AFW Pumps is correct, however, the reason as stated is incorrect in that the start signal only ensures at least ONE SG is available and not TWO. The distractor is plausible as Unit 1 requires two SG for SGTR recovery in accordance with the FSAR.

Technical Reference(s) Technical Specification LCO 3.3.2 Bases Attached w/ Revision: See IPO-003B, Step 4.1.17 Comments / Reference Proposed references to be provided during examination: None Learning Objective: APPLY the administrative requirements of the Auxiliary Feedwater System including Technical Specifications, TRM and ODCM.

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

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 2 Page 23 of 27 CPNPP 2014 NRC SRO Written Exam Worksheet 76 to 80

ES-401 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

IPO-003B, Step 4.1.17 Revision: 19 Page 24 of 27 CPNPP 2014 NRC SRO Written Exam Worksheet 76 to 80

ES-401 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

Technical Specification LCO 3.3.2 Bases Revision: 68 Page 25 of 27 CPNPP 2014 NRC SRO Written Exam Worksheet 76 to 80

ES-401 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

Technical Specification 3.3.2 Bases Revision: 68 Page 26 of 27 CPNPP 2014 NRC SRO Written Exam Worksheet 76 to 80

ES-401 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

Technical Specification 3.3.2 Bases Revision: 68 Page 27 of 27 CPNPP 2014 NRC SRO Written Exam Worksheet 76 to 80

ES-401 SRO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. Date: 05/19/2014 Tier 1 Change: 5 Group 1 K/A 057 G.2.2.22 Level of Difficulty: 3 Importance Rating 4.8 Loss of Vital AC Electrical Instrument Bus: Equipment control: Knowledge of limiting conditions for operations and safety limits.

Proposed Question: 81 Given the following conditions:

Unit 1 is at 100% power.

IV1EC1, 118 VAC SAFEGUARDS BALANCE OF PLANT INVERTER has failed.

In accordance with ABN-603, Loss of Protection or Instrument Bus, panel 1EC1, 118 VAC Distribution Panel 1EC1 was transferred to the alternate source.

Which of the following describes the correct action required to restore panel 1EC1 to full compliance with Technical Specification 3.8.9, Distribution Systems--Operating?

A. Energize Inverter IV1EC1/3 from Battery 1ED1, transfer Panel 1EC1 to the preferred source and place the inverter in INVERTER TO LOAD.

B. Energize Inverter IV1EC1/3 from Battery 1ED3, transfer Panel 1EC1 to the preferred source and place the inverter in BYPASS SOURCE TO LOAD.

C. Energize Inverter IV1EC1/3 from Battery 1ED1, transfer Panel 1EC1 to the preferred source and place the inverter in BYPASS SOURCE TO LOAD.

D. Energize Inverter IV1EC1/3 from Battery 1ED3, transfer Panel 1EC1 to the preferred source and place the inverter in INVERTER TO LOAD.

Proposed Answer: A Page 1 of 45 CPNPP 2014 NRC SRO Written Exam Worksheet 81 to 90

ES-401 SRO Written Exam Worksheet Form ES-401-5 Explanation:

A. Correct. T.S. 3.8.7 requires that an operable inverter is powered from the associated Battery and supplying the vital bus. INVERTER TO LOAD indicated that inverted DC from the safeguards battery is supplying the AC vital bus.

B. Incorrect. Plausible because T.S. 3.8.7 requires that an operable inverter is powered from the associated Battery and supplying the inverter. INVERTER TO LOAD indicated that inverted DC from the safeguards battery is supplying the AC vital bus. With 1ED3 being provided as the appropriate bus and BYPASS SOURCE to load powering the bus, the bus remains inoperable due to both the source and the bypass source to load on the inverter.

C. Incorrect. Plausible because T.S. 3.8.7 requires that an Operable inverter is powered from the associated Battery and supplying the inverter. INVERTER TO LOAD indicates that inverted DC from the safeguards battery is supplying the AC vital bus. While the DC source is 1ED1, the inverter being in BYPASS SOURCE TO LOAD makes the associated vital bus inoperable since it is being supplied from bypass power.

D. Incorrect. Plausible because T.S. 3.8.7 requires that an operable inverter is powered from the associated Battery and supplying the vital bus. INVERTER TO LOAD indicated that inverted DC from the safeguards battery is supplying the AC vital bus. With 1ED3 being provided as the inappropriate DC bus, with INVERTER TO LOAD, the bus remains inoperable due to the inappropriate DC source.

Technical Reference(s) Technical Specification Bases LCO 3.8.7 Attached w/ Revision: See Comments / Reference Proposed references to be provided during examination: None Learning Objective: ANALYZE the indications and DESCRIBE the mitigation strategy and major steps taken, both initial and subsequent, for ABN-604, Loss of Non-1E Instrument Bus.

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 Page 2 of 45 CPNPP 2014 NRC SRO Written Exam Worksheet 81 to 90

ES-401 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

Technical Specifications Bases 3.8.7 Table B Revision: 68 3.8.7-1 Page 3 of 45 CPNPP 2014 NRC SRO Written Exam Worksheet 81 to 90

ES-401 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

Technical Specifications Bases 3.8.7 Table B Revision: 68 3.8.7-1 Page 4 of 45 CPNPP 2014 NRC SRO Written Exam Worksheet 81 to 90

ES-401 SRO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. Date: 05/19/2013 Tier 1 Change: 4 Group 2 K/A W/E02.EA2.1 Level of Difficulty: 4 Importance Rating 4.2 SI Termination: Ability to determine and interpret the following as they apply to (SI Termination): Facility conditions and selection of appropriate procedures during abnormal and emergency operations.

Proposed Question: 82 Given the following conditions:

EOS-1.1A, Safety Injection Termination is in progress on Unit 1.

Centrifugal Charging Pump (CCP) 1-02 has been stopped and placed in standby.

While checking RCS pressure at Step 8 of EOS-1.1A, the following conditions are observed:

RCS pressure is 1400 psig and lowering.

RCS Subcooling is 22°F and becoming less subcooled.

Pressurizer Level is 25% and lowering.

Containment pressure is 1.0 psig and rising.

Safety Injection Pump (SIP) 1-01 and SIP 1-02 are both running.

NO Steam Generators are faulted or ruptured.

CCP 1-01 is aligned via the injection line flow path.

Which of the following actions should be taken by the Unit Supervisor, in accordance with EOS-1.1A?

A. Continue with EOS-1.1A, Safety Injection Termination and re-start CCP 1-02.

B. Transition to EOS-1.2A, Post LOCA Cooldown and Depressurization.

C. Continue with EOS-1.1A, Safety Injection Termination and establish charging flow.

D. Transition to EOP-1.0A, Loss of Reactor or Secondary Coolant.

Proposed Answer: B Page 5 of 45 CPNPP 2014 NRC SRO Written Exam Worksheet 81 to 90

ES-401 SRO Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible because it could be thought that transition to EOP-1.0A does not apply until SI is terminated so re-starting CCP 1-02 would be correct since it was just stopped.

B. Correct. In accordance with step 8 of EOS-1.1A, Safety Injection Termination, if RCS pressure is NOT stable or rising, the operator is directed to EOS-1.2A, Post LOCA Cooldown, for additional actions.

C. Incorrect. Plausible because it could be thought that there is no reason to leave EOS-1.1A, SI Termination and the lowering pressure and level might be corrected by adjusting charging. This is incorrect because Step 8 states that the conditions given require transition to EOS-1.2, Post LOCA Cooldown for additional actions.

D. Incorrect. Plausible because it could be thought that due to subcooling transition to EOP-1.0A is required per the Foldout Page, however the transition to EOP-1.0A only applies after SI is terminated at Step 12..

Technical Reference(s) EOS-1.1A, Rev.8, Steps 7, 8 & 10 Attached w/ Revision: See EOS-1.1A, Rev.8, Foldout Page Comments / Reference Proposed references to be provided during examination: None Learning Objective:

Question Source: Bank ILOT7307 Modified Bank (Note changes or attach parent)

New Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 5 Page 6 of 45 CPNPP 2014 NRC SRO Written Exam Worksheet 81 to 90

ES-401 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

EOS-1.1A Step 8 Revision: 8 Page 7 of 45 CPNPP 2014 NRC SRO Written Exam Worksheet 81 to 90

ES-401 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

EOS-1.1A Step 10 Revision: 8 Page 8 of 45 CPNPP 2014 NRC SRO Written Exam Worksheet 81 to 90

ES-401 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

EOS-1.1A, Fold Out Page Step 1 Amendment: 8 Page 9 of 45 CPNPP 2014 NRC SRO Written Exam Worksheet 81 to 90

ES-401 SRO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. Date: 2/27/2014 Tier 1 Change: 2 Group 2 K/A 028 AA2.08 Level of Difficulty: 3 Importance Rating 3.5 Pressurizer Level Control Malfunction: Ability to determine and interpret the following as they apply to the Pressurizer Level Control Malfunctions: PZR level as a function of power level Proposed Question: 83 Given the following conditions:

Unit 1 is at 20% ramping to 100% power.

Rod Control is in MANUAL.

Average TAVE is 563°F.

Pressurizer level is 32%.

Loop 1 TAVE fails to 630°F.

Assuming no operator action is taken, which of the following indicates final Pressurizer level and the applicable Technical Specification LCO?

Pressurizer level is A. 53%. Technical Specification LCO 3.3.1 Condition M for Pressurizer Level.

B. 60%. Technical Specification LCO 3.3.1 Condition M for Pressurizer Level.

C. 53%. Technical Specification LCO 3.3.1 Condition E for OT N16 and OP N16.

D. 60%. Technical Specification LCO 3.3.1 Condition E for OT N16 and OP N16.

Proposed Answer: C Page 10 of 45 CPNPP 2014 NRC SRO Written Exam Worksheet 81 to 90

ES-401 SRO Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible because 53% PRZR level is the correct final PRZR level, however, PRZR level Technical Specification (TS) LCO 3.3.1, Condition M is not entered, LCO 3.3.1 Condition E is entered for the failed TAVE channel.

B. Incorrect. Plausible because it could be thought that PRZR level would fail to maximum level due to a single TAVE channel failing high, however, the PRZR level program uses AVE TAVE, not high TAVE. It could be thought that PRZR level TS requires entry but it does not, TS LCO 3.3.1 Condition E for the failed TAVE channel must be entered.

C. Correct. From 0 to 100% power TAVE rises from 557°F to 585.4°F. PRZR programmed level rises from 25% to 60% from 0 to 100% power due to AVE TAVE input to the PRZR level controller. When Loop 1 TAVE fails to 630°F, AVE TAVE goes to 580°F. 580°F AVE Tave is equivalent to 80% power.

PRZR program level changes 0.35% for every 1% power change. Therefore, 0.35% x 80% + 25%

= 53% PRZR level. TS LCO 3.3.1, Condition E must be entered for the TAVE failure.

D. Incorrect. Plausible because the correct TS condition is 3.3.1 Condition E the correct TS entry, however, final PRZR level will be 53% not 60% because PRZR programmed level uses AVE TAVE not high TAVE for input.

Technical Reference(s) ABN-704, Sections 2.1 & 2.2 Attached w/ Revision: See ABN-704, Attachment 3 Comments / Reference TDM-301, Programmed TAVE Curve Technical Specification Table 3.3.1-1 Proposed references to be provided during examination: None Learning Objective: APPLY the administrative requirements of the Pressurizer Level Control System including Technical Specifications, TRM and ODCM.

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 Page 11 of 45 CPNPP 2014 NRC SRO Written Exam Worksheet 81 to 90

ES-401 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

ABN-704, Section 2.1 Revision: 10 Page 12 of 45 CPNPP 2014 NRC SRO Written Exam Worksheet 81 to 90

ES-401 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

ABN-704, Section 2.2 Revision: 10 Page 13 of 45 CPNPP 2014 NRC SRO Written Exam Worksheet 81 to 90

ES-401 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

ABN-704, Attachment 3 Revision: 10 Page 14 of 45 CPNPP 2014 NRC SRO Written Exam Worksheet 81 to 90

ES-401 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

TDM-301A, Programmed TAVE Curve Revision: 10 Page 15 of 45 CPNPP 2014 NRC SRO Written Exam Worksheet 81 to 90

ES-401 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

Technical Specification Table 3.3.1-1 Amendment: 161 Page 16 of 45 CPNPP 2014 NRC SRO Written Exam Worksheet 81 to 90

ES-401 SRO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. Date: 3/26/2014 Tier 1 Change: 3 Group 2 K/A W/E06 EA2.2 Level of Difficulty: 3 Importance Rating 4.1 Degraded Core Cooling: Ability to determine and interpret the following as they apply to the Degraded Core Cooling:

Adherence to appropriate procedures and operation within the limitations in the facilitys license and amendments to degraded core cooling Proposed Question: 84 Given the following conditions:

Unit 1 is responding to a large break Loss of Coolant Accident (LOCA).

Cold leg recirculation alignment is in progress in accordance with EOS-1.3A, Transfer to Cold Leg Recirculation.

The Reactor Operator is closing 1/1-8812A, RWST TO RHRP 1 SUCT VLV and 1/1-8812B, RWST TO RHRP 2 SUCT VLV.

The Shift Technical Advisor announces that a loss of Critical Safety Function exists on Core Cooling due to Core Exit Thermocouple (CET) readings on six CETs as follows; CET A06 = 763°F CET J12 = 781°F CET N08 = 792°F CET G08 = 794°F CET R10 = 787°F CET L04 = 767°F Applying the Emergency Response Guidelines rules of usage which of the following describes the required procedural actions?

A. Immediately transition to FRC-0.1A, Response to Inadequate Core Cooling and complete alignment of ECCS for cold leg recirculation after completing FRC-0.1A.

B. Immediately transition to FRC-0.2A, Response to Degraded Core Cooling and complete alignment of ECCS for cold leg recirculation after completing FRC-0.2A.

C. Complete alignment of ECCS for cold leg recirculation then transition to FRC-0.1A, Response to Inadequate Core Cooling.

D. Complete alignment of ECCS for cold leg recirculation then transition to FRC-0.2A, Response to Degraded Core Cooling.

Proposed Answer: D Page 17 of 45 CPNPP 2014 NRC SRO Written Exam Worksheet 81 to 90

ES-401 SRO Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible because it could be thought that immediate transition due to inadequate core cooling is required this is partially based on the belief that a RED path exists but CET temperatures must be greater than 1200°F for the RED path.

B. Incorrect. Plausible because it could be thought that immediate transition is required due to the degraded core cooling (FRC-0.2A) but the NOTE prior to Step 1 in EOS-1.3A requires completion of ECCS cold leg Recirc alignment before addressing the FRC status tree.

C. Incorrect. Plausible because alignment of ECCS cold leg Recirc must be completed prior to addressing the FRC status tree but FRC-0.1A requires CET temperatures to be greater than 1200°F.

D. Correct. The NOTE prior to Step 1 of EOS-1.3A states that FRGs should not be implemented until after completion of step 3 (ECCS Cold Leg Recirc alignment). FRC-0.2A is the correct transition after aligning ECCS for cold leg Recirc due to CETs greater than 750°F but less than 1200°F.

Technical Reference(s) EOS-1.3A, Step 1 NOTE Attached w/ Revision: See FRC-0.1A, Core Cooling CSFST Comments / Reference Proposed references to be provided during examination: None Learning Objective:

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:

EOS-1.3A, Step 1 NOTE Revision: 8 Page 18 of 45 CPNPP 2014 NRC SRO Written Exam Worksheet 81 to 90

ES-401 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

FRC-0.1A, Core Cooling CSFST Revision: 8 Page 19 of 45 CPNPP 2014 NRC SRO Written Exam Worksheet 81 to 90

ES-401 SRO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. Date: 3/27/2014 Tier 1 Change: 4 Group 2 K/A 067 AA2.15 Level of Difficulty: 2 Importance Rating 3.9 Plant Fire on Site: Ability to determine and interpret the following as they apply to the Plant Fire on Site: Requirements for establishing a fire watch Proposed Question: 85 Given the following conditions:

Unit 2 is MODE 3 preparing for refueling.

RCS temperature is 410°F with a cooldown in progress.

The Unit 2 Safeguards NEO reports that Door S2-18, to the TDAFWP 2-01 Room is ajar and will NOT latch closed.

Door S2-18 is rated a 3-hour fire door.

Fire detection status on the 790' elevation of the Unit 2 Safeguards Building has been verified OPERABLE.

Which of the following are the MINIMUM required Fire Protection actions and status of the Turbine Driven Auxiliary Feedwater Pump (TDAFWP) operability, in accordance with STA-738, Fire Protection Systems/Equipment Impairments and Technical Specification LCO 3.7.5, Auxiliary Feedwater (AFW) System?

A. Establish a continuous fire watch with backup fire suppression within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

The TDAFWP remains OPERABLE.

B. Establish a continuous fire watch with backup fire suppression within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

The TDAFWP is inoperable.

C. Establish an hourly fire watch patrol of Door S2-18 within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

The TDAFWP remains OPERABLE.

D. Establish an hourly fire watch patrol of Door S2-18 within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

The TDAFWP is inoperable.

Proposed Answer: C Page 20 of 45 CPNPP 2014 NRC SRO Written Exam Worksheet 81 to 90

ES-401 SRO Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible because in accordance with STA-738, Attachment 8.A 7), with Door S2-18 impaired, establish a continuous fire watch on one side of impaired fire rated assembly within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> OR verify operability of fire detection on one side of impaired fire rated assembly and establish an hourly fire watch. The establishment of the continuous fire watch would meet this requirement, however, the question asks for the minimum required and requiring backup fire suppression is not needed for this impairment as it would be for an impairment to a Spray and/or Sprinkler System. TDAFWP remains OPERABLE per TS LCO 3.7.5 due to compensatory actions of STA-738.

B. Incorrect. Plausible because in accordance with STA-738, Attachment 8.A 7), with Door S2-18 impaired, establish a continuous fire watch on one side of impaired fire rated assembly within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> OR verify operability of fire detection on one side of impaired fire rated assembly and establish an hourly fire watch. The establishment of the continuous fire watch would meet this requirement, however, the question asks for the minimum required and requiring backup fire suppression is not needed for this impairment as it would be for an impairment to a Spray and/or Sprinkler System. Additionally, the TDAFWP remains OPERABLE per TS LCO 3.7.5 due to compensatory actions of STA-738.

C. Correct. In accordance with STA-738, Attachment 8.A 7), with Door S2-18 impaired, establish a continuous fire watch on one side of impaired fire rated assembly within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> OR verify operability of fire detection on one side of impaired fire rated assembly and establish an hourly fire watch. As the operability of the fire detection has been verified, the one hour fire watch patrol satisfies the compensatory measure. TDAFWP remains OPERABLE per TS LCO 3.7.5 due to compensatory actions of STA-738.

D. Incorrect. Plausible because in accordance with STA-738, Attachment 8.A 7), with Door S2-18 impaired, establish a continuous fire watch on one side of impaired fire rated assembly within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> OR verify operability of fire detection on one side of impaired fire rated assembly and establish an hourly fire watch. As the operability of the fire detection has been verified, the one hour fire watch patrol satisfies the compensatory measure. However, the TDAFWP remains OPERABLE per TS LCO 3.7.5 due to compensatory actions of STA-738.

Technical Reference(s) STA-738, Attachment 8.A Attached w/ Revision: See Comments / Reference Proposed references to be provided during examination: STA-738, Attachment 8.A Learning Objective: INTERPRET and ENSURE compliance with plant administrative and operational procedures, guidelines, and policies.

Question Source: Bank Modified Bank ILOT8068 (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 Page 21 of 45 CPNPP 2014 NRC SRO Written Exam Worksheet 81 to 90

ES-401 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

STA-738, Attachment 8.A Revision: 6 Page 22 of 45 CPNPP 2014 NRC SRO Written Exam Worksheet 81 to 90

ES-401 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

STA-738, Attachment 8.A Revision: 6 Page 23 of 45 CPNPP 2014 NRC SRO Written Exam Worksheet 81 to 90

ES-401 SRO Written Exam Worksheet Form ES-401-5 Original Question: ILOT8068 Given the following conditions:

Unit 2 is in the process of being shutdown for refueling.

RCS temperature is 410°F with a cooldown in progress.

The Unit 2 Safeguards NEO reports that Door S2-18, to the TDAFWP 2-01 Room is ajar and will NOT latch closed.

Door S2-18 is rated a 3-hour fire door.

Which of the following Fire Protection actions and Technical Specification conditions are applicable?

A. Establish a continuous fire watch on either side of Door S2-18 within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, all three (3)

Auxiliary Feedwater trains remain OPERABLE per TS LCO 3.7.5.

B. Establish a continuous fire watch on either side of Door S2-18 within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, one (1) Auxiliary Feedwater train is INOPERABLE per TS LCO 3.7.5.

C. Verify operability of fire detection on either side of Door S2-18 within 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />, all three (3)

Auxiliary Feedwater trains remain OPERABLE per TS LCO 3.7.5.

D. Verify operability of fire detection on either side of Door S2-18 within 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />, one (1) Auxiliary Feedwater train is INOPERABLE per TS LCO 3.7.5.

Answer: A Page 24 of 45 CPNPP 2014 NRC SRO Written Exam Worksheet 81 to 90

ES-401 SRO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. Date: 2/27/2014 Tier 2 Change: 2 Group 1 K/A 003 A2.05 Level of Difficulty: 2 Importance Rating 2.8 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: Effects of VCT pressure on RCP seal leak off flows Proposed Question: 86 Given the following conditions:

Unit 1 is at 100% power when the following are observed:

1-PI-115, VCT PRESS indicates 68 psig.

1-ALB-6A, Window 1.5 - VCT PRESS HI/LO alarms.

Volume Control Tank (VCT) level is 52% and stable.

Which of the following describes the effect of the rising VCT pressure on Reactor Coolant Pump seal 1 leakoff flow and what procedure is used to MITIGATE the malfunction?

Reactor Coolant Pump 1 seal leakoff flow...

A. ...increases; Refer to ALM-0061A, Alarm Procedure 1-ALB-6A.

B. ...increases; Refer to ABN-105, Chemical and Volume Control System Malfunction.

C. ...decreases; Refer to ALM-0061A, Alarm Procedure 1-ALB-6A.

D. ...decreases; Refer to ABN-105, Chemical and Volume Control System Malfunction.

Proposed Answer: C Explanation:

A. Incorrect. Plausible because seal 2 leakoff would increase and ALM-0061A is the correct procedure.

B. Incorrect. Plausible because seal 2 leakoff would increase, but ABN-105 is not the correct procedure. A misconception could exist that the CVCS Malfunction ABN would provide the necessary procedural guidance.

C. Correct. When VCT pressure increases RCP 1 seal leakoff flow decreases. ALM-0061A directs the operator to SOP-103 which is used to lower VCT pressure.

D. Incorrect. Plausible because seal 1 leakoff will decrease, but ABN-105 is not the correct procedure. A misconception could exist that the CVCS Malfunction ABN would provide the necessary procedural guidance.

Page 25 of 45 CPNPP 2014 NRC SRO Written Exam Worksheet 81 to 90

ES-401 SRO Written Exam Worksheet Form ES-401-5 Technical Reference(s) ALM-0061A, 1-ALB-6A, Window 1.5 Attached w/ Revision: See ABN-105, Section 1.0 Comments / Reference Proposed references to be provided during examination: None Learning Objective: EVALUATE plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrumentation while responding to a Reactor Coolant Pump System malfunction.

Question Source: Bank Modified Bank ILOT5838 (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:

ABN-105, Section 1.0 Revision: 7 Page 26 of 45 CPNPP 2014 NRC SRO Written Exam Worksheet 81 to 90

ES-401 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

ALM-0061A, 1-ALB-6A, Window 1.5 Revision: 7 Page 27 of 45 CPNPP 2014 NRC SRO Written Exam Worksheet 81 to 90

ES-401 SRO Written Exam Worksheet Form ES-401-5 Original Question: CPNPP Exam Bank ILOT5838 Volume Control Tank pressure lowers and ALB-6B, Window 1.5, VCT PRESS HI/LO, alarms.

Which of the following describes the effect of the lowering pressure on Reactor Coolant Pump seal #1 leakoff flow and what procedure should be used to mitigate the malfunction?

A. Seal leakoff flow increases; ALM-0061A, ALARM PROCEDURE ALB-6B, should be used.

B. Seal leakoff flow increases; SOP-108A, REACTOR COOLANT PUMP, should be used.

C. Seal leakoff flow decreases; ALM-0061A, ALARM PROCEDURE ALB-6B, should be used.

D. Seal leakoff flow decreases; SOP-108A, REACTOR COOLANT PUMP, should be used.

Answer: A Page 28 of 45 CPNPP 2014 NRC SRO Written Exam Worksheet 81 to 90

ES-401 SRO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. Date: 02/27/2014 Tier 2 Change: 3 Group 1 K/A 005 G 2.2.25 Level of Difficulty: 2 Importance Rating 4.2 Residual Heat Removal System: Emergency Procedures/Plan: Knowledge of the bases in Technical Specifications for limiting conditions for operations and safety limits Proposed Question: 87 Given the following conditions:

Unit 1 is in MODE 3 with Reactor Startup preparations in progress.

Valve lineups are being performed and it has been determined that 1-HCV-606, RHR HX 1-01 FLO CTRL VLV has a broken instrument air supply line.

Which of the following explains the condition of the Train A Emergency Core Cooling System (ECCS) Safety Function as a result of the Residual Heat Removal (RHR) System malfunction?

Train A ECCS Safety Function is...

A. ...maintained because 1-HCV-606 fails CLOSED on a loss of instrument air.

B. ...lost because 1-HCV-606 fails CLOSED on a loss of instrument air.

C. ...maintained because 1-HCV-606 fails OPEN on a loss of instrument air.

D. ...lost because 1-HCV-606 fails OPEN on a loss of instrument air.

Proposed Answer: C Explanation:

A. Incorrect. Plausible because Train A ECCS Safety Function is maintained, however 1-HCV-606 actually fails open which maintains the Safety Function.

B. Incorrect. Plausible because if 1-HCV-606 did fail closed Train A ECCS Safety Function is lost, however 1-HCV-606 actually fails open which maintains the Safety Function.

C. Correct. Train A ECCS Safety Function is maintained because its flowpath is capable of injecting cooled water from the RWST or containment sump into the RCS.

D. Incorrect. Plausible because 1-HCV-606 fails open, however the Safety Function is maintained.

Technical Reference(s) Technical Specification LCO 3.5.2 Bases Attached w/ Revision: See Comments / Reference Proposed references to be provided during examination: None Page 29 of 45 CPNPP 2014 NRC SRO Written Exam Worksheet 81 to 90

ES-401 SRO Written Exam Worksheet Form ES-401-5 Learning Objective: APPLY the administrative requirements of the Residual Heat Removal System including Technical Specifications, TRM and ODCM.

Question Source: Bank ILOT1877 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 Page 30 of 45 CPNPP 2014 NRC SRO Written Exam Worksheet 81 to 90

ES-401 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

Technical Specification LCO 3.5.2 Bases Revision: 68 Page 31 of 45 CPNPP 2014 NRC SRO Written Exam Worksheet 81 to 90

ES-401 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

Technical Specification LCO 3.5.2 Bases Revision: 68 Page 32 of 45 CPNPP 2014 NRC SRO Written Exam Worksheet 81 to 90

ES-401 SRO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. Date: 2/27/14 Tier 2 Change: 2 Group 1 K/A 008 A2.09 Level of Difficulty: 3 Importance Rating 2.8 Component Cooling Water System: Ability to (a) predict the impacts of the following malfunctions or operations on the CCWS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Results of excessive exit temperature from the letdown cooler, including the temperature effects on ion-exchange resins Proposed Question: 88 Given the following conditions:

Unit 1 is in MODE 1 at 100% power.

The following Annunciators are in alarm:

1-ALB-06A, Window 1.3 - LTDN HX OUT TEMP HI.

1-ALB-06A, Window 2.3 - LTDN HX NORM OUT FLO DIVERT.

1-TK-130, LTDN HX OUT TEMP CTRL was placed in MANUAL and output stopped at 12%.

1-TK-130, LTDN HX OUT TEMP CTRL output CANNOT be adjusted further.

Alarm Response Procedure field actions have not reduced Letdown Heat Exchanger outlet temperature.

1-TI-130, LTDN HX OUT TEMP indicates 137°F.

1-FI-132, LTDN FLO indicates 139 gpm.

1-FI-121A, CHRG FLO indicates 151 gpm.

Which of the following lists the appropriate procedure to CONTROL the condition and the primary operational/component concern?

A. ABN-105, Chemical and Volume Control System Malfunctions.

Damage to CVCS demineralizer resin due to high temperature.

B. ABN-105, Chemical and Volume Control System Malfunctions.

Damage to the Regenerative Heat Exchanger due to tube vibration.

C. ABN-502, Component Cooling Water System Malfunctions.

Damage to CVCS demineralizer resin due to high temperature.

D. ABN-502, Component Cooling Water System Malfunctions.

Damage to the Regenerative Heat Exchanger due to tube vibration.

Proposed Answer: A Page 33 of 45 CPNPP 2014 NRC SRO Written Exam Worksheet 81 to 90

ES-401 SRO Written Exam Worksheet Form ES-401-5 Explanation:

A. Correct. 1/1-TCV-129 automatically diverts to the VCT when 1-TI-130, LTDN HX OUT TEMP reaches 135F. ALM-0061A, Window 1.3 - LTDN HX OUT TEMP HI, directs isolating Letdown which stops temperature induced resin effects and then directs entry into ABN-105 for loss of Letdown.

B. Incorrect. Plausible because 139 gpm letdown is higher than normal but is less than 140 gpm at which point tube vibration becomes a concern. ABN-105 is the correct procedure for controlling the condition as letdown must be isolated.

C. Incorrect. Plausible because 1/1-TCV-129 automatically diverts to the VCT when 1-TI-130 reaches 135F. It could be thought that ABN-502 would correct the temperature of the Letdown stream.

D. Incorrect. Plausible because it could be thought that ABN-502 would correct the temperature of the Letdown stream.

Technical Reference(s) ALM-0061A, 1-ALB-6A, Window 1.3 Attached w/ Revision: See LO21.SYS.CS1, Pages 19 & 63 Comments / Reference Proposed references to be provided during examination: None Learning Objective: EVALUATE plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrumentation while responding to a Component Cooling Water System malfunction.

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

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 5 Page 34 of 45 CPNPP 2014 NRC SRO Written Exam Worksheet 81 to 90

ES-401 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

ALM-0061A, 1-ALB-6A, Window 1.3 Revision: 7 Comments /

Reference:

ALM-0061A, 1-ALB-6A, Window 1.3 Revision: 7 Page 35 of 45 CPNPP 2014 NRC SRO Written Exam Worksheet 81 to 90

ES-401 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

LO21.SYS.CS1, Page 19 & 63 Revision: 04/28/2011 Page 19 of 86 LETDOWN HIGH TEMPERATURE DIVERSION VALVE Letdown High Temperature Diversion Valve, u-TCV-0129 is located downstream of u-PCV-0131, just before the letdown demineralizers, in the upper valve gallery on 842 ft elevation of the auxiliary building. It functions automatically to divert letdown flow to the volume control tank, bypassing the letdown demineralizers, on high temperature conditions to protect the ion exchanger resin.

Page 63 of 86 Letdown components were originally designed for a nominal maximum flowrate of 120 gpm associated with a 45 gpm orifice and a 75 gpm orifice in service. Subsequent analysis was performed to verify that a flowrate of 140 gpm is acceptable for all components. This evaluation of potentially limiting components reveals that demineralizer resin beds can operate at flowrates significantly higher than 140 gpm, although the effectiveness of the demineralizers is reduced at higher flowrates. The potential for channeling of flow through resin beds at high flow rates is avoided by maintaining flow within the conservative limit of 140 gpm.

Procedures caution against exceeding 140 gpm through demineralizers to address the historical concern of resin channeling and also to minimize the likelihood of a resin intrusion event into the reactor coolant system. As the flowrate through a demineralizer increases, the vessel differential pressure increases. The operational limit for vessel differential pressure is 20 psid. If the differential pressure across an ion exchanger vessel is not maintained within limits, the possibility exists for a failure of the resin retention element and subsequent resin loss to the system. Any resin loss from a vessel should be stopped by the reactor coolant filter so that resin does not get transferred into the reactor coolant system.

As letdown flow rate increases, the regenerative heat exchanger tubes experience increased vibration.

This tube vibration is the most limiting factor associated with the maximum allowed letdown flowrate of 140 gpm.

Page 36 of 45 CPNPP 2014 NRC SRO Written Exam Worksheet 81 to 90

ES-401 SRO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. Date: 3/27/2014 Tier 2 Change: 3 Group 1 K/A 026 A2.05 Level of Difficulty: 2 Importance Rating 4.1 Containment Spray System: Ability to (a) predict the impacts of the following malfunctions or operations on the CSS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Failure of chemical addition tanks to inject Proposed Question: 89 Given the following conditions:

Unit 1 is responding to a Large Break Loss of Coolant Accident (LOCA) in accordance with EOP-0.0A, Reactor Trip or Safety Injection.

Containment pressure is 35 psig and lowering.

While performing Step 7 of EOP-0.0A the following is observed:

1-HS-4754, CHEM ADD TK DISCH VLV is CLOSED.

1-HS-4755, CHEM ADD TK DISCH VLV is CLOSED.

1-HS-4752, CHEM ADD TK DISCH VLV is OPEN.

1-HS-4753, CHEM ADD TK DISCH VLV is OPEN.

Containment Spray Chemical Additive Tank level is 93% and stable.

Which of the following describes the expected action at EOP-0.0A, Reactor Trip or Safety Injection, Step 7 and the Technical Specification Bases for this action?

A. Manually OPEN 1-HS-4754 and 1-HS-4755.

Make spray more acidic to enhance iodine absorption and retention.

B. Manually OPEN 1-HS-4754 and 1-HS-4755.

Make spray more alkaline to enhance iodine absorption and retention.

C. Locally OPEN 1-LV-4754 and 1-LV-4755.

Make spray more acidic to enhance iodine absorption and retention.

D. Locally OPEN 1-LV-4754 and 1-LV-4755.

Make spray more alkaline to enhance iodine absorption and retention.

Proposed Answer: B Page 37 of 45 CPNPP 2014 NRC SRO Written Exam Worksheet 81 to 90

ES-401 SRO Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible because the valves should be manually opened however, NaOH will make spray more alkaline not more acidic.

B. Correct. EOP-0.0A, Step 7 RNO says to manually align valves and refer to Attachment 6 as necessary. TS bases for LCO 3.6.7, Spray Additive System states that making the spray more alkaline will enhance iodine absorption and retention.

C. Incorrect. Plausible because the valves should be opened but EOP-0.0A, Step 7 states to manually align valves not locally open valves and adding NaOH will make spray more alkaline not more acidic.

D. Incorrect. Plausible because the valves should be opened but EOP-0.0A, Step 7 states to manually align valves not locally open valves and adding NaOH will make spray more alkaline.

Technical Reference(s) EOP-0.0A, Step 7 & Attachment 6 Attached w/ Revision: See Technical Specification LCO 3.6.7 Bases Comments / Reference Proposed references to be provided during examination: None Learning Objective: COMPREHEND the normal, abnormal and emergency operation of the Containment Spray System.

APPLY the administrative requirements of the Containment Spray System including Technical Specifications, TRM and ODCM.

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

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 2 Page 38 of 45 CPNPP 2014 NRC SRO Written Exam Worksheet 81 to 90

ES-401 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

EOP-0.0A, Step 7 Revision: 8 Page 39 of 45 CPNPP 2014 NRC SRO Written Exam Worksheet 81 to 90

ES-401 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

EOP-0.0A, Attachment 6 Revision: 8 Page 40 of 45 CPNPP 2014 NRC SRO Written Exam Worksheet 81 to 90

ES-401 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

Technical Specification LCO 3.6.7 Bases Revision: 68 Page 41 of 45 CPNPP 2014 NRC SRO Written Exam Worksheet 81 to 90

ES-401 SRO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. Date: 2/27/2014 Tier 2 Change: 2 Group 1 K/A 061 G 2.4.18 Level of Difficulty: 4 Importance Rating 4.0 Auxiliary/Emergency Feedwater System: Emergency Procedures/Plan: Knowledge of the specific bases for EOPs Proposed Question: 90 Given the following conditions:

Unit 2 is responding to a Pressurized Thermal Shock condition in accordance with FRP-0.1B, Response to Imminent Pressurized Thermal Shock Condition.

While checking Reactor Coolant System (RCS) Cold Leg temperatures, all four loops indicate 170°F and lowering.

All four Steam Generators (SG) are faulted.

Which of the following describes the expected action with regard to Auxiliary Feedwater (AFW) flow and bases for the action in accordance with FRP-0.1B?

A. Control AFW flow at 100 gpm to each SG.

Minimize the effects of the RCS cooldown due to secondary depressurization.

B. Control AFW flow at 100 gpm to each SG.

Minimize thermal shock to Steam Generator components due to dryout and subsequent feeding.

C. Control AFW flow at 150-200 gpm to each SG.

Minimize the effects of the RCS cooldown due to secondary depressurization.

D. Control AFW flow at 150-200 gpm to each SG.

Minimize thermal shock to Steam Generator components due to dryout and subsequent feeding.

Proposed Answer: A Page 42 of 45 CPNPP 2014 NRC SRO Written Exam Worksheet 81 to 90

ES-401 SRO Written Exam Worksheet Form ES-401-5 Explanation:

A. Correct. AFW flow is throttled to 100 gpm per SG to minimize the effects of RCS cooldown while the secondary depressurizes.

B. Incorrect. Plausible because 100 gpm is correct flow but SG dryout concern is from ECA-2.1B and is used to control thermal shock to the SG not the reactor vessel.

C. Incorrect. Plausible because it could be thought that 150 to 200 gpm is required for secondary heat sink in FRP-0.1B and that 100 gpm is only applicable to ECA-2.1B. Also, the reason for the action is correct.

D. Incorrect. Plausible because it could be thought that 150 to 200 gpm is required for secondary heat sink in FRP-0.1B and that 100 gpm is only applicable to ECA-2.1B. Also it could be thought that SG dryout is concern in FRP-0.1B based on thermal shock but ECA-2.1B addresses SG thermal shock not reactor vessel thermal shock.

Technical Reference(s) FRP-0.1B, Attachment 4, Step 2 Bases Attached w/ Revision: See ECA-2.1B, Attachment 4, Step 2 CAUTION Comments / Reference Proposed references to be provided during examination: None Learning Objective: DISCUSS the operator actions, including all cautions, notes, RNOs, and bases associated with FRP-0.1, Response to Imminent Pressurized Thermal Shock 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 X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 5 Page 43 of 45 CPNPP 2014 NRC SRO Written Exam Worksheet 81 to 90

ES-401 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

FRP-0.1B, Attachment 4, Step 2 Bases Revision: 8 Page 44 of 45 CPNPP 2014 NRC SRO Written Exam Worksheet 81 to 90

ES-401 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

ECA-2.1B, Attachment 4, Step 2 CAUTION Revision: 8 Page 45 of 45 CPNPP 2014 NRC SRO Written Exam Worksheet 81 to 90

ES-401 SRO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. Date: 5/19/2014 Tier 2 Change: 3 Group 2 K/A 015 A2.01 Level of Difficulty: 3 Importance Rating 3.9 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: Power supply loss or erratic operation Proposed Question: 91 Given the following conditions:

Unit 1 is operating at 100% power when the following alarms are received:

1-ALB-5C, Window 2.5 - 1 OF 4 OT N16 HI.

1-ALB-6D, Window 1.4 - RX > 50% PWR UP DET FLUX DEV HI.

1-ALB-6D, Window 3.4 - PR CHAN DEV.

1-ALB-6D, Window 3.14 - 1 OF 4 OT N16 ROD STOP & TURB RUNBACK.

1-ALB-6D, Window 4.10 - QUADRANT PWR TILT.

Rod Control is in AUTOMATIC.

Control Bank D rods remain at 215 steps.

ABN-703, Power Range Instrumentation Malfunction is in progress.

Which of the following describes the cause of the alarms and the expected response in accordance with ABN-703, Power Range Instrumentation Malfunction?

A Power Range nuclear instrument channel...

A. ...upper detector has failed low.

Place control rods in MANUAL and withdraw rods to restore TAVE - TREF deviation.

B. ...upper detector has failed low.

Verify QPTR is within limit in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

C. ...upper detector has failed high.

Place control rods in MANUAL and withdraw rods to restore TAVE-TREF deviation.

D. ...upper detector has failed high.

Verify QPTR is within limit in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

Proposed Answer: B Page 1 of 53 CPNPP 2014 NRC SRO Written Exam Worksheet 91 to 100

ES-401 SRO Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible because the upper detector has failed low based on control bank D rod position with rod control in automatic. Rods are placed in MANUAL but not to restore TAVE-TREF deviation, there is no deviation because rods do not move when the instrument fails low.

B. Correct. With Control Bank D rods at 215 steps with rod control in automatic indicates the upper detector failed low. Rods are placed in MANUAL but not to restore TAVE-TREF deviation. QPTR surveillance is required 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after the detector failure when above 75% power.

C. Incorrect. Plausible because placing rods in MANUAL and restoring TAVE-TREF deviation would be required if the detector failed high however based on control rod response the detector has failed low.

D. Incorrect. Plausible because verifying QPTR in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is required but the detector has failed low not high based on control rod response.

Technical Reference(s) ABN-703, Section 2.1 & 2.2 Attached w/ Revision: See ABN-703, Steps 2.3.1, 2.3.2, & 2.3.7 Comments / Reference Technical Specification LCO 3.2.4 OPT-302, Step 5.2.3 Proposed references to be provided during examination: None Learning Objective: ANALYZE the response to a Power Range Instrumentation Malfunction in accordance with ABN-703, Power Range Nuclear Instrument Malfunction.

Question Source: Bank Modified Bank ILOT8203 (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 Page 2 of 53 CPNPP 2014 NRC SRO Written Exam Worksheet 91 to 100

ES-401 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

ABN-703, Section 2.1 Revision: 8 Page 3 of 53 CPNPP 2014 NRC SRO Written Exam Worksheet 91 to 100

ES-401 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

ABN-703, Section 2.2 Revision: 8 Page 4 of 53 CPNPP 2014 NRC SRO Written Exam Worksheet 91 to 100

ES-401 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

ABN-703, Step 2.3.1 & 2.3.2 Revision: 8 Page 5 of 53 CPNPP 2014 NRC SRO Written Exam Worksheet 91 to 100

ES-401 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

ABN-703, Step 2.3.7 Revision: 8 Page 6 of 53 CPNPP 2014 NRC SRO Written Exam Worksheet 91 to 100

ES-401 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

Technical Specification LCO 3.2.4 Amendment: 161 Page 7 of 53 CPNPP 2014 NRC SRO Written Exam Worksheet 91 to 100

ES-401 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

OPT-302, Step 5.2.3 Revision: 11 Page 8 of 53 CPNPP 2014 NRC SRO Written Exam Worksheet 91 to 100

ES-401 SRO Written Exam Worksheet Form ES-401-5 Original Question: CPNPP Exam Bank ILOT8203 Given the following conditions with Unit 1 operating at 100% power:

The following annunciators are in alarm:

1-ALB-6D-2.4, RX > 50% PWR LOW PR DET FLUX DEV HI 1-ALB-6D-3.4, PR CHAN DEV 1-ALB-6D-4.10, QUADRANT PWR TILT Rod Control is in AUTOMATIC.

NO other alarms or automatic control actions occurred.

ABN-703, Power Range Instrumentation Malfunction is in progress.

Which of the following describes the cause of the alarms and what action should be taken to mitigate the situation?

A. A Power Range NI Lower Detector has failed low.

Direct a power reduction to < 75% RTP due to QPTR being greater than Technical Specification limit.

B. A Power Range NI Lower Detector has failed high.

Perform the required channel bypasses that will allow the remaining channels to calculate QPTR.

C. A Power Range NI Lower Detector has failed low.

Verify QPTR within limits using the Core Power Distribution Measurement every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

D. A Power Range NI Lower Detector has failed high.

Place Rod Control in MANUAL until the channel is restored.

Answer: C Page 9 of 53 CPNPP 2014 NRC SRO Written Exam Worksheet 91 to 100

ES-401 SRO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. Date: 5/22/2014 Tier 2 Change: 6 Group 2 K/A 035 A2.01 Level of Difficulty: 3 Importance Rating 4.6 Steam Generator System: Ability to (a) predict the impacts of the following malfunctions or operations on the SGS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Faulted or ruptured steam generators Proposed Question: 92 Given the following conditions:

A Steam Generator Tube Rupture (SGTR) has occurred on Unit 2.

Actions have been taken in EOP-3.0B, Steam Generator Tube Rupture, to the point where a recovery procedure will be selected.

Which of the following procedures is the preferred method of conducting a post-SGTR cooldown, and describe the advantage of using this procedure?

A. EOS-3.3B, Post-SGTR Cooldown Using Steam Dump.

Provides fastest means of depressurizing the Reactor Coolant System.

B. EOS-3.3B, Post-SGTR Cooldown Using Steam Dump.

Minimizes radiological release.

C. EOS-3.1B, Post-SGTR Cooldown Using Backfill.

Provides fastest means of depressurizing the Reactor Coolant System.

D. EOS-3.1B, Post-SGTR Cooldown Using Backfill.

Minimizes radiological release.

Proposed Answer: D Page 10 of 53 CPNPP 2014 NRC SRO Written Exam Worksheet 91 to 100

ES-401 SRO Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible because this method is the fastest means for performing the cooldown, however, it is not the preferred method.

B. Incorrect. Plausible because this method is the fastest means for performing the cooldown which may be thought to minimize radiological release. However, the release of radioactive steam to the condenser is not a radiological release minimizing method.

C. Incorrect. Plausible because this is the preferred method for performing the cooldown, but it is not the fastest means of performing the cooldown, thus the reason is incorrect.

D. Correct. In accordance with the bases of EOP-3.0B Step 41, the preferred method which minimizes radiological release is performing a cooldown using backfill.

Technical Reference(s) EOP-3.0B, Step 41 Bases Attached w/ Revision: See Comments / Reference Proposed references to be provided during examination: None Learning Objective: Given a set of plant conditions (or an actual or simulated Control Room status) and a set of Critical Safety Function Status Trees, correctly DETERMINE the status of the Critical Safety functions and IDENTIFY any applicable Functional Restoration Guidelines.

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 11 of 53 CPNPP 2014 NRC SRO Written Exam Worksheet 91 to 100

ES-401 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

EOP-3.0B, Step 41 Bases Revision: 8 Page 12 of 53 CPNPP 2014 NRC SRO Written Exam Worksheet 91 to 100

ES-401 SRO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. Date: 5/19/2014 Tier 2 Change: 4 Group 2 K/A 086 G 2.4.11 Level of Difficulty: 3 Importance Rating 4.2 Fire Protection System: Emergency Procedures/Plan: Knowledge of abnormal condition procedures Proposed Question: 93 Given the following conditions:

Both Units are at 100% power.

A fire is burning in the Unit 1 Cable Spreading Room.

Both Units are shutting down.

The Shift Manager is injured and CANNOT perform his duties.

Who will assume the duties of the Shift Manager and which Unit will control operation of systems and equipment common to both Units?

In accordance with ABN-803A, Response to a Fire in the Control Room or Cable Spreading Room the duties of the Shift Manager will be assumed by the...

A. ...Unit 1 Unit Supervisor, Unit 1 will control common systems/equipment.

B. ...Unit 1 Unit Supervisor, Unit 2 will control common systems/equipment.

C. ...CPC Supervisor, Unit 1 will control common systems/equipment.

D. ...CPC Supervisor, Unit 2 will control common systems/equipment.

Proposed Answer: C Page 13 of 53 CPNPP 2014 NRC SRO Written Exam Worksheet 91 to 100

ES-401 SRO Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible because it could be thought that the Unit 1 US would assume SM duties. Unit 1 will control common systems and equipment.

B. Incorrect. Plausible because it could be thought that the Unit 1 US would assume SM duties and then Unit 2 would control common systems and equipment.

C. Correct. ABN-803A directs the Unit 2 or CPC Supervisor to assume SM duties and also directs Unit 1 to control common systems and equipment.

D. Incorrect. Plausible because it could be thought that because the fire is in the Unit 1 Cable Spreading Room that the CPC Supervisor would assume SM duties and Unit 2 would control common systems and equipment.

Technical Reference(s) ABN-803A, Step 2.3.1 NOTE Attached w/ Revision: See ABN-803A, Step 2.3.6 NOTE Comments / Reference Proposed references to be provided during examination: None Learning Objective: ANALYZE the response to a Fire in the Electrical or Control Building in accordance with ABN-803, Response to A Fire In the Control Room or Cable Spreading Room.

Question Source: Bank ILOT8073 Modified Bank (Note changes or attach parent)

New Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 5 Page 14 of 53 CPNPP 2014 NRC SRO Written Exam Worksheet 91 to 100

ES-401 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

ABN-803A, Step 2.3.1 NOTE Revision: 11 Page 15 of 53 CPNPP 2014 NRC SRO Written Exam Worksheet 91 to 100

ES-401 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

ABN-803A, Step 2.3.6 NOTE Revision: 11 Page 16 of 53 CPNPP 2014 NRC SRO Written Exam Worksheet 91 to 100

ES-401 SRO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. Date: 3/27/2014 Tier 3 Change: 3 Category 1 K/A G 2.1.34 Level of Difficulty: 4 Importance Rating 3.5 Conduct of Operations: Knowledge of primary and secondary plant chemistry limits Proposed Question: 94 Given the following conditions:

Unit 2 is at 100% power.

At 1000 on June 16th, Chemistry reports RCS DOSE EQUIVALENT I-131 sample results for the past 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />s:

0600 - 1.85 µCi/gm 0700 - 12.6 µCi/gm 0800 - 53.2 µCi/gm 0900 - 75.7 µCi/gm Which of the following states the REQUIRED ACTION based on the Chemistry report?

A. Restore RCS DOSE EQUIVALENT I-131 to within limit by 0900 on June 18th or be in HOT STANDBY by 1500 on June 18th and in COLD SHUTDOWN by 2100 on June 19th.

B. Restore RCS DOSE EQUIVALENT I-131 to within limit by 1000 on June 18th or be in HOT STANDBY by 1600 on June 18th and in COLD SHUTDOWN by 2200 on June 19th.

C. Be in HOT STANDBY by 1600 on June 16th and in COLD SHUTDOWN by 2200 on June 17th.

D. Be in HOT STANDBY by 1500 on June 16th and in COLD SHUTDOWN by 2100 on June 17th.

Proposed Answer: D Page 17 of 53 CPNPP 2014 NRC SRO Written Exam Worksheet 91 to 100

ES-401 SRO Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible if thought that 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> are available to restore DEI-131 to within limits before required entry into MODE 3, however because at 0900 DEI-131 is greater than 60µCi/gm MODE 3 must be entered in 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and MODE 5 in 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

B. Incorrect. Plausible if thought that 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> are available to restore DEI-131 to within limits before required entry into MODE 3, however because at 0900 DEI-131 is greater than 60µCi/gm MODE 3 must be entered in 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. Also the start time for exceeding 60µCi/gm is 0900 not 1000 when reported.

C. Incorrect. Plausible because it could be thought that the clock for action starts at 1000 vice 0900.

D. Correct. Based on sample results at 0900 LCO 3.4.16 Condition C is entered and the unit must be in MODE 3 by 1500 on June 16th and in MODE 5 by 2100 on June 17th.

Technical Reference(s) Technical Specification LCO 3.4.16 Attached w/ Revision: See Comments / Reference Proposed references to be provided during examination: Technical Specification LCO 3.4.16 Learning Objective: APPLY the administrative requirements of the Reactor Coolant System including Technical Specifications, TRM and ODCM.

Question Source: Bank ILOT6157 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 Page 18 of 53 CPNPP 2014 NRC SRO Written Exam Worksheet 91 to 100

ES-401 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

Technical Specification LCO 3.4.16 Amendment: 161 Page 19 of 53 CPNPP 2014 NRC SRO Written Exam Worksheet 91 to 100

ES-401 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

Technical Specification LCO 3.4.16 Amendment: 161 Page 20 of 53 CPNPP 2014 NRC SRO Written Exam Worksheet 91 to 100

ES-401 SRO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. Date: 2/27/2014 Tier 3 Change: 2 Category 2 K/A G 2.2.42 Level of Difficulty: 4 Importance Rating 4.6 Equipment Control: Ability to recognize system parameters that are entry-level conditions for Technical Specifications Proposed Question: 95 Given the following conditions:

Unit 1 is at 100% power with the following indications:

Containment narrow range pressure is -0.25 psig.

Containment temperature is 112°F.

Emergency Core Cooling System Accumulator indications are as follows:

1-01 1-02 1-03 1-04 Pressure (psig) 635 636 640 645 Level (%) 60 59 60 62 Boron (ppm) 2399 2426 2431 2416 Refueling Water Storage Tank (RWST) indications are as follows:

Level is 95%.

Temperature is 121°F.

Boron concentration is 2432 ppm.

Based on the above indications what Technical Specification Limiting Condition(s) for Operation must be entered and what action must be taken?

A. LCO 3.6.4, Containment Pressure; restore within limits in 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

LCO 3.6.5, Containment Air Temperature; restore within limits in 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

B. LCO 3.5.1, ECCS Accumulators; restore Accumulator 1-01 boron concentration within limits in 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

LCO 3.6.5, Containment Air Temperature; restore within limits in 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

C. LCO 3.5.4, Refueling Water Storage Tank; restore Refueling Water Storage Tank temperature within limits in 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

LCO 3.5.4, Refueling Water Storage Tank; restore Refueling Water Storage Tank level within limits in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

D. LCO 3.5.1, ECCS Accumulators; restore Accumulator 1-04 pressure and level within limits in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

LCO 3.5.4, Refueling Water Storage Tank; restore Refueling Water Storage Tank temperature within limits in 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

Page 21 of 53 CPNPP 2014 NRC SRO Written Exam Worksheet 91 to 100

ES-401 SRO Written Exam Worksheet Form ES-401-5 Proposed Answer: D Explanation:

A. Incorrect. Plausible because it could be thought that containment pressure of -.0.25 (alarm would be annunciated) psig would require entry into LCO 3.6.4, however pressure must be less than -0.3 psig. Also, it could be thought that containment temperature of 112°F (alarm setpoint) would require entry into LCO 3.6.5; however temperature must be greater than 120°F.

B. Incorrect. Plausible because it could be thought that boron concentration must be greater than 2400 ppm as is the case with the RWST; however boron must be greater than 2300 ppm for the accumulators. Also, it could be thought that containment temperature of 112°F (alarm setpoint) would require entry into LCO 3.6.5; however temperature must be greater than 120°F.

C. Incorrect. Plausible because in accordance with LCO 3.5.4 RWST temperature must be restored in 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. Also, it could be thought that RWST level must be greater than 95% to not enter LCO 3.5.4 on level.

D. Correct. In accordance with LCO 3.5.1; 1-04 accumulator pressure and level must be restored in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Also, in accordance with LCO 3.5.4 RWST temperature must be restored in 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

Technical Reference(s) Technical Specification LCO 3.5.1 Attached w/ Revision: See Technical Specification LCO 3.5.4 Comments / Reference Technical Specification LCO 3.6.4 Technical Specification LCO 3.6.5 OPT-102A-1 Proposed references to be provided during examination: None Learning Objective: APPLY the administrative requirements of the Emergency Core Cooling System including Technical Specifications, TRM and ODCM.

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

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 2 Page 22 of 53 CPNPP 2014 NRC SRO Written Exam Worksheet 91 to 100

ES-401 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

Technical Specification LCO 3.5.1 Amendment: 161 Page 23 of 53 CPNPP 2014 NRC SRO Written Exam Worksheet 91 to 100

ES-401 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

Technical Specification LCO 3.5.4 Amendment: 161 Page 24 of 53 CPNPP 2014 NRC SRO Written Exam Worksheet 91 to 100

ES-401 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

Technical Specification LCO 3.6.4 Amendment: 161 Page 25 of 53 CPNPP 2014 NRC SRO Written Exam Worksheet 91 to 100

ES-401 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

Technical Specification LCO 3.6.5 Amendment: 161 Page 26 of 53 CPNPP 2014 NRC SRO Written Exam Worksheet 91 to 100

ES-401 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

OPT-102A-1 Revision: 37 Page 27 of 53 CPNPP 2014 NRC SRO Written Exam Worksheet 91 to 100

ES-401 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

OPT-102A-1 Revision: 37 Page 28 of 53 CPNPP 2014 NRC SRO Written Exam Worksheet 91 to 100

ES-401 SRO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. Date: 3/27/2014 Tier 3 Change: 3 Category 2 K/A G 2.2.20 Level of Difficulty: 3 Importance Rating 3.8 Equipment Control: Knowledge of the process for managing troubleshooting activities Proposed Question: 96 Given the following condition:

Troubleshooting is in progress under a troubleshooting Work Order for IV1EC1/3, TRN A 118 VAC RPS/SFGD BOP INSTALLED SPARE INVERTER IV1EC1/3.

Which of the following describes the process to complete troubleshooting and repair activities on IV1EC1/3, TRN A 118 VAC RPS/SFGD BOP INSTALLED SPARE INVERTER IV1EC1/3 once it is determined that several inverter internal components must be replaced?

A. The Responsible Work Organization Supervisor should initial and date the Work Order changes and provide verbal authorization to continue the work.

B. The Responsible Work Organization Supervisor should perform a technical review on the Work Order and provide written authorization to continue the work.

C. The existing Work Order must be revised, Operations must re-impact the Work Order and the Shift Manager must authorize continuing the work.

D. The existing Work Order must be edited, Operations re-impact is NOT required and the Shift Manager need NOT authorize continuing the work.

Proposed Answer: C Page 29 of 53 CPNPP 2014 NRC SRO Written Exam Worksheet 91 to 100

ES-401 SRO Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible because the RWO supervisor could authorize an editorial change by initial and date of the change and verbally authorizing work to continue.

B. Incorrect. Plausible because the RWO supervisor normally performs a technical review of work orders but cannot authorize continuing work IAW requirements of STI-606.01 because revision is required.

C. Correct. STI-606-0.1 requires the work order be revised when work order intent changes and the work order must be re-impacted when troubleshooting becomes corrective. The SM must review and authorize work when a troubleshooting work order scope is expanded to include corrective actions.

D. Incorrect. Plausible because an editorial change would not require re-impact or SM authorization.

Technical Reference(s) STI-606.01, Step 6.11.18 Attached w/ Revision: See STA-202, Step 6.10 Comments / Reference Proposed references to be provided during examination: None Learning Objective: INTERPRET and ENSURE compliance with plant administrative and operational procedures, guidelines, and policies.

Question Source: Bank ILOT7240 Modified Bank (Note changes or attach parent)

New Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 3 Page 30 of 53 CPNPP 2014 NRC SRO Written Exam Worksheet 91 to 100

ES-401 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

STI-606.01, Step 6.11.18 Revision: 0 Page 31 of 53 CPNPP 2014 NRC SRO Written Exam Worksheet 91 to 100

ES-401 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

STI-606-01, Step 6.11.18 Revision: 0 Page 32 of 53 CPNPP 2014 NRC SRO Written Exam Worksheet 91 to 100

ES-401 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

STA-202, Step 6.10 Revision: 36 Page 33 of 53 CPNPP 2014 NRC SRO Written Exam Worksheet 91 to 100

ES-401 SRO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. Date: 5/19/2014 Tier 3 Change: 4 Category 3 K/A G 2.3.11 Level of Difficulty: 4 Importance Rating 4.3 Radiation Control: Ability to control radiation releases Proposed Question: 97 Given the following conditions:

Unit 2 is in MODE 1 at 100% power.

ABN-106, High Secondary Activity is in progress.

The unit has been operating with a 50 gpd Steam Generator Tube Leak for the past 8 days.

Chemistry notifies the control room that the Specific Activity of the Condensate system is 0.12 µCi/gm Dose Equivalent I-131.

What procedure should be used for the power reduction and which Technical Specification action is required to limit the potential radiation release?

Reduce power to 20% and trip the reactor in accordance with A. IPO-003B, Power Operations.

Be in MODE 4 in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> in accordance with LCO 3.7.18, Secondary Specific Activity.

B. IPO-003B, Power Operations.

Be in MODE 5 in 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> in accordance with LCO 3.7.18, Secondary Specific Activity.

C. ABN-106, High Secondary Activity.

Be in MODE 4 in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> in accordance with LCO 3.7.18, Secondary Specific Activity.

D. ABN-106, High Secondary Activity.

Be in MODE 5 in 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> in accordance with LCO 3.7.18, Secondary Specific Activity.

Proposed Answer: B Page 34 of 53 CPNPP 2014 NRC SRO Written Exam Worksheet 91 to 100

ES-401 SRO Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible because shutdown of the plant using IPO-003B will be followed such that a power ramp to 20% is performed and the reactor tripped in order to 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 />.

It could be thought that LCO 3.7.18 only is applicable in MODES 1, 2 & 3 which makes the MODE 4 in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> plausible. Also the 50 gpd tube leak does not require a plant shutdown.

B. Correct. ABN-106, Section 2 does not require a plant shutdown based on the 50 gpd tube leak; however TS LCO 3.7.18 does require the plant to be in MODE 3 in 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and MODE 5 in 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. To shutdown the plant IPO-003B will be followed such that a power ramp to 20% is performed and the reactor tripped in order to 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 />. Then the unit would be placed in MODE 5 within the next 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

C. Incorrect. Plausible because it could be thought that ABN-106 would require a plant shutdown based on the 50 gpd tube leak however the plant shutdown would be accomplished with IPO-003B and is required by LCO 3.7.18 not ABN-106. It could be thought that LCO 3.7.18 only is applicable in MODES 1, 2 & 3 which makes the MODE 4 in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> plausible.

D. Incorrect. Plausible because it could be thought that ABN-106 would require a plant shutdown based on the 50 gpd tube leak however the plant shutdown would be accomplished with IPO-003B and is required by LCO 3.7.18 not ABN-106. LCO 3.7.18 requires the unit to 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 /> and then in MODE 5 within the next 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

Technical Reference(s) ABN-106, Steps 2.3.12 & 2.3.14 Attached w/ Revision: See Technical Specification LCO 3.7.18 & Bases Comments / Reference Proposed references to be provided during examination: None Learning Objective: APPLY the administrative requirements of the Liquid Waste systems including Technical Specifications, TRM and ODCM.

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

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 4 Page 35 of 53 CPNPP 2014 NRC SRO Written Exam Worksheet 91 to 100

ES-401 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

ABN-106, Step 2.3.12 Revision: 10 Page 36 of 53 CPNPP 2014 NRC SRO Written Exam Worksheet 91 to 100

ES-401 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

ABN-106, Step 2.3.14 Revision: 10 Page 37 of 53 CPNPP 2014 NRC SRO Written Exam Worksheet 91 to 100

ES-401 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

Technical Specification LCO 3.7.18 Amendment: 161 Page 38 of 53 CPNPP 2014 NRC SRO Written Exam Worksheet 91 to 100

ES-401 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

Technical Specification LCO 3.7.18 Bases Revision: 68 Page 39 of 53 CPNPP 2014 NRC SRO Written Exam Worksheet 91 to 100

ES-401 SRO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. Date: 5/19/2014 Tier 3 Change: 4 Category 3 K/A G 2.3.15 Level of Difficulty: 3 Importance Rating 3.7 Radiation Control: Knowledge of radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc.

Proposed Question: 98 Given the following conditions:

Both Units are in MODE 1.

The following process radiation monitor status is observed; X-RE-5895A, CR VENT N. INTK (CRV053) status is BLUE.

X-RE-5895B, CR VENT N. INTK (CRV054) status is GREEN.

X-RE-5896A, CR VENT S. INTK (CRV091) status is GREEN.

X-RE-5896B, CR VENT S. INTK (CRV092) status is GREEN.

Which of the following identifies the REQUIRED Technical Specification actions in accordance with LCO 3.3.7, Control Room Emergency Filtration/Pressurization System (CREFS) Actuation Instrumentation?

Place affected CREFS train in emergency recirculation...

A. ...immediately AND secure the Control Room makeup air supply fan from the North Air Intake immediately.

B. ...immediately OR secure the Control Room makeup air supply fan from the North Air Intake immediately.

C. ...within 7 days AND secure the Control Room makeup air supply fan from the North Air Intake within 7 days.

D. ...within 7 days OR secure the Control Room makeup air supply fan from the North Air Intake within 7 days.

Proposed Answer: D Page 40 of 53 CPNPP 2014 NRC SRO Written Exam Worksheet 91 to 100

ES-401 SRO Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible because it could be thought that the OR statement from LCO 3.3.7 action A is an AND statement not an OR statement and that the action time is immediate as in LCO 3.3.7, Condition B.

B. Incorrect. Plausible because it could be thought that the action time is immediate as in LCO 3.3.7, Condition B. The OR statement is correct the immediate action time is not correct.

C. Incorrect. Plausible because it could be thought that the OR statement from LCO 3.3.7 action A is an AND statement not an OR statement.

D. Correct. With one air intake radiation monitor inoperable, TS LCO 3.3.7, CREFS Actuation Instrumentation Condition A requires either placing the affected train of CREFS in emergency recirculation within 7 days or securing the CR makeup supply fan from the affected air intake within 7 days.

Technical Reference(s) Technical Specification LCO 3.3.7 Attached w/ Revision: See Technical Specification Table 3.3.7-1 Comments / Reference Proposed references to be provided during examination: None Learning Objective: LIST and DESCRIBE the following Technical Specifications (i.e., LCOs, action statements and conditional surveillance requirements of one hour and less, if applicable) for the Control Room Ventilation 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 4 Page 41 of 53 CPNPP 2014 NRC SRO Written Exam Worksheet 91 to 100

ES-401 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

Technical Specification LCO 3.3.7 Amendment: 161 Page 42 of 53 CPNPP 2014 NRC SRO Written Exam Worksheet 91 to 100

ES-401 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

Technical Specification Table 3.3.7-1 Amendment: 161 Page 43 of 53 CPNPP 2014 NRC SRO Written Exam Worksheet 91 to 100

ES-401 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

Technical Specification LCO 3.3.7 Amendment: 161 Page 44 of 53 CPNPP 2014 NRC SRO Written Exam Worksheet 91 to 100

ES-401 SRO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. Date: 03/03/2014 Tier 3 Change: 2 Category 4 K/A G 2.4.26 Level of Difficulty: 3 Importance Rating 3.6 Emergency Procedures/Plan: Knowledge of facility protection requirements, including fire brigade and portable firefighting equipment usage Proposed Question: 99 Given the following conditions:

At 1500 hours0.0174 days <br />0.417 hours <br />0.00248 weeks <br />5.7075e-4 months <br />, a Nuclear Equipment Operator (NEO) designated as Nozzleman 1 on the watchbill has been notified that his wife has gone to the hospital in labor.

The Shift Manager has decided to release the NEO to go to the hospital.

Which of the following lists the action required with regard to the Fire Brigade in accordance with ODA-102, Conduct of Operations and STA-727, Fire Brigade?

Release the NEO from the Fire Brigade and immediately...

A. ...take action such that the Nozzleman 1 position will be filled by 1700 with a qualified Nozzleman.

B. ...assign the Safe Shutdown 1 NEO as Nozzleman 1 and take action to ensure relief by 1700 with a qualified Nozzleman.

C. ...assign an NEO to proceed to the scene of the fire if one was to occur in lieu of the unavailable Nozzleman 1.

D. ...assign Hoseman 1 as Nozzleman 1 and fill the Hoseman 1 position with another Prompt Team Hoseman.

Proposed Answer: A Page 45 of 53 CPNPP 2014 NRC SRO Written Exam Worksheet 91 to 100

ES-401 SRO Written Exam Worksheet Form ES-401-5 Explanation:

A. Correct. Fire Brigade manning may be reduced by 1 for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> excluding shift turnover if immediate action is taken to fill the position and the position is filled within 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 it could be thought that another shift NEO could fill the position as long as the position is filled within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, however, the safe shutdown NEOs are specifically excluded from filling Fire Brigade positions.

C. Incorrect. Plausible because in accordance with STA-727 Step 6.3.2.1 Note the task of one nozzleman to proceed to the scene of the fire to perform a preliminary estimate of the size and type of fire and report the status to the Fire Brigade Leader may be designated. However, this does not release the Shift Manager from restoring the minimum Fire Brigade composition within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

D. Incorrect. Plausible because it could be thought the plant knowledge requirement can be relaxed to allow filling the position; however, an Operations advisor would have to be assigned to allow the use of Hoseman as a Nozzleman.

Technical Reference(s) ODA-102, Attachment 8.A Attached w/ Revision: See STA-727, Steps 6.1.2.1, 6.2.2 & 6.3.2.1 Comments / Reference Proposed references to be provided during examination: None Learning Objective: INTERPRET and ENSURE compliance with plant administrative and operational procedures, guidelines, and policies.

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

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 1 Page 46 of 53 CPNPP 2014 NRC SRO Written Exam Worksheet 91 to 100

ES-401 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

ODA-102, Attachment 8.A Revision: 26 Comments /

Reference:

STA-727, Step 6.1.2.1 Revision: 5 Page 47 of 53 CPNPP 2014 NRC SRO Written Exam Worksheet 91 to 100

ES-401 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

STA-727, Step 6.2.2.2 Revision: 5 Page 48 of 53 CPNPP 2014 NRC SRO Written Exam Worksheet 91 to 100

ES-401 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

STA-727, Step 6.2.2.5 Revision: 5 Page 49 of 53 CPNPP 2014 NRC SRO Written Exam Worksheet 91 to 100

ES-401 SRO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. Date: 03/03/2014 Tier 3 Change: 2 Category 4 K/A G 2.4.16 Level of Difficulty: 2 Importance Rating 4.4 Emergency Procedures/Plan: Knowledge of EOP implementation hierarchy and coordination with other support procedures or guidelines such as operating procedures, abnormal operating procedures, and severe accident management guidelines Proposed Question: 100 Given the following conditions:

A Station Blackout has been in progress for several hours.

Unit 1 is responding to the Station Blackout in accordance with ECA-0.0A, Loss of All AC Power.

While checking Core Exit Thermocouple (CET) temperatures they are found to be 1220°F and rising.

Which of the following is the required action?

The Unit Supervisor should...

A. ...remain in ECA-0.0A, Loss of All AC Power and ensure actions required to restore power to any AC Safeguards bus are in progress.

B. ...transition to SACRG-1, Severe Accident Control Room Guideline Initial Response and verify a GENERAL EMERGENCY has been declared.

C. ...enter ABN-601, Response to a 138/345 KV System Malfunction concurrent with ECA-0.0A and restore power to any AC Safeguards bus.

D. ...transition to ECA-0.2A, Loss of All AC Power Recovery With SI Required and manually align SI valves in preparation for power restoration.

Proposed Answer: B Page 50 of 53 CPNPP 2014 NRC SRO Written Exam Worksheet 91 to 100

ES-401 SRO Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible because this action would continuously be performed, however, once CET temperatures exceed 1200°F entry into SACRG-1 is required.

B. Correct. IAW Step 24 of ECA-0.0A when CET temperatures exceed 1200°F entry into SACRG-1 is required.

C. Incorrect. Plausible because this action would be in progress, however, once CET temperatures exceed 1200°F entry into SACRG-1 is required.

D. Incorrect. Plausible because entry into ECA-0.2A would be required because a SI would be needed with CET temperatures greater than 1200°F, however, at least one safeguards bus is needed to enter ECA-0.2A.

Technical Reference(s) ECA-0.0A, Flowchart Attached w/ Revision: See ECA-0.0A, Step 24 Comments / Reference Proposed references to be provided during examination: None Learning Objective: IDENTIFY the proper transitions out of ECA-0.0, Loss of All AC Power.

Question Source: Bank ILOT8323 Modified Bank (Note changes or attach parent)

New Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 5 Page 51 of 53 CPNPP 2014 NRC SRO Written Exam Worksheet 91 to 100

ES-401 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

ECA-0.0A, Flowchart Revision: 8 Page 52 of 53 CPNPP 2014 NRC SRO Written Exam Worksheet 91 to 100

ES-401 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

ECA-0.0A, Step 24 Revision: 8 Page 53 of 53 CPNPP 2014 NRC SRO Written Exam Worksheet 91 to 100

CPNPP NRC 2014 RO Written Exam Reference List

1. NRC Generic Fundamentals Equation Sheet

2. Steam Tables CPNPP NRC 2014 RO Written Exam Reference List

GENERIC FUNDAMENTALS EXAMINATION EQUATIONS AND CONVERSIONS HANDOUT SHEET EQUATIONS



Q  pT mc P = Po10SUR(t)



Q 

mh P = Poe(t/)

 A = Aoe-t Q UAT CRS/D = S/(1 - Keff)

  m Q

3

 Nat Circ CR1(1 - Keff1) = CR2(1 - Keff2) 2 T  m  Nat Circ 1/M = CR1/CRX Keff = 1/(1 - ) A = r 2

= (Keff - 1)/Keff F = PA SUR = 26.06/ m = Av eff W 

mP Pump eff E = IR 5 eff

 Thermal Efficiency = Net Work Out/Energy In 1  eff g(z2 - z1) + (v22 - v12) + (P2 - P1) + (u2 - u1) + (q - w) = 0 5* = 1 x 10-4 sec gc 2gc eff = 0.1 sec-1 (for small positive ) gc = 32.2 lbm-ft/lbf-sec2 2 2 DRW  tip / avg CONVERSIONS 1 Mw = 3.41 x 106 Btu/hr 1 Curie = 3.7 x 1010 dps 1 hp = 2.54 x 103 Btu/hr 1 kg = 2.21 lbm 1 Btu = 778 ft-lbf 1 galwater = 8.35 lbm (C = (5/9)((F - 32) 1 ft3water = 7.48 gal (F = (9/5)((C) + 32 CPNPP NRC 2014 RO Written Exam Reference List

CPNPP NRC 2014 SRO Written Exam Reference List

1. Technical Specification LCO 3.4.16, RCS Specific Activity

2. NRC Generic Fundamentals Equation Sheet

3. STA-738, Fire Protection Systems/Equipment Impairments

4. Steam Tables CPNPP NRC 2014 SRO Written Exam Reference List

RCS Specific Activity 3.4.16 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.16 RCS Specific Activity LCO 3.4.16 RCS DOSE EQUIVALENT I-131 and DOSE EQUIVALENT XE-133 specific activity shall be within limits.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. DOSE EQUIVALENT I-131 -----------------------NOTE------------------------

not within limit. LCO 3.0.4.c is applicable.

A.1 Verify DOSE EQUIVALENT I-131 Once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> 60Ci/gm.

AND A.2 Restore DOSE EQUIVALENT I-131 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> to within limit.

B. DOSE EQUIVALENT B.1 --------------------NOTE---------------------

XE-133 not within limit. LCO 3.0.4.c is applicable.

Restore DOSE EQUIVALENT 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> XE-133 to within limit.

COMANCHE PEAK - UNITS 1 AND 2 3.4-41 Amendment No. 150, 156 CPNPP NRC 2014 SRO Written Exam Reference List

RCS Specific Activity 3.4.16 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C. Required Action and C.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time of Condition A or B AND not met.

C.2 Be in MODE 5. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> OR DOSE EQUIVALENT I-131

> 60 Ci/gm.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.16.1 -----------------------------------NOTE-----------------------------------

Only required to be performed in MODE 1.

Verify reactor coolant DOSE EQUIVALENT XE-133 In accordance with specific activity 500 Ci/gm. the Surveillance Frequency Control Program.

COMANCHE PEAK - UNITS 1 AND 2 3.4-42 Amendment No. 150, 156 CPNPP NRC 2014 SRO Written Exam Reference List

GENERIC FUNDAMENTALS EXAMINATION EQUATIONS AND CONVERSIONS HANDOUT SHEET EQUATIONS



Q  pT mc P = Po10SUR(t)



Q 

mh P = Poe(t/)

 A = Aoe-t Q UAT CRS/D = S/(1 - Keff)

  m Q

3

 Nat Circ CR1(1 - Keff1) = CR2(1 - Keff2) 2 T  m  Nat Circ 1/M = CR1/CRX Keff = 1/(1 - ) A = r 2

= (Keff - 1)/Keff F = PA SUR = 26.06/ m = Av eff W 

mP Pump eff E = IR 5 eff

 Thermal Efficiency = Net Work Out/Energy In 1  eff g(z2 - z1) + (v22 - v12) + (P2 - P1) + (u2 - u1) + (q - w) = 0 5* = 1 x 10-4 sec gc 2gc eff = 0.1 sec-1 (for small positive ) gc = 32.2 lbm-ft/lbf-sec2 2 2 DRW  tip / avg CONVERSIONS 1 Mw = 3.41 x 106 Btu/hr 1 Curie = 3.7 x 1010 dps 1 hp = 2.54 x 103 Btu/hr 1 kg = 2.21 lbm 1 Btu = 778 ft-lbf 1 galwater = 8.35 lbm (C = (5/9)((F - 32) 1 ft3water = 7.48 gal (F = (9/5)((C) + 32 CPNPP NRC 2014 SRO Written Exam Reference List

CPSES PROCEDURE NO.

STATION ADMINISTRATION MANUAL STA-738 FIRE PROTECTION SYSTEMS/EQUIPMENT IMPAIRMENTS REVISION NO. 6 PAGE 13 OF 45 ATTACHMENT 8.A PAGE 1 OF 6 GUIDELINES FOR COMPENSATORY MEASURES THE FOLLOWING COMPENSATORY MEASURES SHALL BE IMPLEMENTED WHEN FIRE PROTECTION SYSTEMS/EQUIPMENT ARE DETERMINED IMPAIRED OR INOPERABLE AS DESCRIBED BY THIS PROCEDURE.

1) FIRE SUPPRESSION WATER SYSTEM a) With one pump and/or one water supply tank inoperable, restore the inoperable equipment to OPERABLE status within 7 days or provide an alternate backup pump or supply.

b) With the fire suppression water system otherwise inoperable, establish a backup fire suppression water supply within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

NOTE: Normally this can be achieved by verifying operability of the emergency refill pump located at the Service Water Intake Structure (SWIS). The valves necessary to achieve the proper alignment should be stroked to verify operability and placed in normal operating position upon completion, to prevent the introduction of lake water into the system.

[C] 2) FIRE DETECTION a) With any, but not more than one-half the total in any fire zone Function A fire detection instruments shown in Attachment 8.B inoperable, the inoperable instrument(s) shall be restored to an operable status within 14 days or within the next 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> a fire watch patrol shall be established to inspect the zone(s) with the inoperable instrument(s) at least once per hour, unless the instrument(s) is located inside Containment or within Zone V radiation areas outside Containment, then that zone shall be inspected at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or monitor the containment air temperature at least once per hour by the Control Room Indicators.

CPNPP NRC 2014 SRO Written Exam Reference List

CPSES PROCEDURE NO.

STATION ADMINISTRATION MANUAL STA-738 FIRE PROTECTION SYSTEMS/EQUIPMENT IMPAIRMENTS REVISION NO. 6 PAGE 14 OF 45 ATTACHMENT 8.A PAGE 2 OF 6 GUIDELINES FOR COMPENSATORY MEASURES

2) b) With more than one-half of the Function A fire detection instruments in any fire zone shown in Attachment 8.B inoperable, or with any Function B fire detection instrument(s) shown in Attachment 8.B inoperable, or with any two or more adjacent fire detection instruments shown in Table 1 inoperable, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> a fire watch patrol shall be established to inspect the zone(s) with inoperable instrument(s) at least once per hour, unless the instrument(s) is located inside Containment or within Zone V radiation areas outside Containment, then that zone shall be inspected at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or, for Containment, monitor the air temperature at least once per hour by the Control Room Indicators.

c) For all other detection systems/equipment compensatory measures should be determined by the Fire Protection Supervisor.

[C] 3) SPRAY AND/OR SPRINKLER SYSTEM a) With the Unit 2 safety chiller room water curtain system inoperable, compensatory measures shall be established depending on the type or extent of impairment.

(1) Fire Barrier and/or Water Curtain Inoperable - Within one hour ensure detection system operability and establish an hourly roving fire watch. If detection is inoperable, establish a continuous fire watch.

(2) Area Wide Suppression System Inoperable - Within one hour establish a continuous fire watch.

(3) Detection System Inoperable - See Attachment 8.A, 2 (a).

(4) Barrier and Suppression System Inoperable - Within one hour establish a continuous fire watch.

CPNPP NRC 2014 SRO Written Exam Reference List

CPSES PROCEDURE NO.

STATION ADMINISTRATION MANUAL STA-738 FIRE PROTECTION SYSTEMS/EQUIPMENT IMPAIRMENTS REVISION NO. 6 PAGE 15 OF 45 ATTACHMENT 8.A PAGE 3 OF 6 GUIDELINES FOR COMPENSATORY MEASURES (5) Suppression and Detection System Inoperable - Within one hour establish a continuous fire watch.

3) b) With one or more of the required spray and/or sprinkler systems in the Diesel Generator Building inoperable, establish an hourly roving fire watch patrol within one hour.

c) With one or more of the required Spray and/or Sprinkler Systems listed in Attachment 8.C inoperable, establish a continuous fire watch with backup fire suppression equipment within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. For Zone V radiation areas, the area shall be inspected at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, with backup fire suppression equipment established within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> for the inoperable system.

d) For Containment, the thermal detectors shall be verified operable in the containment pre-access filtration units.

e) For all other areas, compensatory measures should be determined by the Fire Protection Supervisor.

[C] 4) HALON With a Cable Spreading Room Halon System (which includes both main and reserve cylinders) inoperable, establish a continuous fire watch with back up fire suppression equipment within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

For all other areas, compensatory measures should be determined by the Fire Protection Supervisor.

CPNPP NRC 2014 SRO Written Exam Reference List

CPSES PROCEDURE NO.

STATION ADMINISTRATION MANUAL STA-738 FIRE PROTECTION SYSTEMS/EQUIPMENT IMPAIRMENTS REVISION NO. 6 PAGE 16 OF 45 ATTACHMENT 8.A PAGE 4 OF 6 GUIDELINES FOR COMPENSATORY MEASURES

[C] 5) FIRE HOSE STATIONS With one or more of the fire hose stations listed in Attachment 8.D inoperable, provide a gated wye on the nearest OPERABLE hose station. One outlet of the gated wye shall be connected to the standard length of fire hose provided for the OPERABLE hose station.

The second outlet of the gated wye shall be connected to a length of hose sufficient to provide coverage for the area left unprotected by the inoperable hose station. The action listed above shall be accomplished within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> if the inoperable fire hose station is the primary means of fire suppression; otherwise, additional fire hose shall be provided within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Signs identifying the purpose and location of the fire hose and related valves shall be mounted above the gated wye and at the inoperable hose station. Where a gated wye on the nearest OPERABLE fire hose station cannot be provided, a hose capable of providing an equivalent quantity of water and pressure may be provided. It is also acceptable to provide coverage to the area left unprotected by the inoperable hose station by using a fire hose from a fire hose station not listed in Attachment 8.D, if that hose can provide an equivalent quantity of water and pressure. Where it can be demonstrated that the physical routing of the fire hose from the OPERABLE hose station to the inoperable hose station would result in a recognizable hazard to plant personnel, plant equipment or the fire hose itself, the fire hose shall not be laid out to the inoperable hose station, but stored at the outlet of the OPERABLE fire hose station.

For all other areas, compensatory measures should be determined by the Fire Protection Supervisor.

CPNPP NRC 2014 SRO Written Exam Reference List

CPSES PROCEDURE NO.

STATION ADMINISTRATION MANUAL STA-738 FIRE PROTECTION SYSTEMS/EQUIPMENT IMPAIRMENTS REVISION NO. 6 PAGE 17 OF 45 ATTACHMENT 8.A PAGE 5 OF 6 GUIDELINES FOR COMPENSATORY MEASURES

[C] 6) YARD FIRE HYDRANTS/FIRE HOSE HOUSES With one or more of the yard fire hydrants or associated hydrant hose houses listed in Attachment 8.E inoperable, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> provide sufficient additional lengths of 2-1/2 inch diameter hose located at an adjacent OPERABLE fire hydrant to provide service to the unprotected area(s) if the inoperable fire hydrant or associated hydrant hose house is the primary means of fire suppression; otherwise, provide additional fire hose within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

It is also acceptable to provide suppression coverage to the area left unprotected by an inoperable yard fire hydrant or hydrant hose house by using a yard fire hydrant not listed in Attachment 8.E, if that fire hydrant can provide an equivalent quantity of water and pressure.

[C] 7) FIRE RATED ASSEMBLIES With one or more of the fire rated assemblies (fire dampers, fire walls, fire doors, penetration seals, thermolag and radiant energy shield) listed in Attachment 8.F impaired or inoperable; establish within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> a continuous fire watch on one side of the affected assembly, or verify operability of the fire detection on at least one side of the impaired/inoperable fire rated assembly and establish an hourly fire watch patrol.

For all other rated assemblies (i.e., Risk Management areas), compensatory measures should be determined by the Fire Protection Supervisor.

CPNPP NRC 2014 SRO Written Exam Reference List

CPSES PROCEDURE NO.

STATION ADMINISTRATION MANUAL STA-738 FIRE PROTECTION SYSTEMS/EQUIPMENT IMPAIRMENTS REVISION NO. 6 PAGE 18 OF 45 ATTACHMENT 8.A PAGE 6 OF 6 GUIDELINES FOR COMPENSATORY MEASURES

8) APPLICABILITY To ascertain applicability consult with the Operations Shift Manager and/or the Fire Protection Supervisor.

a) Fire Suppression Water System - AT ALL TIMES.

b) Fire Detection - Whenever the equipment protected is required to be OPERABLE.

c) Spray and/or Sprinkler Systems - Whenever the equipment protected is required to be OPERABLE.

d) Halon Fire Suppression System - Whenever the equipment protected is required to be OPERABLE.

e) Fire Hose Stations - Whenever the equipment protected is required to be OPERABLE.

f) Yard Hydrants/Hydrant Hose Houses - Whenever the equipment protected is required to be OPERABLE.

g) Fire Rated Assemblies - Whenever equipment protected is required to be OPERABLE.

CPNPP NRC 2014 SRO Written Exam Reference List

CPNPP 2014 NRC Written Examination Senior Reactor Operator Answer Key

1. C 26. C 51. A 76. A

2. C 27. A 52. A 77. A

3. C 28. C 53. D 78. B

4. A 29. D 54. A 79. B

5. B 30. B 55. A 80. B

6. B 31. C 56. C 81. A

7. D 32. B 57. A 82. B

8. A 33. C 58. C 83. C

9. B 34. B 59. A 84. D

10. B 35. C 60. A 85. C

11. A 36. D 61. D 86. C

12. D 37. A 62. D 87. C

13. B 38. C 63. D 88. A

14. D 39. C 64. A 89. B

15. B 40. A 65. C 90. A

16. C 41. B 66. C or D 91. B

17. A 42. C 67. C 92. D

18. D 43. B 68. C 93. C

19. A 44. B 69. B 94. D

20. A 45. D 70. D 95. D

21. C 46. B 71. B 96. C

22. C 47. D 72. A 97. B

23. D 48. C 73. D 98. D

24. D 49. C 74. C 99. A

25. C 50. D 75. D 100. B CPNPP NRC 2014 SRO Written Exam Answer Key