ML13108A205

2013-04 Final Written Exam

ML13108A205
Person / Time
Site:	Comanche Peak
Issue date:	04/08/2013
From:	Kelly Clayton Operations Branch IV
To:	Luminant Generation Co
laura hurley
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Download: ML13108A205 (394)

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ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 003 K2.02 Importance Rating 2.5 Reactor Coolant Pump System: Knowledge of bus power supplies to the following: CCW pumps Proposed Question: Common 1 What are the power supplies for Component Cooling Water (CCW) Pumps 2-01 and 2-02?

CCWP 2-01 CCWP 2-02 A. 2EA1 2EB1 B. 2EA1 2EB2 C. 2EB1 2EB2 D. 2EA1 2EA2 Proposed Answer: D Explanation:

A. Incorrect. Plausible if believed that the A and B in the bus designator are related to the Train they supply power from.

B. Incorrect. Plausible if believed that the A and B in the bus designator are related to the Train they supply power from and the 1 and 2 are also part of the Train designator.

C. Incorrect. Plausible if believed that the CCWPs are powered form 480V AC.

D. Correct. 2-EA1 and 2EA2 are the correct power supplies.

Technical Reference(s) LO21.SYS.CC1 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 Component Cooling Water 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 Page 1 of 93 CPNPP NRC 2013 RO Written Exam Worksheet 1 to 25 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comprehension or Analysis 10 CFR Part 55 Content: 55.41 3, 7 55.43 Comments /

Reference:

From LO21.SYS.CC1.LN, Page 14 Revision 5/1/2011 CCW PUMPS The CCW pumps are located on the centerline of the Auxiliary Building, elevation 810'. They are 100% capacity, centrifugal, horizontal, double suction, single stage, motor-driven pumps with a nominal capacity of 14,700 gpm each at a head of 226 ft. The shafts have minimum leakage mechanical seals cooled by the discharge of the pump. The journal and thrust bearings are self-lubricated by oil rings.

The pumps are normally powered from uEA1 and uEA2. On a loss of power, they will be supplied from the train related emergency diesel generator. Control power for the pumps is from uED1-2 for Train A and uED2-2 for Train B.

Page 2 of 93 CPNPP NRC 2013 RO Written Exam Worksheet 1 to 25 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 004 K2.06 Importance Rating 2.6 Chemical and Volume Control System: Knowledge of bus power supplies to the following: Control instrumentation Proposed Question: Common 2 Given the following conditions:

• 1-PT-0456, PRESSURIZER 1-01 PRESSURE TRANSMITTER 0456 PROT CHAN II has failed.

• Maintenance was dispatched and reported that the transmitter had no power.

What is the power supply for 1-PT-0456, PRESSURIZER 1-01 PRESSURE TRANSMITTER 0456 PROT CHAN II?

A. 1PC1, 118 VAC DISTRIBUTION PANEL 1PC1 B. 1EC1, 118 VAC DISTRIBUTION PANEL 1EC1 C. 1PC2, 118 VAC DISTRIBUTION PANEL 1PC2 D. 1EC2, 118 VAC DISTRIBUTION PANEL 1EC2 Proposed Answer: C Explanation:

A. Incorrect. Plausible if thought that because the channel is a controlling channel it is powered from channel I, however PT-456 is a channel II instrument.

B. Incorrect. Plausible if thought that because the channel is a controlling channel it is powered from channel I, however PT-456 is a channel II instrument that it is powered from an instrument bus vice a protection bus.

C. Correct. The power supply for PT-456 is 1PC2.

D. Incorrect. Plausible if thought that PT-456 is powered from an instrument bus vice a protection bus.

Technical Reference(s) ABN-603, Attachment 1 Attached w/ Revision # See Comments / Reference Proposed references to be provided during examination: None Learning Objective: DEMONSTRATE an understanding of the components of the Chemical and Volume Control System including interrelations with other systems to include interlocks and control loops.

Page 3 of 93 CPNPP NRC 2013 RO Written Exam Worksheet 1 to 25 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Question Source: Bank #

Modified Bank # (Note changes or attach parent)

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

Reference:

From ABN-603, Attachment 1 Revision 8, PCN 1 Page 4 of 93 CPNPP NRC 2013 RO Written Exam Worksheet 1 to 25 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Residual Heat Removal System: Ability to (a) predict the impacts of the following malfunctions or operations on the RHRS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: RHR pump/motor malfunction Proposed Question: Common 3 Given the following conditions:

• Unit 2 is in a Mid-Loop condition when the following alarms are received:

• 2-ALB-4B, Window 2.4 - RHRP 1/2 OVRLOAD TRIP.

• 2-ALB-4B, Window 4.4 - RHRP 1/2 TO CL INJ FLO LO.

• The running Train A Residual Heat Removal (RHR) Pump has tripped.

• The standby Train B RHR Pump will NOT start.

• The Reactor Vessel Head is removed.

Which of the following actions should be performed per ABN-104, Residual Heat Removal System Malfunction?

Initially attempt to...

A. ...align the Refueling Water Storage Tank for gravity feed.

B. ...initiate Hot Leg injection with a Safety Injection Pump.

C. ...initiate Cold Leg injection with a Centrifugal Charging Pump.

D. ...align the Volume Control Tank for gravity feed.

Proposed Answer: B Explanation:

A. Incorrect. Plausible because RWST gravity feed is a subsequent action if Hot Leg and Cold Leg injection fail.

B. Correct. Given the conditions listed, this is the correct action per ABN-104.

C. Incorrect. Plausible because Cold Leg injection is a subsequent action if Hot Leg injection fails.

D. Incorrect. Plausible because VCT gravity feed is a subsequent action if Hot Leg and Cold Leg injection fail.

Technical Reference(s) ABN-104, Section 8.1 Attached w/ Revision # See ABN-104, Steps 8.3.9, 8.3.10, & 8.3.11 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.

Page 5 of 93 CPNPP NRC 2013 RO Written Exam Worksheet 1 to 25 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Question Source: Bank #

Modified Bank # X (Note changes or attach parent)

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

Reference:

From ABN-104, Section 8.1 Revision # 8 Page 6 of 93 CPNPP NRC 2013 RO Written Exam Worksheet 1 to 25 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ABN-104, Step 8.3.9 Revision # 8 Page 7 of 93 CPNPP NRC 2013 RO Written Exam Worksheet 1 to 25 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ABN-104, Step 8.3.9 Revision # 8 Page 8 of 93 CPNPP NRC 2013 RO Written Exam Worksheet 1 to 25 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ABN-104, Step 8.3.10 Revision # 8 Page 9 of 93 CPNPP NRC 2013 RO Written Exam Worksheet 1 to 25 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ABN-104, Step 8.3.11 Revision # 8 Page 10 of 93 CPNPP NRC 2013 RO Written Exam Worksheet 1 to 25 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From CPNPP Exam Bank Revision # 10/06/97 Given the following conditions:

• During Mid-Loop operations on unit 2, the following alarms are received:

• 2-ALB-4B, Window 2.4 - RHRP 1/2 OVRLOAD TRIP.

• 2-ALB-4B, Window 4.4 - RHRP 1/2 TO CL INJ FLO LO.

• The Reactor Operator determines the running RHR Pump (Train A) has tripped and is unable to start the Train B RHR Pump.

Which of the following actions should be performed per ABN-104, Residual Heat Removal System Malfunction?

A. Isolate the RHR Hot Leg suctions, and initiate Hot Leg Injection with an SIP.

B. Ensure a Hot Leg vent path, and initiate Hot Leg injection with an SIP.

C. Open PORV's, initiate Cold Leg injection with an SIP.

D. Open PORV's, initiate Hot Leg injection with a CCP.

Page 11 of 93 CPNPP NRC 2013 RO Written Exam Worksheet 1 to 25 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 005 K6.03 Importance Rating 2.5 Residual Heat Removal System: Knowledge of the effect that a loss or malfunction of the following will have on the RHRS:

RHR heat exchanger Proposed Question: Common 4 Given the following conditions:

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

• All protection systems responded as expected during the initial phase of the accident.

• Containment pressure rose to 43 psig before being lowered by Containment Spray.

• The crew has placed the unit in hot leg recirculation per EOS-1.4A, Transfer to Hot Leg Recirculation.

• 1-HCV-607, RHR HX 1-02 FLO CTRL VLV has closed and CANNOT be reopened.

Which of the following describes the impact on the Emergency Core Cooling System and design analysis assumptions with regard to the Emergency Core Cooling System?

Overall cooling to the core is A. unchanged because only one train of Emergency Core Cooling is assumed available.

B. reduced but design analysis assumptions remain valid.

C. unchanged because only one train of Emergency Core Cooling is used during hot leg recirculation.

D. reduced and design analysis assumptions are NO longer valid.

Proposed Answer: B Explanation:

A. Incorrect. Plausible if thought that loss of a single train of RHR is assumed in the design analysis and that that therefore does not impact cooling.

B. Correct. Core cooling is reduced but design analysis assumptions remain valid because loss of a single train is covered in the analysis.

C. Incorrect. Plausible if thought that a single train is used for hot leg Recirc due to how the system is configured (only hot legs 2 & 3 are injected into by RHR in hot leg Recirc).

D. Incorrect. Plausible because overall core cooling is reduced but the design assumptions account for the loss of a single train of cooling.

Page 12 of 93 CPNPP NRC 2013 RO Written Exam Worksheet 1 to 25 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Technical Reference(s) LO21.SYS.SI1 Attached w/ Revision # See LO21.SYS.RH1 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 #

Modified Bank # (Note changes or attach parent)

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

Reference:

From LO21.SYS.SI1, PAGES 7 & 8 Revision 5/2/2011 The Emergency Core Cooling System is designed to remain functional after a Safe Shutdown Earthquake. The system is also protected from flooding, pipe whips, jet forces, and missiles. The system is designed to tolerate a single failure without a loss of its core protective functions. This failure is limited to an active failure during the short-term (injection) phase following a LOCA. When applied to the Emergency Core Cooling System, an active failure is defined as the failure of a powered component, such as a pump, power actuated valve, a component of the electrical supply system or instrumentation and control equipment, to act to accomplish its design functions. The worst-case single active failure would be the loss of one entire safeguards electrical train. The single failure is also limited to an active or passive failure during the long-term (recirculation) cooling phase. A passive failure is defined as the structural failure of a static component. When applied to a fluid system, this means a break in the pressure boundary resulting in abnormal leakage not exceeding 50 gpm for 30 minutes. Such leaks are consistent with limited cracks in pipes, sprung flanges, valve packing leaks or pump seal failures.

Page 13 of 93 CPNPP NRC 2013 RO Written Exam Worksheet 1 to 25 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From LO21.SYS.RH1, Page 25 Revision 10/20/2011 The rate of heat removal is controlled manually by the operator by adjusting u-HC-606 (Train A) or u-HC-607 (Train B) to position the Flow Control Valves. The RHR Heat Exchanger Bypass Valves are operated in automatic control. As the operator manually positions the RHR Heat Exchanger Flow Control Valve to control Reactor Coolant temperature or cooldown rate, the RHR Heat Exchanger Bypass Valve repositions to maintain constant flow through the system Page 14 of 93 CPNPP NRC 2013 RO Written Exam Worksheet 1 to 25 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 006 A1.16 Importance Rating 4.1 Emergency Core Cooling System: Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the ECCS controls including: RCS temperature, including superheat, saturation and subcooled Proposed Question: Common 5 Given the following conditions:

• Unit 1 is responding to a Small Break Loss of Coolant Accident per EOS-1.2A, Post LOCA Cooldown and Depressurization.

• All Reactor Coolant Pumps have been stopped.

• Residual Heat Removal Pumps are in STANDBY.

• Safety Injection Pumps are in STANDBY.

• Normal charging flow has been established.

• Containment pressure is 6 psig and stable.

• The following indications are observed;

• Reactor Coolant System Hot Leg temperatures are all 460°F and rising.

• Reactor Coolant System pressure is 835 psig and lowering.

• Subcooling is 65°F and becoming less subcooled.

• Pressurizer level is 38% and slowly lowering.

Which of the following actions should be taken in response to the given conditions per EOS-1.2A, Post LOCA Cooldown and Depressurization?

A. Immediately manually start and align Emergency Core Cooling System pumps.

B. Immediately re-actuate Safety Injection to align Emergency Core Cooling System.

C. When subcooling or pressurizer level meets foldout page criteria then manually start and align Emergency Core Cooling System pumps.

D. When subcooling or pressurizer level meets foldout page criteria then re-actuate Safety Injection to align Emergency Core Cooling System.

Proposed Answer: C Page 15 of 93 CPNPP NRC 2013 RO Written Exam Worksheet 1 to 25 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible because manually starting and aligning ECCS pumps is required per EOS-1.2 foldout page, however, subcooling is NOT less than 55°F or PZRZ level is NOT less than 34% for adverse containment.

B. Incorrect. Plausible because it may be thought that re-actuating SI is the proper action because EOS-1.2A, foldout page title for step is SI REINITIATION CRITERIA, however, subcooling is NOT less than 55°F or PZRZ level is NOT less than 34% for adverse containment.

C. Correct. Manually starting and aligning ECCS pumps is correct action when subcooling is less than 55°F or PZRZ level is less than 34%.

D. Incorrect. Plausible because it may be thought that re-actuating SI is the proper action because EOS-1.2A, foldout page title for step is SI REINITIATION CRITERIA when subcooling is less than 55°F or PZRZ level is less than 34%.

Technical Reference(s) EOS-1.2A, foldout page and bases Attached w/ Revision # See Comments / Reference Proposed references to be provided during examination: None Learning Objective: DISCUSS the symptoms or Entry Conditions for FRC-0.3, Response to Saturated 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 Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 Page 16 of 93 CPNPP NRC 2013 RO Written Exam Worksheet 1 to 25 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From EOS-1.2A Foldout Page Revision 8 Page 17 of 93 CPNPP NRC 2013 RO Written Exam Worksheet 1 to 25 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From EOS-1.2A, Attachment 1.A bases Revision 8 Page 18 of 93 CPNPP NRC 2013 RO Written Exam Worksheet 1 to 25 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 006 K5.06 Importance Rating 3.5 Emergency Core Cooling System: Knowledge of the operational implications of the following concepts as they apply to the ECCS: Relationship between ECCS flow and RCS pressure Proposed Question: Common 6 Given the following:

• A 3 Small Break Loss of Coolant Accident has occurred on Unit 1.

Which of the following describes the expected response of the Safety Injection Pump discharge flow as Reactor Coolant System pressure lowers from 1400 psig to 1200 psig?

Each Safety Injection Pump is initially discharging at...

A. 100 gpm and stable.

B. 0 gpm and stable.

C. 100 gpm and steadily rising to 250 gpm.

D. 0 gpm and steadily rising to 150 gpm.

Proposed Answer: C Explanation:

A. Incorrect. Plausible if thought that the Safety Injection Pump curves are vertical between 1400 and 1200 psig RCS pressure, however the flow is rising at a fairly substantial rate along this portion of the pump curve as the curve is more horizontal than vertical.

B. Incorrect. Plausible if thought that the Safety Injection Pump begin injecting into the RCS below 1200 psig, however, based on shutoff head and line losses the pumps begin injecting into the RCS at a pressure of approximately 1450 psig.

C. Correct. At 1400 psig RCS pressure (which equates to approximately 3500 ft of head, when RWST suction pressure and injection line losses are considered), each Safety Injection Pump would be injecting approximately 100 gpm with 45 gpm of recirculation flow. At 1200 psig RCS pressure (with similar assumptions as the 1400 psig condition) the pump would be injecting approximately 250 gpm based on the CPNPP Safety Injection Pumps.

D. Incorrect. Plausible if thought that the Safety Injection Pump begin injecting slightly below 1400 psig RCS pressure and flow would then increase at approximately the same rate as the actual pump head curve, however, based on shutoff head and line losses the pumps begin injecting into the RCS at a pressure of approximately 1450 psig.

Page 19 of 93 CPNPP NRC 2013 RO Written Exam Worksheet 1 to 25 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Technical Reference(s) OPT-513A-1 Attached w/ Revision # See Comments / Reference Proposed references to be provided during examination: None Learning Objective: DESCRIBE the basic design and flow path of the Emergency Core Cooling System.

DEMONSTRATE an understanding of the components of the Emergency Core Cooling 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 10, 14 55.43 Page 20 of 93 CPNPP NRC 2013 RO Written Exam Worksheet 1 to 25 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From OPT-513A-1 4/17/2001 Data Page 21 of 93 CPNPP NRC 2013 RO Written Exam Worksheet 1 to 25 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 007 K3.01 Importance Rating 3.3 Pressurizer Relief/Quench Tank System: Knowledge of the effect that a loss or malfunction of the PRTS will have on the following: Containment Proposed Question: Common 7 Given the following conditions:

• A Pressurizer (PRZR) Power Operated Relief Valve (PORV) opened at 2335 psig and will NOT close.

• The associated PRZR PORV Block Valve failed to close manually and the Pressurizer Relief Tank rupture disk has blown.

• PRZR PORV Outlet (Tailpipe) temperature is indicating 228°F.

Which of the following is the expected Containment pressure for the conditions listed?

A. 2 psig B. 5 psig C. 10 psig D. 15 psig Proposed Answer: B Explanation:

A. Incorrect. Plausible if Mollier Diagram is improperly read.

B. Correct. With a nominal opening pressure of 2335 psig, the isenthalpic expansion occurs at approximately ~1110 BTUs/lbm. Following this on the curve to the point where 228°F intersects the Saturation Curve corresponds to a Containment pressure of approximately 20 psia, which is equal to 5 psig.

C. Incorrect. Plausible if Mollier Diagram is improperly read.

D. Incorrect. Plausible if Mollier Diagram is improperly read.

Technical Reference(s) Mollier Diagram Attached w/ Revision # See Comments / Reference Proposed references to be provided during examination: Steam Tables Learning Objective: DISCUSS the operator actions, including all cautions, notes, RNOs and bases associated with EOP-1.0, Loss of Reactor or Secondary Coolant.

Page 22 of 93 CPNPP NRC 2013 RO Written Exam Worksheet 1 to 25 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Question Source: Bank #

Modified Bank # X (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, 14 55.43 Comments /

Reference:

From Mollier Diagram Revision N/A Page 23 of 93 CPNPP NRC 2013 RO Written Exam Worksheet 1 to 25 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Original Bank Question Given the following conditions:

• A Pressurizer (PRZR) Power Operated Relief Valve (PORV) opened at 2335 psig and will NOT close.

• The associated PRZR PORV Block Valve failed to close manually and the Pressurizer Relief Tank rupture disk has blown.

• PRZR PORV Outlet (Tailpipe) temperature is indicating 260°F.

Which of the following is the expected Containment pressure for the conditions listed?

A. ~5 psig.

B. ~20 psig.

C. ~35 psig.

D. ~50 psig.

Page 24 of 93 CPNPP NRC 2013 RO Written Exam Worksheet 1 to 25 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 008 A4.01 Importance Rating 3.3 Component Cooling Water System: Ability to manually operate and/or monitor in the control room: CCW indications and controls Proposed Question: Common 8 Given the following conditions:

• Unit 1 is operating at 100% power.

• Train A Component Cooling Water System is in service.

• Annunciator 1-ALB-3B, Window 2.2 - CCW SRG TK TRN A/B EMPTY is in alarm.

• 1-LR-4500, TRN A SRG TK LVL is reading 53% and lowering.

Which of the following is the response of the Component Cooling Water (CCW) System?

A. Train A CCW Pump trips; Train B CCW Pump AUTO starts.

B. Train A CCW Pump trips; Train B CCW Pump remains in STANDBY.

C. Train A Safeguards Loop Isolation Valves close; Train B CCW Pump AUTO starts.

D. Train A Safeguards Loop Isolation Valves close; Train B CCW Pump remains in STANDBY.

Proposed Answer: D Explanation:

A. Incorrect. Plausible because it is a misconception that an empty CCW Surge Tank would trip Train A CCW Pump causing an AUTO start of the Train B CCW Pump. An empty CCW Surge Tank does not trip Train A CCW Pump, therefore, Train B CCW Pump does not auto start.

B. Incorrect. Plausible because it is a misconception that an empty CCW Surge Tank would trip Train A CCW Pump causing an AUTO start of the Train B CCW Pump. An empty CCW Surge Tank does not trip Train A CCW Pump, therefore, Train B CCW Pump remains in standby.

C. Incorrect. Plausible because it is a misconception that closure of Train A Safeguards Loop Isolation valves would lead to an AUTO start of Train B CCW Pump.

D. Correct. Train A Safeguards Loop Isolation Valves will close with CCW Surge Tank level at < 57%

and the Train B Component Cooling Water Pump remains in standby.

Technical Reference(s) ALM-0032A, 1-ALB-3B, Window 2.2 Attached w/ Revision # See ABN-502, Steps 2.1, 2.2, & 3.2 Comments / Reference Page 25 of 93 CPNPP NRC 2013 RO Written Exam Worksheet 1 to 25 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Proposed references to be provided during examination: None Learning Objective: IDENTIFY the Main Control Board/Plant Computer controls, alarms and indications associated with the Component Cooling Water System.

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

New Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 4, 7 55.43 Page 26 of 93 CPNPP NRC 2013 RO Written Exam Worksheet 1 to 25 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ALM-0032A, 1-ALB-3B, Window 2.2 Revision # 7 Page 27 of 93 CPNPP NRC 2013 RO Written Exam Worksheet 1 to 25 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ABN-502, Steps 2.1 & 2.2 Revision # 6 Comments /

Reference:

From ABN-502, Step 3.2 Revision # 6 Page 28 of 93 CPNPP NRC 2013 RO Written Exam Worksheet 1 to 25 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 010 K6.01 Importance Rating 2.7 Pressurizer Pressure Control System: Knowledge of the effect that a loss or malfunction of the following will have on the PPCS: Pressure detection systems Proposed Question: Common 9 Given the following conditions:

• Unit 1 is operating at 50% power.

• PS-455F, PRZR PRESS CTRL CHAN SELECT, is in the 455/456 position.

• PT-456, PRZR PRESS CHAN II, fails high.

Assuming NO operator action, which of the following is the expected plant response?

A. A high pressure Reactor Trip occurs.

B. A low pressure Reactor Trip and Safety Injection occur.

C. Unit remains at power, with pressure being controlled at approximately 2185 psig.

D. Unit remains at power, with pressure being controlled at approximately 2335 psig.

Proposed Answer: C Explanation:

A. Incorrect. Plausible because pressure would increase with no operator action if PT-455 were to fail low and, if PORV 456 were to also fail, a high pressure Reactor Trip would occur, but with PT-456 failing high, pressure stabilizes around 2185 psig.

B. Incorrect. Plausible because a low pressure Reactor Trip and Safety Injection would occur with no operator action if PT-455 were to fail high, but with PT-456 failing high pressure stabilizes around 2185 psig.

C. Correct. With PT-456 failing high, PORV 456 opens and actual pressure begins lowering. The PRZR Spray Valves remain closed and the PRZR Heaters will energize as pressure sensed by PT-455 lowers, but the open PORV causes pressure to continue to lower. When PT-457 senses pressure below 2185 psig, the open interlock for PORV 456 is lost and the PORV closes. As pressure increases above 2185 psig due to the heaters, the open interlock for the PORV is restored and pressure will cycle around 2185 psig.

D. Incorrect. Plausible because pressure would increase and cycle around 2335 psig with no operator action if PT-455 were to fail low, but with PT-456 failing high pressure stabilizes around 2185 psig.

Technical Reference(s) ABN-705, Step 2.2 Attached w/ Revision # See Comments / Reference Page 29 of 93 CPNPP NRC 2013 RO Written Exam Worksheet 1 to 25 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 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 # X Modified Bank # (Note changes or attach parent)

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

Reference:

From ABN-705, Step 2.2 Revision # 12 Page 30 of 93 CPNPP NRC 2013 RO Written Exam Worksheet 1 to 25 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 012 K1.02 Importance Rating 3.4 Reactor Protection System: Knowledge of the physical connections and/or cause-effect relationships between the RPS and the following systems: 125V DC system Proposed Question: Common 10 Given the following conditions:

• Unit 2 is operating at 100% power.

• 125 VDC Bus 2ED1 de-energizes due to a fault on the bus.

Which of the following describes the effect of a loss of DC bus 2ED1?

A. Reactor will trip due to a shunt trip of Train A Reactor Trip Breaker.

B. Reactor will trip due to an undervoltage trip of Train A Reactor Trip Breaker.

C. A shunt trip signal will NOT be capable of opening Train A Reactor Trip Breaker.

D. An undervoltage trip signal will NOT be capable of opening Train A Reactor Trip Breaker.

Proposed Answer: C Explanation: With the Reactor Trip breakers being part of the Reactor Protection System the following plausibility statements are provided; A. Incorrect. Plausible because most actuations are de-energized to actuate, but the shunt trip requires that 125 VDC be applied to the shunt trip coil to cause the breaker to open.

B. Incorrect. Plausible because the UV trip receives power from a DC power supply, but the power supply is 48 VDC within SSPS.

C. Correct. 125 VDC Bus 2ED1 supplies power to the shunt trip coil for Train A Trip Breaker. UV coils and shunt trip relays are supplied from 48 VDC from SSPS. The shunt trip coil is normally de-energized and without power available, a shunt trip of Train A Trip Breaker is not possible.

D. Incorrect. Plausible because the UV trip receives power from a DC power supply, but the power supply is 48 VDC within SSPS and is de-energized to actuate.

Technical Reference(s) LO21.SYS.ES2 Attached w/ Revision # See Comments / Reference Proposed references to be provided during examination: None Page 31 of 93 CPNPP NRC 2013 RO Written Exam Worksheet 1 to 25 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 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 # X 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 6 55.43 Page 32 of 93 CPNPP NRC 2013 RO Written Exam Worksheet 1 to 25 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From LO21.SYS.ES2.LN, Page 66 Revision # 09/02/04 THE SHUNT TRIP COIL The Shunt Trip Coil on the Reactor Trip and Bypass breakers is actuated by any of the following:

• Either Manual Reactor Trip switch

• Either Manual SI Actuation switch

• Auto Shunt Trip Relay STA(B) - only on the Reactor Trip Breakers

• Both Bypass Breakers connected and closed- only on the Bypass Breakers The Shunt Trip Coil is normally de-energized. When it is actuated by applying 125 VDC to its coil, the coil attracts an armature which pushes the trip lever on the breaker trip shaft, causing the breaker to trip (Figure 20). This trip device is a mechanically less complicated and more forceful mechanism than the undervoltage trip coil. The power to the shunt trip coils comes from uED1 (2)-2.

Page 33 of 93 CPNPP NRC 2013 RO Written Exam Worksheet 1 to 25 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 013 K5.01 Importance Rating 2.8 Engineered Safety Features Actuation System: Knowledge of the operational implications of the following concepts as they apply to the ESFAS: Definitions of safety train and ESF channel Proposed Question: Common 11 What pumps make up a complete Emergency Core Cooling System safety train in MODE 4?

A. A Centrifugal Charging Pump, a Safety Injection Pump and a Residual Heat Removal Pump.

B. A Safety Injection Pump and a Residual Heat Removal Pump.

C. A Safety Injection Pump and a Centrifugal Charging Pump.

D. A Centrifugal Charging Pump and a Residual Heat Removal Pump.

Proposed Answer: D Explanation: In order to define a safety train the Technical Specification bases are used as the source of the definition. That definition changes based on MODE.

A. Incorrect. Plausible because a Centrifugal Charging Pump, a Safety Injection Pump and a Residual Heat Removal Pump are required in MODE 1, 2 & 3 per T/S LCO 3.5.2.

B. Incorrect. Plausible if thought that a Safety Injection Pump and a Residual Heat Removal Pump meet the requirement in MODE 4 per T/S LCO 3.5.3.

C. Incorrect. Plausible if thought that a Safety Injection Pump and a Centrifugal Charging Pump meet the requirement in MODE 4 per T/S LCO 3.5.3.

D. Correct. A Centrifugal Charging Pump and a Residual Heat Removal Pump are required in MODE 4 per T/S LCO 3.5.3.

Technical Reference(s) Technical Specification 3.5.2 Attached w/ Revision # See Technical Specification 3.5.3 Comments / Reference Proposed references to be provided during examination: None Learning Objective: COMPREHEND the normal, abnormal and emergency operation of the Solid State Protection System.

COMPREHEND the normal, abnormal and emergency operation of the Safety Injection and Blackout Sequencers.

Question Source: Bank #

Page 34 of 93 CPNPP NRC 2013 RO Written Exam Worksheet 1 to 25 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 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, 7, 8 55.43 Page 35 of 93 CPNPP NRC 2013 RO Written Exam Worksheet 1 to 25 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From Technical Specification B3.5.2 Revision 67 Page 36 of 93 CPNPP NRC 2013 RO Written Exam Worksheet 1 to 25 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From Technical Specification B3.5.3 Revision 67 Page 37 of 93 CPNPP NRC 2013 RO Written Exam Worksheet 1 to 25 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 022 A1.02 Importance Rating 3.6 Containment Cooling System: Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the CCS controls including: Containment pressure Proposed Question: Common 12 Given the following conditions:

• Unit 1 Containment pressure is 1.2 psig and slowly lowering.

• A Containment Vent is in progress per SOP-801A, Containment Ventilation System.

What MIMIMUM containment pressure must be maintained during the containment vent and what is the basis for this value?

A. -0.3 psig to ensure containment will NOT exceed design negative differential pressure following inadvertent Containment Spray system actuation.

B. 0.0 psig to ensure containment will NOT exceed design negative differential pressure following inadvertent Containment Spray system actuation.

C. -0.3 psig to ensure instrumentation impacted by containment pressure will operate as designed during normal, abnormal and emergency conditions.

D. 0.0 psig to ensure instrumentation impacted by containment pressure will operate as designed during normal, abnormal and emergency conditions.

Proposed Answer: A Explanation:

A. Correct. Per Technical Specification LCO 3.6.4 bases.

B. Incorrect. Plausible because the reason is correct but the minimum pressure is not correct.

C. Incorrect. Plausible because the pressure is correct but the reason is not correct.

D. Incorrect. Plausible if thought the pressure and the reason are correct.

Technical Reference(s) Technical Specification LCO 3.6.4 Attached w/ Revision # See Technical Specification LCO 3.6.4 bases Comments / Reference Proposed references to be provided during examination: None Learning Objective: COMPREHEND the normal, abnormal and emergency operation of the Containment Ventilation System.

Page 38 of 93 CPNPP NRC 2013 RO Written Exam Worksheet 1 to 25 Rev h

ES-401 CPNPP NRC 2013 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 9, 10 55.43 Comments /

Reference:

From Technical Specification LCO 3.6.4 Amendment 158 Page 39 of 93 CPNPP NRC 2013 RO Written Exam Worksheet 1 to 25 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From Technical Specification LCO 3.6.4 bases Revision 67 Page 40 of 93 CPNPP NRC 2013 RO Written Exam Worksheet 1 to 25 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 022 A4.01 Importance Rating 3.6 Containment Cooling System: Ability to manually operate and/or monitor in the control room: CCS fans Proposed Question: Common 13 Assuming Containment Recirculation Cooler Fan 1-01 was operating, which of the following identifies the expected handswitch indication lights for 1-HS-5405A, CNTMT FN CLR FN 1, two minutes following a Safety Injection?

GREEN AMBER RED A. OFF OFF ON B. ON ON ON C. ON ON OFF D. OFF OFF OFF Proposed Answer: C Explanation:

A. Incorrect. Plausible because these are the indications that would be available if the fan were still operating which would occur 40 seconds after a Blackout instead of a Safety Injection (SI), but the fan will load shed on the SI.

B. Incorrect. Plausible if thought that the shunt trip of the breaker caused all lights to illuminate.

C. Correct. The fan is load shed via a shunt trip of the breaker upon receipt of an SI signal. With the handswitch in the AUTO AFTER START position, this will cause the amber light to energize. The green light is on because the breaker is open, which is also why the red light is off.

D. Incorrect. Plausible if thought that the shunt trip of the breaker caused all lights to extinguish, however control power for the circuit is 125 VDC.

Technical Reference(s) EOP-0.0A, Attachment 8 Attached w/ Revision # See ALM-0031A, 1-ALB-3A, Window 1.2 Comments / Reference Proposed references to be provided during examination: None Learning Objective: COMPREHEND the normal, abnormal and emergency operation of the Containment Ventilation System.

Page 41 of 93 CPNPP NRC 2013 RO Written Exam Worksheet 1 to 25 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Question Source: Bank # X 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, 9 55.43 Page 42 of 93 CPNPP NRC 2013 RO Written Exam Worksheet 1 to 25 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From EOP-0.0A, Attachment 8 Revision # 8 Page 43 of 93 CPNPP NRC 2013 RO Written Exam Worksheet 1 to 25 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ALM-0031A, 1-ALB-3A, Window 1.2 Revision # 8 Page 44 of 93 CPNPP NRC 2013 RO Written Exam Worksheet 1 to 25 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 026 A2.07 Importance Rating 3.6 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: Loss of containment spray pump suction when in recirculation mode, possibly caused by clogged sump screen, pump inlet high temperature exceeded cavitation, voiding or sump level below cutoff (interlock) limit Proposed Question: Common 14 Given the following conditions:

• Unit 1 experienced a Large Break Loss of Coolant Accident (LOCA) about 20 minutes ago.

• Train A Emergency Core Cooling System (ECCS) Pumps are running in Cold Leg Recirculation Mode per EOS-1.3 A, Transfer to Cold Leg Recirculation.

• Transfer of Containment Spray Pumps to recirculation is complete with the following indications:

• Containment Spray Pumps 1-01 and 1-03 are running with:

• 3750 gpm and stable discharge flow on each pump.

• 265 psig and stable discharge pressure on each pump.

• Containment Spray Pumps 1-02 and 1-04 are running with:

• 3800 to 2800 gpm fluctuating discharge flow on each pump.

• 270 to 160 psig fluctuating discharge pressure on each pump.

Which of the following lists the action required per EOS-1.3 A, Transfer to Cold Leg Recirculation?

A. Close 1-HV-4777, CS HX 1-02 OUT VLV.

B. Open 1-HV-4759, RWST TO CS PMP 1-02 & 1-04 SUCT VLV.

C. Place Containment Spray Pumps 1-02 & 1-04 handswitches in PULLOUT.

D. Place Containment Spray Pump 1-02 in PULLOUT and check for flow and pressure improvement.

Proposed Answer: C Page 45 of 93 CPNPP NRC 2013 RO Written Exam Worksheet 1 to 25 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible because 1-HV-4777, CS HX 1-02 OUT VLV is closed if Containment Spray Pumps have been stopped due to low RWST level, however, the valve would be reopened when the Containment Spray Pumps were started.

B. Incorrect. Plausible if thought that opening 1-HV-4759, RWST TO CS PMP 1-02 & 1-04 SUCT VLV would assist with eliminating cavitation in the Containment Spray Pumps, however, under no condition in EOS-1.3A is this valve opened. Step 12 of EOS-1.3A does have the operator makeup to the RWST which could lead an individual to select distractor B.

C. Correct. Per the EOS-1.3A, Step 3 CAUTION.

D. Incorrect. Plausible if thought that closing 1-HV-4783, CNTMT SMP TO CS PMP 1-02 & 1-04 SUCT ISOL VLV would assist with eliminating cavitation in the Containment Spray Pumps, however, under no condition in EOS-1.3A is this valve closed.

Technical Reference(s) EOS-1.3A, Step 3 CAUTION Attached w/ Revision # See EOS-1.3A, Step 3 & Step 4 RNO Comments / Reference 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 10, 14 55.43 Page 46 of 93 CPNPP NRC 2013 RO Written Exam Worksheet 1 to 25 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From EOS-1.3A, Step 3 CAUTION Revision # 8 Page 47 of 93 CPNPP NRC 2013 RO Written Exam Worksheet 1 to 25 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From EOS-1.3A, Step 3 Revision # 8 Page 48 of 93 CPNPP NRC 2013 RO Written Exam Worksheet 1 to 25 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From EOS-1.3A, Step 4 RNO Revision # 8 Page 49 of 93 CPNPP NRC 2013 RO Written Exam Worksheet 1 to 25 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 039 A1.05 Importance Rating 3.2 Main and Reheat Steam System: Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the MRSS controls including: RCS Tave Proposed Question: Common 15 Given the following conditions on Unit 1:

• A Reactor and Turbine Trip have just occurred from 100% power.

• Loop 1 T AVE Channel has failed HIGH.

Which of the following describes the effect on the Steam Dump System due to the Loop 1 T AVE Channel failure?

A. Steam Dumps will OPEN and reduce T AVE to LESS than No-Load T AVE .

B. Steam Dumps will OPEN and reduce T AVE to No-Load T AVE .

C. Steam Dump System will automatically shift from the Plant Trip Controller to the Steam Pressure Mode and reduce steam pressure to LESS than 1092 psig.

D. Steam Dump System will automatically shift from the Plant Trip Controller to the Steam Pressure Mode and reduce steam pressure to 1092 psig.

Proposed Answer: A Page 50 of 93 CPNPP NRC 2013 RO Written Exam Worksheet 1 to 25 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Explanation:

A. Correct. The Steam Dump Valves will control based on the differential temperature between the Average T AVE and No-Load T AVE . The Average T AVE will be higher based on the channel failure and keep the valves open past where they should be closed.

B. Incorrect. Plausible because the Steam Dump Valves will normally control temperature at no load temperature without the channel failure.

C. Incorrect. Plausible because it could be thought that the system would swap to Pressure Control because this is the mode normally used when below no load temperature, however, the system does not automatically shift to Pressure Control mode.

D. Incorrect. Plausible because it could be thought that the system would swap to Pressure Control because this is the mode normally used when below no load temperature, however, the system does not automatically shift to Pressure Control mode and temperature will stabilize below no load temperature.

Technical Reference(s) ABN-704, Steps 2.2.a, 2.3.2 and 2.3.3 Attached w/ Revision # See EOP-0.0A, Step 9 RNO a Comments / Reference Proposed references to be provided during examination: None Learning Objective: COMPREHEND the normal, abnormal and emergency operation of the Steam Dump System.

Question Source: Bank #

Modified Bank # X (Note changes or attach parent)

New Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 4, 5, 7, 10 55.43 Page 51 of 93 CPNPP NRC 2013 RO Written Exam Worksheet 1 to 25 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ABN-704, Step 2.2.a Revision # 10 Page 52 of 93 CPNPP NRC 2013 RO Written Exam Worksheet 1 to 25 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ABN-704, 2.3.2 and 2.3.3 Revision # 10 Page 53 of 93 CPNPP NRC 2013 RO Written Exam Worksheet 1 to 25 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From EOP-0.0A, Step 9 RNO a Revision #

Original Bank Question Page 54 of 93 CPNPP NRC 2013 RO Written Exam Worksheet 1 to 25 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 059 K4.16 Importance Rating 3.1 Main Feedwater System: Knowledge of the MFW design feature(s) and/or interlock(s) that provide for the following:

Automatic trips for MFW pumps Proposed Question: Common 16 Which of the following conditions or situations will result in an automatic trip of a Main Feedwater Pump?

A. Safety Injection Signal.

B. High Main Feedwater Pump vibration.

C. Low Condenser vacuum of 21 HG.

D. LO-LO T AVE with Reactor Trip Signal.

Proposed Answer: A Explanation:

A. Correct. This is the only signal listed that will result in trip of the Main Feedwater Pump.

B. Incorrect. Plausible because the Main Feedwater Pump is equipped with a vibration monitor, however, high vibration does not generate a Main Feedwater Pump trip.

C. Incorrect. Plausible because the Main Feedwater Pump will trip on low vacuum, however, the setpoint is 17.5 HG.

D. Incorrect. Plausible because this permissive will trip the Main Turbine and initiate automatic Feedwater Isolation, however, it does not generate a Main Feedwater Pump trip.

Technical Reference(s) SOP-302A, Step 4.2.C Attached w/ Revision # See ALM-0065A, 1-PCIP, Window 1.5 Comments / Reference LO21.SYS.MF1 Proposed references to be provided during examination: None Learning Objective: EXPLAIN the instrumentation and controls of the Main Feedwater System and PREDICT the system response.

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

New Question History: Last NRC Exam Page 55 of 93 CPNPP NRC 2013 RO Written Exam Worksheet 1 to 25 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 Comments /

Reference:

From SOP-302A, Step 4.2.C Revision # 17 Page 56 of 93 CPNPP NRC 2013 RO Written Exam Worksheet 1 to 25 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ALM-0065A, 1-PCIP, Window 1.5 Revision # 4 Comments /

Reference:

From LO21.SYS.MF1.LN, Page 55 Revision # 05/31/07 FEED PUMP VIBRATION MONITOR The Unit 1 main feed pumps use the Bentley-Nevada proximity probe type vibration monitors. VD1 and VD2 sense vibration amplitude in the X direction while VD3 and VD4 sense amplitude in the Y direction. The X and Y signals are auctioneered high for display. These where chosen because they can interface directly with the MK-V digital control system and are displayed on the and plant computer. Unit 2 has the original Bentley-Nevada vibration detectors and the MK-V. The Unit 2 MK-V will input to the plant computer just like Unit 1 but does not have the axial vibration data input. Unit 1 uses all live data input.

Page 57 of 93 CPNPP NRC 2013 RO Written Exam Worksheet 1 to 25 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 061 K5.03 Importance Rating 2.6 Auxiliary/Emergency Feedwater System: Knowledge of the operational implications of the following concepts as they apply to the AFW: Pump head effects when control valve is shut Proposed Question: Common 17 Given the following conditions:

• Unit 1 has experienced a Main Steam Line Break inside Containment from Steam Generator 1-02.

• Containment pressure is 12 psig and rising.

• Following isolation of Steam Generator 1-02, Steam Generator 1-01 is identified as ruptured.

• Steam Generator 1-01 narrow range level is 40% and rising.

Once Steam Generator 1-01 level reaches [1] %, Auxiliary Feedwater flow should be stopped to Steam Generator 1-01. When the Flow Control Valve is closed Motor Driven Auxiliary Feedwater Pump 1-01 discharge pressure will [2].

[1] [2]

A. 43 increase B. 43 decrease C. 50 increase D. 50 decrease Proposed Answer: C Page 58 of 93 CPNPP NRC 2013 RO Written Exam Worksheet 1 to 25 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible because the pump head will increase when the Flow Control Valve is closed, however, with Adverse Containment parameters this action should not be taken until the level in the ruptured Steam Generator is a minimum of 50%.

B. Incorrect. Plausible if thought that overall flow would increase when the Flow Control Valve is closed as occurs with other pumps (i.e., the miniflow valve opens when the normal discharge path is closed). However, with the Auxiliary Feedwater Pumps the miniflow is open throughout the evolution. Additionally, with Adverse Containment parameters this action should not be taken until the level in the ruptured Steam Generator is a minimum of 50%.

C. Correct. Given the conditions listed, once Steam Generator level rises above 50%, the Flow Control Valve should be closed restricting the pump discharge to only miniflow, this restriction of the discharge flow will result in an increase in discharge pressure.

D. Incorrect. Plausible because once Steam Generator level rises above 50%, the Flow Control Valve should be closed restricting the pump discharge to only miniflow. If thought that overall flow would increase when the Flow Control Valve is closed as occurs with other pumps (i.e., the miniflow valve opens when the normal discharge path is closed), and discharge pressure could decrease.

However, with the Auxiliary Feedwater Pumps the miniflow is open throughout the evolution.

Technical Reference(s) EOP-3.0A, Step 4 Attached w/ Revision # See Comments / Reference Proposed references to be provided during examination: None Learning Objective: STATE the bases for operator actions, notes and cautions from EOP-3.0, 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 7, 10 55.43 Page 59 of 93 CPNPP NRC 2013 RO Written Exam Worksheet 1 to 25 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From EOP-3.0A, Step 4 Revision 8 Page 60 of 93 CPNPP NRC 2013 RO Written Exam Worksheet 1 to 25 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 062 K4.02 Importance Rating 2.5 AC Electrical Distribution System: Knowledge of the AC electrical distribution design feature(s) and/or interlock(s) which provide for the following: Circuit breaker automatic trips Proposed Question: Common 18 Which of the following describes a design feature of the Second Level Undervoltage Relays?

A. Starts the Emergency Diesel Generators and initiates the Blackout Sequencer.

B. Provides transfer to an energized Offsite Power source upon a loss of bus voltage.

C. Protects 1E motors from a low voltage/high current condition by opening the Preferred and Alternate Feeder Breakers.

D. Ensures the opening of the Alternate Feeder Breaker within one second after the Preferred Feeder Breaker closes.

Proposed Answer: C Explanation:

A. Incorrect. Plausible because the Emergency Diesel Generators will start due to bus undervoltage, however, the Blackout Sequencer is initiated from another source.

B. Incorrect. Plausible if thought that was the function of these relays, however, the 2nd Level Undervoltage Relays are used on the 1E Safeguards Buses.

C. Correct. Per ABN-602, automatic actions.

D. Incorrect. Plausible because this action can occur, however, it is not associated with these relays.

Technical Reference(s) ABN-602, Step 2.2 & Attachment 3 Attached w/ Revision # See Comments / Reference 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 # X Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam Page 61 of 93 CPNPP NRC 2013 RO Written Exam Worksheet 1 to 25 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 Comments /

Reference:

From ABN-602, Step 2.2 Revision # 8 Page 62 of 93 CPNPP NRC 2013 RO Written Exam Worksheet 1 to 25 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ABN-602, Attachment 3 Revision # 8 Page 63 of 93 CPNPP NRC 2013 RO Written Exam Worksheet 1 to 25 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 063 A3.01 Importance Rating 2.7 DC Electrical Distribution System: Ability to monitor automatic operation of the DC electrical distribution, including: Meters, annunciators, dials, recorders and indicating lights Proposed Question: Common 19 Given the following indications on Unit 1 DC Safeguards Switch Panels (Buses) and Batteries:

• DC Switch Panel 1ED1 is at 135 volts with battery BT1ED1 at 2 amp CHARGE.

• DC Switch Panel 1ED2 is at 136 volts with battery BT1ED2 at 2 amp CHARGE.

• DC Switch Panel 1ED3 is at 135 volts with battery BT1ED3 at 1 amp CHARGE.

• DC Switch Panel 1ED4 is at 140 volts with battery BT1ED4 at 10 amp CHARGE.

Which of the following lists the status of the Unit 1 DC Safeguards batteries?

BT1ED1 BT1ED2 BT1ED3 BT1ED4 A. FLOAT EQUALIZE FLOAT EQUALIZE B. EQUALIZE FLOAT EQUALIZE FLOAT C. EQUALIZE EQUALIZE EQUALIZE FLOAT D. FLOAT FLOAT FLOAT EQUALIZE Proposed Answer: D Explanation:

A. Incorrect. Plausible because it could be thought that the higher voltages on 1ED2 and 1ED4 indicate the battery is in EQUALIZE.

B. Incorrect. Plausible because it could be thought that the higher voltages on 1ED2 and 1ED4 indicate the battery is in FLOAT.

C. Incorrect. Plausible because of a misconception with regard to indications of FLOAT and EQUALIZE being reversed.

D. Correct. Voltage of between 138 VDC and 140 VDC is expected during an EQUALIZE charge on a battery with increased charging current.

Technical Reference(s) SOP-605A Attached w/ Revision # See Comments / Reference Page 64 of 93 CPNPP NRC 2013 RO Written Exam Worksheet 1 to 25 Rev h

ES-401 CPNPP NRC 2013 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 DC 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 Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 5, 7, 8 55.43 Page 65 of 93 CPNPP NRC 2013 RO Written Exam Worksheet 1 to 25 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SOP-605A Revision 11 Page 66 of 93 CPNPP NRC 2013 RO Written Exam Worksheet 1 to 25 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 064 A2.18 Importance Rating 2.6 Emergency Diesel Generator System: Ability to (a) predict the impacts of the following malfunctions or operations on the EDG system; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Consequences of premature opening of breaker under load Proposed Question: Common 20 Given the following conditions:

• Unit 1 is in MODE 1.

• OPT-214A, Diesel Generator Operability Test, is in progress with Emergency Diesel Generator (EDG) 1-01 paralleled to the grid.

• CS-1EA1-1, INCOMING BKR 1EA1-1 and CS-1EG1, DG BKR 1EG1 are closed.

• With the test in progress a loss of several other units connected to the grid results in grid frequency dropping to 58.6 Hertz.

Which of the following states the required operator action and the reason for that action?

The operator will...

A. open CS-1EG1, DG BKR 1EG1 to prevent excessive loading on the EDG.

B. pullout CS-1EA1-2, INCOMING BKR 1EA1-2 and CS-1EA1-1, INCOMING BKR 1EA1-1 to prevent excessive loading on the EDG.

C. open CS-1EG1, DG BKR 1EG1 to prevent excessive loading on CS-1EA1-1, INCOMING BKR 1EA1-1.

D. pullout CS-1EA1-2, INCOMING BKR 1EA1-2 and CS-1EA1-1, INCOMING BKR 1EA1-1 to prevent excessive loading on CS-1EA1-1, INCOMING BKR 1EA1-1.

Proposed Answer: A Page 67 of 93 CPNPP NRC 2013 RO Written Exam Worksheet 1 to 25 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Explanation:

A. Correct. Given the conditions listed, the output breaker is opened to prevent excessive loading on the EDG.

B. Incorrect. Plausible because preventing excessive loading on the EDG is the proper concern and opening both the alternate and preferred offsite breakers would disconnect the bus from the degrading grid. However, proper procedural guidance is to open the EDG output breaker.

C. Incorrect. Plausible because the output breaker is opened, however, the reason is not to prevent excessive loading on the incoming breaker but to prevent excessive loading on the EDG.

D. Incorrect. Plausible because preventing excessive loading is correct but not excessive loading on the incoming breaker. Opening both the alternate and preferred offsite breakers would disconnect the bus from the degrading grid. However, proper procedural guidance is to open the EDG output breaker.

Technical Reference(s) OPT-214A, Step 8.1.T Attached w/ Revision # See Comments / Reference Proposed references to be provided during examination: None Learning Objective: EXPLAIN the instrumentation and controls of the Emergency Diesel Generator 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, 10 55.43 Page 68 of 93 CPNPP NRC 2013 RO Written Exam Worksheet 1 to 25 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From OPT-214A, Step 8.1.T Revision # 22 Page 69 of 93 CPNPP NRC 2013 RO Written Exam Worksheet 1 to 25 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 064 K4.10 Importance Rating 3.5 Emergency Diesel Generator System: Knowledge of the EDG design feature(s) and/or interlock(s) that provide for the following: Automatic load sequencer: blackout Proposed Question: Common 21 Given the following conditions:

• Work is in progress to place XST2A, Spare 345 kV Startup Transformer, in service.

• XST2, 345 kV Startup Transformer, is currently tagged out to support the work activities.

• 1EA1, 6.9kV Safeguards Bus, is being supplied from XST1, 138 kV Startup Transformer, via CS-1EA1-2, INCOMING BKR 1EA1-2.

With this configuration, if a sudden pressure fault were to occur on XST1, Bus 1EA1 would be de-energized, Emergency Diesel Generator (EDG) 1-01 would start, CS-1EG1, DG BKR 1EG1 would ...

A. automatically CLOSE, and 1EA1 would have to be manually loaded by the operators.

B. automatically CLOSE, then the Blackout Sequencer would load 1EA1.

C. remain OPEN, may be manually closed, then the Blackout Sequencer would load 1EA1.

D. remain OPEN, may be manually closed, then manually loaded by the operators.

Proposed Answer: B Page 70 of 93 CPNPP NRC 2013 RO Written Exam Worksheet 1 to 25 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible because when 1EA1 is de-energized, the EDG would auto start, the output breaker would automatically close, however, the BOS would automatically load the bus. If the BOS sequencer failed to run, the operator may manually load the bus per ABN-602.

B. Correct. When 1EA1 is de-energized, the EDG would auto start, the output breaker would automatically close and the BOS would automatically start the prescribed loads.

C. Incorrect. Plausible because when 1EA1 is de-energized, the EDG would auto start, the output breaker would automatically close and the BOS would automatically start the prescribed loads.

However, if the EDG output breaker were to remain open, the operator may manually close the breaker and then the BOS would load the bus.

D. Incorrect. Plausible because when 1EA1 is de-energized, the EDG would auto start, the output breaker would automatically close and the BOS would automatically start the prescribed loads.

However, if the EDG output breaker were to remain open, the operator may manually close the breaker and if the BOS sequencer failed to run, the operator may manually load the bus per ABN-602.

Technical Reference(s) ABN-602, Section 2.2 Attached w/ Revision # See ABN-602, Steps 2.3.4 & 2.3.5 Comments / Reference Proposed references to be provided during examination: None Learning Objective: COMPREHEND the normal, abnormal and emergency operation of the Emergency Diesel Generator System.

Question Source: Bank #

Modified Bank # X (Note changes or attach parent)

New Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 5, 7 55.43 Page 71 of 93 CPNPP NRC 2013 RO Written Exam Worksheet 1 to 25 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ABN-602, Section 2.2 Revision # 8 Page 72 of 93 CPNPP NRC 2013 RO Written Exam Worksheet 1 to 25 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ABN-602, Step 2.3.4 Revision # 8 Page 73 of 93 CPNPP NRC 2013 RO Written Exam Worksheet 1 to 25 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ABN-602, Step 2.3.4 Revision # 8 Page 74 of 93 CPNPP NRC 2013 RO Written Exam Worksheet 1 to 25 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ABN-602, Step 2.3.4 Revision # 8 Page 75 of 93 CPNPP NRC 2013 RO Written Exam Worksheet 1 to 25 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ABN-602, Step 2.3.5 Revision # 8 Page 76 of 93 CPNPP NRC 2013 RO Written Exam Worksheet 1 to 25 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From CPNPP Exam Bank Revision # 01/23/00 Given the following conditions:

• Work is in progress to place XST2A, Spare 345 kV Startup Transformer, in service.

• XST2, 345 kV Startup Transformer, is currently de-energized to support the work activities.

• 1EA1, 6.9kV Safeguards Bus, is being supplied from XST1, 345 kV Startup Transformer, via 1EA1-2, 6.9 kV Incoming Breaker.

With this configuration, if a sudden pressure fault were to occur on XST1, Bus 1EA1 would be load shed, Emergency Diesel Generator (EDG) 1-01 would start, 1EG1, EDG 1-01 Output Breaker would ...

A. automatically close, and the Blackout Sequencer (BOS) would load Bus 1EA1.

B. automatically close, but Bus 1EA1 would have to be manually loaded by the operators.

C. remain open but could be manually closed by the operators, and Bus 1EA1 would have to be manually loaded by the operators.

D. remain open and could NOT be manually closed by the operators, and Bus 1EA1 would remain de-energized.

Page 77 of 93 CPNPP NRC 2013 RO Written Exam Worksheet 1 to 25 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 073 K3.01 Importance Rating 3.6 Process Radiation Monitoring System: Knowledge of the effect that a loss or malfunction of the PRM system will have on the following: Radioactive effluent releases Proposed Question: Common 22 Given the following conditions on Unit 1:

• Turbine Building Sumps are being released per STA-603-16, Secondary Waste Release Data Sheet.

• 1-RE-5100 (TBD172) TURBINE BUILDING DRAINS RADIATION MONITOR has experienced a loss of power.

Which of the following actions will occur?

A. 1-RV-5100A, TURB BLDG SMP 1-02 DISCH DRN HDR TO LVW/EVAP POND ISOL VLV will CLOSE and 1-RV-5100B, TURB BLDG SMP 1-02 DISCH HDR TO WWHT ISOL VLV will CLOSE.

B. 1-RV-5100A, TURB BLDG SMP 1-02 DISCH DRN HDR TO LVW/EVAP POND ISOL VLV will OPEN and 1-RV-5100B, TURB BLDG SMP 1-02 DISCH HDR TO WWHT ISOL VLV will CLOSE.

C. 1-RV-5100A, TURB BLDG SMP 1-02 DISCH DRN HDR TO LVW/EVAP POND ISOL VLV will CLOSE and 1-RV-5100B, TURB BLDG SMP 1-02 DISCH HDR TO WWHT ISOL VLV will OPEN.

D. 1-RV-5100A, TURB BLDG SMP 1-02 DISCH DRN HDR TO LVW/EVAP POND ISOL VLV will OPEN and 1-RV-5100B, TURB BLDG SMP 1-02 DISCH HDR TO WWHT ISOL VLV will OPEN.

Proposed Answer: C Page 78 of 93 CPNPP NRC 2013 RO Written Exam Worksheet 1 to 25 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible if thought that all discharge will isolate on a loss of power to the radiation monitor, however the loss of power to the monitor has the same effect as a high radiation signal which would cause 1-RV-5100A to close isolating the Low Volume Waste flowpath and 1-RV-5100B to open aligning the Co-Current Waste flowpath.

B. Incorrect. Plausible if thought normal release flowpath is to Co-Current Waste, however 1-RV-5100A will close isolating the Low Volume Waste flowpath and 1-RV-5100B will open aligning the Co-Current Waste flowpath.

C. Correct. The normal release path is to Low Volume Waste, so1-RV-5100A will close isolating the Low Volume Waste flowpath and 1-RV-5100B will open aligning the Co-Current Waste flowpath.

D. Incorrect. Plausible if thought that 1-RV-5100A and 1-RV-5100B open in response to the loss of power to the radiation monitor which controls valve positions.

Technical Reference(s) ALM-3200 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.

STATE the functions and EXPLAIN the design criteria of the Waste Management System.

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

New Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 11, 13 55.43 Page 79 of 93 CPNPP NRC 2013 RO Written Exam Worksheet 1 to 25 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ALM-3200 Revision 4 Page 80 of 93 CPNPP NRC 2013 RO Written Exam Worksheet 1 to 25 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 073 G 2.4.6 Importance Rating 3.7 Process Radiation Monitoring System: Emergency Procedures/Plan: Knowledge of EOP mitigation strategies Proposed Question: Common 23 Given the following conditions:

• Unit 1 is in MODE 3.

• The Main Steam Isolation Valves are CLOSED.

• Reactor Coolant System temperature is being maintained by steaming through the Atmospheric Relief Valves.

• Steam Generator Blowdown is being maintained at maximum for Chemistry control.

• A Safety Injection actuates on Low Pressurizer Pressure.

When performing EOP-0.0A, Reactor Trip or Safety Injection, which of the following Process Radiation Monitor Trends would be used in determining that a Steam Generator Tube Rupture was the initiating event for the Safety Injection?

A. Condenser Off Gas B. Steam Generator Blowdown Sample C. Main Steamline Radiation D. Main Steamline N-16 Radiation Proposed Answer: B Page 81 of 93 CPNPP NRC 2013 RO Written Exam Worksheet 1 to 25 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible because the Condenser Off Gas Monitor is commonly used as indication of a Steam Generator Tube Rupture, however with the MSIVs closed the radioactive steam is bypassing the condenser and the trend would not aid is identifying a SGTR.

B. Correct. In accordance with EOP-0.0A the Steam Generator Blowdown Sample Radiation Monitor would indicate an increasing trend until either reaching the point that the monitor isolates Steam Generator Blowdown or the Safety Injection isolated the Steam Generator Blowdown.

C. Incorrect. Plausible because the Main Steamline Radiation Monitors are commonly used as indication of a Steam Generator Tube Rupture, however with the MSIVs closed the radioactive steam is bypassing the Main Steamline Monitors and the trend would not aid is identifying a SGTR.

D. Incorrect. Plausible because the Main Steamline N-16 monitors are the most sensitive during full power operation, however in MODE 3, no N-16 is being produced. Thus the Main Steamline N-16 Monitors would not aid in identifying a SGTR.

Technical Reference(s) EOP-0.0A, Step 13 Attached w/ Revision # See LO21.SYS.MR1 Comments / Reference LO21.SYS.SB1 Proposed references to be provided during examination: None Learning Objective: COMPREHEND the normal, abnormal and emergency operation of the Component Cooling Water System.

EXPLAIN the instrumentation and controls of the Digital Radiation Monitoring 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 5, 11 55.43 Page 82 of 93 CPNPP NRC 2013 RO Written Exam Worksheet 1 to 25 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From EOP-0.0A, Step 13 Revision # 8 Page 83 of 93 CPNPP NRC 2013 RO Written Exam Worksheet 1 to 25 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From LO21.SYS.MR1 p. 32 Revision 6-9-2011 Main Steam Line Radiation Monitors To detect SG primary-to-secondary tube leakage, each Main Steam line is equipped with a two types of adjacent-to-line (ATL) radiation monitors. Using different detector types and discrimination settings, the monitors are capable of detecting a range of very small to very large leakage amounts. The first monitor type meets the accuracy and range requirements of Regulatory Guide 1.97 and detects gross gamma radiation emitted during radioactive decay of noble fission product gases, particularly Xe-133.

The second monitor type detects N-16 gamma radiation produced during reactor operation.

Located between the ARV and the first SG safety valve of each steam line, Main Steam Line Monitors (u-RE-2325 u-RE-2328) are Geiger Mueller type detectors. Because direct measurement of the steam beta radiation activity concentration is not possible, Main Steam Line Monitors measure the equivalent gamma radiation dose rates adjacent to the steam line. Gamma radiation is given off during beta minus decay of noble gases such as Xe-133. To prevent area (background) radiation from affecting the detector response and indication, the detector is mounted inside a 3 1/4" lead shield with a thin steel window facing the steam pipe. The major limitation of these monitors is they are not sensitive to small leak rate changes and are therefore limited to post-accident assessment of significant releases. They are only capable of detecting primary-to-secondary leakage beginning at around 3600 gallons per day (2.5 gpm). Two Main Steam Line Monitors interface with a single remote RM-80 microprocessor.

Located just upstream of the MSIV for each steam line, the Steam Generator Leak Rate Monitors (u-RE-2325A u-RE-2328A) is an insulated, lightly shielded gamma scintillator type detector, sensitive to N-16 at 6.13 MeV. By utilizing the response of the N-16 gamma, this monitor provides early detection of very small amounts of primary-to-secondary leakage in the gallons-per-day (gpd) range. If primary-to-secondary leakage or a Steam Generator Tube Rupture (SGTR) were to occur while the reactor is at power, this monitor provides a rapid and clear indication of a primary-to-secondary leak. Once the reactor trips, production of N-16 essentially ceases, so readouts typically return to their normal levels.

This radiation monitor is capable of detecting SG tube leakage ranging from about 1 gallon per day

(.0007 gpm) up to 150 gallons per day. Two Steam Generator Leak Rate Monitors interface with a single remote RM-80 microprocessor.

Page 84 of 93 CPNPP NRC 2013 RO Written Exam Worksheet 1 to 25 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From LO21.SYS.SB1 p. 27 Revision 5-7-2011 EFFECTS ON SG BLOWDOWN The effects of a Steam Generator Tube Leak/Rupture on the SG Blowdown System are significant. If activity levels reach the threshold limit for radiation monitor u-RE-4200, then flow through the system is stopped by the SG Blowdown Isolation Valves closing. This prevents the SG Blowdown System from spreading the radioactive elements to other secondary plant components. Another effect is exhaustion of the resin in the systems demineralizers. The SG Blowdown Systems demineralizers utilize resin specifically designed to remove ionic impurities found in the secondary plant water. Introduction of reactor coolant into the blowdown stream provides the demineralizers access to additional ionic impurities and boron. These additional ionic impurities and boric acid cause rapid depletion of the resin.

Once the resin is exhausted, radioactivity levels increase in the blowdown stream. When radioactivity levels reach 1.0 X 10-5 µci/cm3, an annunciator on the Main Control Board warns the operator of a trouble alarm on the SG Blowdown Control Panel. Prior to receiving the annunciator alarm, the PC-11 will sound an audible alarm and provide information of the radiation monitor in "ALERT" status.

Shortly after these alarms, flow through the SG Blowdown System isolates when the radiation monitor reaches its "ALARM" setpoint.

Page 85 of 93 CPNPP NRC 2013 RO Written Exam Worksheet 1 to 25 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 076 A4.04 Importance Rating 3.5 Service Water System: Ability to manually operate and/or monitor in the control room: Emergency heat loads Proposed Question: Common 24 Given the following conditions:

• Unit 1 is operating at 100% power.

• Station Service Water Pump 1-01 trips.

• The crew is securing affected components per ABN-501, Station Service Water System Malfunction.

• Component Cooling Water Heat Exchanger 1-01 outlet temperature is 95°F and stable.

Which of the following components are secured using ABN-501, Station Service Water System Malfunction?

1. Centrifugal Charging Pump 1-01

2. Emergency Diesel Generator 1-01

3. Component Cooling Water Pump 1-01

4. Containment Spray Pumps 1-01 & 1-03

5. Safety Injection Pump 1-01

6. Safety Chiller 1-05 A. 1, 5, 6 B. 2, 3, 4 C. 2, 4, 6 D. 1, 2, 5 Proposed Answer: D Page 86 of 93 CPNPP NRC 2013 RO Written Exam Worksheet 1 to 25 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible because CCP 1-01 and SIP 1-01 must be secured and Safety Chiller 1-05 is secured if CCWP 1-01 trips.

B. Incorrect. Plausible because EDG 1-01 and CTPs 1-01 & 1-03 must be secured, CCWP 1-01 may be secured if CCW HX outlet temperature exceeds 122°F.

C. Incorrect. Plausible because EDG 1-01, CTPs 1-01 & 1-03 and Safety Chiller 1-05 is secured if CCWP 1-01 trips.

D. Correct. CCP 1-01, EDG 1-01 and SIP 1-01 must be secured.

Technical Reference(s) ABN-501, Steps 2.3.1 & 2.3.5 Attached w/ Revision # See ABN-502, Step 2.3.6 Comments / Reference Proposed references to be provided during examination: None Learning Objective: ANALYZE the response to Station Service Water Pump Trip per 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 X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7, 8 55.43 Page 87 of 93 CPNPP NRC 2013 RO Written Exam Worksheet 1 to 25 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ABN-501, Step 2.3.1 Revision # 9 Page 88 of 93 CPNPP NRC 2013 RO Written Exam Worksheet 1 to 25 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ABN-502, Step 2.3.5 Revision # 9 Page 89 of 93 CPNPP NRC 2013 RO Written Exam Worksheet 1 to 25 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ABN-502, Step 2.3.6 Revision # 6 Page 90 of 93 CPNPP NRC 2013 RO Written Exam Worksheet 1 to 25 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 076 K1.15 Importance Rating 2.5 Service Water System: Knowledge of the physical connections and/or cause-effect relationships between the SWS and the following systems: FPS Proposed Question: Common 25 Given the following conditions on Unit 1:

• Unit 1 is tripped due to a normal plant shutdown.

• Station Service Water Pump 1-01 was tagged out 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> ago due to the pump bearing overheating.

• Station Service Water Pump 1-02 tripped on overcurrent.

• During the recovery actions it becomes necessary to supply cooling water to the Unit 1 Safety Injection Pumps, Containment Spray Pumps, and Centrifugal Charging Pumps.

Which of the following answers complete the statement below per ABN-501, Station Service Water Malfunctions?

IF [1] can NOT supply Unit 1 Station Service Water, THEN supply cooling water to essential equipment from [2].

[1] [2]

A. Fire Protection Water Unit 2 Station Service Water B. Unit 2 Station Service Water Fire Protection Water C. Fire Protection Water Demineralized Water D. Unit 2 Station Service Water Demineralized Water Proposed Answer: B Explanation:

A. Incorrect. Plausible because Unit 2 Station Service Water is the preferred source and fire protection water is the supply the required loads.

B. Correct. As outlined in ABN-501.

C. Incorrect. Plausible because fire protection water is the secondary source, however, demineralized water is insufficient to supply the required loads D. Incorrect. Plausible because Unit 2 Station Service Water is the preferred source, however, demineralized water is insufficient to supply the required loads.

Page 91 of 93 CPNPP NRC 2013 RO Written Exam Worksheet 1 to 25 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Technical Reference(s) ABN-501, Step 5.3.7 Attached w/ Revision # See ABN-501, Attachment 1 Comments / Reference Proposed references to be provided during examination: None Learning Objective: ANALYZE the response to Station Service Water Pump Trip per ABN-501, Station Service Water System Malfunction.

Question Source: Bank # X 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, 8 55.43 Comments /

Reference:

From ABN-501, Step 5.3.7 Revision # 9 Page 92 of 93 CPNPP NRC 2013 RO Written Exam Worksheet 1 to 25 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ABN-501, Attachment 1 Revision # 9 Page 93 of 93 CPNPP NRC 2013 RO Written Exam Worksheet 1 to 25 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 078 K2.01 Importance Rating 2.7 Instrument Air System: Knowledge of bus power supplies to the following: Instrument air compressor Proposed Question: Common 26 What is the power supply for Instrument Air Compressor 1-01?

A. 1EB3-1 B. 1EB4-1 C. 1EB3 D. 1EB4 Proposed Answer: C Explanation:

A. Incorrect. Plausible because this is the control power supply to 1-01 Instrument Air Dryer, however, the power supply to Instrument Air Compressor 1-01 is 1EB3.

B. Incorrect. Plausible because this is the control power supply to 1-02 Instrument Air Dryer, however, the power supply to Instrument Air Compressor 1-01 is 1EB3.

C. Correct. This is the power supply to Instrument Air Compressor 1-01.

D. Incorrect. Plausible because this is the power supply to Instrument Air Compressor 1-02.

Technical Reference(s) LO21.SYS.IA1 Attached w/ Revision # See Comments / Reference Proposed references to be provided during examination: None Learning Objective: DEMONSTRATE an understanding of the components of the Instrument Air System including interrelations with other systems to include interlocks and control loops.

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

New Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis Page 1 of 92 CPNPP NRC 2013 RO Written Exam Worksheet 26 to 50 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 10 CFR Part 55 Content: 55.41 5, 7 55.43 Comments /

Reference:

From LO21.SYS.IA1.LN, Page 23 Revision # 05/07/11 ELECTRICAL The following table provides the source of power for the major components of the Instrument Air system:

Component Component Power Supply Control Power 1-01 Instrument Air Compressor 1EB3/11D 1EB3/11D 1-02 Instrument Air Compressor 1EB4/11D 1EB4/11D X-01 Instrument Air uB4/11C uB4/11C Compressor X-02 Instrument Air XB1/5C XB1/5C Compressor 2-01 Instrument Air Compressor 2EB3/10C 2EB3/10C 2-02 Instrument Air Compressor 2EB4/10D 2EB4/10D 1-01 Instrument Air Dryer --- 1EB3-1/2BR 1-02 Instrument Air Dryer --- 1EB4-1/12FL X-01 Instrument Air Dryer --- XB2-1/3M X-02 Instrument Air Dryer --- XB1-6/4RB 2-01 Instrument Air Dryer --- 2EB3-1/2BR 2-02 Instrument Air Dryer --- 2EB4-2/9FL Page 2 of 92 CPNPP NRC 2013 RO Written Exam Worksheet 26 to 50 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 078 K3.03 Importance Rating 3.0 Instrument Air System: Knowledge of the effect that a loss or malfunction of the IAS will have on the following: Cross-tied units Proposed Question: Common 27 Given the following conditions:

• In response an instrument air malfunction Unit 1 and Unit 2 instrument air headers are cross-connected in accordance with SOP-509A, Instrument Air System.

• Instrument Air compressors 1-01 and 1-02 are not operable and instrument air compressor X-01 is aligned to Unit 1 running at reduced capacity.

• The air leak has NOT been found.

• Instrument Air compressors 2-01, 2-02 and X-02 are all aligned to Unit 2 as follows;

• Instrument Air Compressor 2 STBY

• Instrument Air Compressor 2 LEAD

• Instrument Air Compressor X STBY

• 2-CI-0050, INST AIR RCVR 2-01 U1 XTIE VLV and 1-CI-0051, INST AIR RCVR 1-01 U2 XTIE VLV are open.

• Instrument Air header pressure dropped to 83 psig and then rose to 111 psig and has been stable for 22 minutes.

Which of the following describes the status of the Unit 2 Instrument Air compressors?

Instrument Air compressor 2-01 is __________________, Instrument Air compressor 2-02 is

__________________, and Instrument Air compressor X-02 is __________________.

A. running loaded running loaded running loaded B. running unloaded running unloaded running unloaded C. not running running unloaded not running D. not running running loaded not running Proposed Answer: A Page 3 of 92 CPNPP NRC 2013 RO Written Exam Worksheet 26 to 50 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Explanation:

A. Correct. Instrument air header pressure dropping to 83 psig causes IAC 2-02 to load at 105 psig, IAC 2-01 to start and load at 100 psig and IAC X-02 to start and load at 95 psig. All IACs must reach 115 psig before they unload or stop.

B. Incorrect. Plausible if thought that 110 psig is the unloading pressure and that automatic shutdown does not occur until 20 minutes of running unloaded (based on Unit difference).

C. Incorrect. Plausible if thought that 110 psig is the unloading pressure and the LEAD compressor (2-02) would remain running unloaded.

D. Incorrect. Plausible if thought that 110 psig and 20 minutes would stop IACs 2-01 and X-02 with IAC 2-02 running loaded as the LEAD compressor.

Technical Reference(s) LO21.SYS.IA1 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 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 Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 5, 7 55.43 Page 4 of 92 CPNPP NRC 2013 RO Written Exam Worksheet 26 to 50 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From LO21.SYS.IA1.LN, Page 13 Revision # 05/07/11 Page 5 of 92 CPNPP NRC 2013 RO Written Exam Worksheet 26 to 50 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 103 G 2.2.42 Importance Rating 3.9 Containment System: Equipment Control: Ability to recognize system parameters that are entry- level conditions for Technical Specifications Proposed Question: Common 28 Which of the following would cause entry into the Containment Isolation Valves Limiting Condition for Operation while in MODE 1?

Damage to the ______________________, which will NOT allow the valve to close.

A. stem for 1-8880, SI/PORV ACCUM N2 ISOL VLV B. operator for 1-LCV-459, U1 LTDN ISOL VLV C. stem for 1MS-0357, SG 1-03 BLDN DNSTRM ISOL VLV D. operator for 1-FCV-0510, SG 1-01 FW FLO CTRL VLV Proposed Answer: A Explanation:

A. Correct. This valve is a Containment Penetration boundary outside of Containment.

B. Incorrect. Plausible because this valve is part of the CVCS System, however, there are 2 automatic downstream isolation valves (1/1-8160 & 8152) providing Containment Isolation.

C. Incorrect. Plausible because this valve is part of the Blowdown System, however, there are 2 automatic upstream isolation valves (1-HV-2399 & 2399A) providing Containment Isolation.

D. Incorrect. Plausible because this valve is a Feedwater Isolation Valve under its own Tech Spec, 3.7.3, which is tied to Tech Spec 3.6.3 in the NOTE above the line.

Technical Reference(s) Technical Specification LCO 3.6.3 Attached w/ Revision # See OPT-408A, Attachment 10.1.1 Comments / Reference Proposed references to be provided during examination: None Learning Objective: APPLY the administrative requirements of the Containment System including Technical Specifications, TRM and ODCM.

Page 6 of 92 CPNPP NRC 2013 RO Written Exam Worksheet 26 to 50 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Question Source: Bank # X 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:

From Technical Specification LCO 3.6.3 Amendment # 150 Page 7 of 92 CPNPP NRC 2013 RO Written Exam Worksheet 26 to 50 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From OPT-408A, Attachment 10.1.1, page 11 Revision 11 Comments /

Reference:

From OPT-408A, Attachment 10.1.1, page 6 Revision 11 Comments /

Reference:

From OPT-408A, Attachment 10.1.1, page 33 Revision 11 Page 8 of 92 CPNPP NRC 2013 RO Written Exam Worksheet 26 to 50 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 2 K/A # 001 K6.12 Importance Rating 2.9 Control Rod Drive System: Knowledge of the effect that a loss or malfunction of the following will have on the CRDS:

Location and interpretation of CRDS AC/DC status alarms Proposed Question: Common 29 Given the following conditions:

• Unit 1 is at 100% power.

• Annunciator 1-ALB-10B, Window 4.12 - 480V ANY NON-1E BUS VOLT LOSS is in alarm.

• The following conditions are report by the Field Support Supervisor:

• CV-1MG1 & CV-1MG2, GROUND PROTECTION RELAY flags are NOT actuated.

• CRDM Generator 1-01 Motor Breaker is closed.

• CRDM Generator 1-02 Motor Breaker is open.

• Both MG 1 and 2 DIRECTIONAL OVERCURRENT A and B relays are NOT actuated.

• MG GENERATOR OVERVOLTAGE TRIP light NOT lit.

• CRDM Generator 1-01 Generator Breaker is closed.

• CRDM Generator 1-02 Generator Breaker is open.

• CRDM GENERATOR LINE VOLTS are 261 volts and stable.

Which of the following describes the effect on the Control Rod Drive System?

The Reactor is [1] due to the loss of 480 Volt Bus [2].

[1] [2]

A. tripped 1B3 B. not tripped 1B3 C. tripped 1B4 D. not tripped 1B4 Proposed Answer: D Page 9 of 92 CPNPP NRC 2013 RO Written Exam Worksheet 26 to 50 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible if believed that the reactor trips with the loss of one CRDS MG and that 480 V Bus 1B3 powers CRDS MG 1-02.

B. Incorrect. Plausible because the Reactor is not tripped and if believed that that 480 V Bus 1B3 powers CRDS MG 1-02.

C. Incorrect. Plausible because 480 V Bus 1B4 powers CRDS MG 1-02 but the Reactor is not tripped..

D. Correct. Annunciator 1-ALB-10B, Window 4.12 monitors the status of AC power to the CRDM Generators. When Bus 1B4 is tripped, the supply breaker to CRDM Generator 1-02 is also tripped which results in a motor breaker OPEN indication as listed in the Stem. Given the conditions listed, the Reactor is not tripped but CRDM Generator 1-02 is lost with Bus 1B4 de-energized.

Technical Reference(s) LO21.SYS.CR1 Attached w/ Revision # See ALM-0102A, 1-ALB-10B, Window 4.12 Comments / Reference Proposed references to be provided during examination: None Learning Objective: EXPLAIN the instrumentation and controls of the Rod Control System and PREDICT the system response.

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 5, 6 55.43 Comments /

Reference:

From LO21.SYS.CR1.LN, Page 24 Revision # 05/02/11 MOTOR GENERATORS (MGS)

Power for the CRDMs is supplied by two motor-generator sets. They are each comprised of a three phase induction motor directly coupled to a solid steel flywheel and a synchronous alternator. Each MG is operating from separate 480-volt, three-phase buses (uB3 and uB4). The generators are paralleled through Westinghouse type DB-416 circuit breakers. Each generator is a synchronous type, rated at 438 KVA, 260 VAC phase to phase, 150 VAC phase to neutral, zig-zag-wye connected, with brushless excitation from a static voltage regulator and 58.5-59.7 Hertz. The generator output breakers are in the bottom of the MG set switchgear panel. Control switches are on the panel and an automatic synchronizing circuit is installed inside.

Page 10 of 92 CPNPP NRC 2013 RO Written Exam Worksheet 26 to 50 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From LO21.SYS.CR1.LN, Page 25 Revision # 05/02/11 Motor Generator Set Control Panel These panels are located in the Safeguards Building 832 level near the Rod Control Logic and Power Cabinets. Each houses the associated Motor Generator output breakers. The associated controls for the generator output breaker as well as the motor generator motor breaker are also on the MG set panel.

When both output breakers are open, either will close when operated by their respective handswitch.

The second must always be closed by the synchronizer. Either breaker may be tripped by its control switch, overexcitation, or by phase overcurrent.

Comments /

Reference:

From ALM-0102A, 1-ALB-10B, Window 4.12 Revision # 12 Page 11 of 92 CPNPP NRC 2013 RO Written Exam Worksheet 26 to 50 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 2 K/A # 011 A1.03 Importance Rating 2.8 Pressurizer Level Control System: Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the PZR LCS system controls including: VCT level Proposed Question: Common 30 Given the following conditions:

• Unit 1 is at 100% power.

• Centrifugal Charging Pump (CCP) 1-01 is running.

• Letdown is established at 120 gpm.

• 1-LCV-112A, VCT LVL CTRL VLV, is in AUTO position.

• Actual Volume Control Tank (VCT) level is 58% and lowering.

• VCT level channel 1-LT-112 fails high.

Which of the following predicts the impact on Pressurizer Level Control?

Pressurizer Level will...

A. ...remain stable as Reactor Makeup initiates automatic makeup at 46% and results in VCT level cycling between 46% and 56%; CCP 1-01 maintains Net Positive Suction Head.

B. ...NOT remain stable as Reactor Makeup initiates automatic makeup at 46% but VCT level will continue to lower; CCP 1-01 will lose Net Positive Suction Head as suction will NOT automatically shift to the Refueling Water Storage Tank.

C. ...remain stable as Reactor Makeup does NOT initiate automatic makeup; CCP 1-01 maintains Net Positive Suction Head as suction automatically shifts to the Refueling Water Storage Tank.

D. ...NOT remain stable as Reactor Makeup does NOT initiate automatic makeup; CCP 1-01 will lose Net Positive Suction Head as suction will NOT automatically shift to the Refueling Water Storage Tank.

Proposed Answer: D Page 12 of 92 CPNPP NRC 2013 RO Written Exam Worksheet 26 to 50 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible because automatic makeup would be initiated when 1-LT-112 level channel lowered to 46% and the cycling setpoint is correct, however, 1-LT-112 will continue to indicate 100% VCT level. Thus automatic makeup will not occur and maintain VCT level in the band.

B. Incorrect. Plausible because automatic makeup would be initiated when 1-LT-112 level channel lowered to 46%, however, charging flow is exceeding automatic makeup capability and therefore VCT level could be thought to continue trending down even with automatic makeup, however, with RCP Seal Return flow included VCT level would not continue to drop if automatic makeup to the VCT was occurring..

C. Incorrect. Plausible because automatic makeup will not be initiated, however, the Centrifugal Charging Pump suction remains aligned to the VCT as both 1-LT-112 and 1-LT-185 must indicate 2% for suction to swap to the RWST and 1-LT-112 is failed to 100%.

D. Correct. As outlined in ALM-0061A, VCT level will not actuate at 46% as 1-LT-112 is failed to 100%. Additionally, CCP suction will not transfer to the RWST as both 1-LT-112 and 1-LT-185 must indicate 2% for suction to swap to the RWST and 1-LT-112 is failed to 100%. CCP 1-01 will lose suction and Pressurizer Level will lower dramatically.

Technical Reference(s) ALM-0061A, 1-ALB-6A, Window 2.5 & 4.5 Attached w/ Revision # See LO21.SYS.CS1 Comments / Reference Proposed references to be provided during examination: None Learning Objective: DEMONSTRATE an understanding of the components of the Chemical and Volume Control 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 5, 7 55.43 Page 13 of 92 CPNPP NRC 2013 RO Written Exam Worksheet 26 to 50 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ALM-0061A, 1-ALB-6A, Window 2.5 Revision # 7 Page 14 of 92 CPNPP NRC 2013 RO Written Exam Worksheet 26 to 50 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ALM-0061A, 1-ALB-6A, Window 4.5 Revision # 7 Page 15 of 92 CPNPP NRC 2013 RO Written Exam Worksheet 26 to 50 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From LO21.SYS.CS1 LN p. 24, 25 Revision 4/28/11 Table 5 summarizes the functions initiated at various level setpoints on u-LT-0112 and u-LT-0185.

Table 1: VCT Level Transmitter Automatic Functions LVL u-LT-0112 u-LT-0185 BOTH u-LT-0112 AND

(%) u-LT-0185 98 --- LCV to HUT ---

high-high level alarm 70 high level alarm high level alarm ---

62 modulate LCV --- ---

56 stop auto MU --- ---

46 start auto MU --- ---

16 low level alarm and low level alarm and ---

allow chg realign allow chg realign 2 low-low level alarm low-low level alarm ---

2 --- --- align charging to RWST At 62% indicated level on u-LI-0112 (or at a value set on the controller), VCT Level Controller u-LK-0112C (a M/A station on CB-06) will send a signal to an I/P converter, which pneumatically modulates u-LCV-0112A to direct letdown to the recycle holdup tank. The valve will continue to position further to the recycle holdup tank if level increases above setpoint. At 98% level from the other level transmitter, u-LT-0185, a solenoid operated valve directs full air pressure to the valve positioner to fully position u-LCV-0112A to the recycle holdup tank.

The 62% setpoint for initiating diversion of letdown to the recycle holdup tank is based on maintaining a gas volume in the volume control tank sufficient to absorb the liquid which results from a 4F error in Tave during a 0 to 100% power swing and which is not absorbed by the pressurizer.

Volume Control Tank Level Transmitter u-LT-0112A actuates level bistables which initiate automatic makeup to the volume control tank at 46% level and stop automatic makeup at 56% level.

At 2% level (low-low level) on both u-LT-0112 and u-LT-0185, the charging pump suction piping will automatically realign to the refueling water storage tanks. The low-low level realignment opens the parallel charging pump suction valves from the RWST (u-LCV-0112D and u-LCV-0112E) and closes the series charging pump suction valves from the VCT (u-LCV-0112B and u-LCV-0112C). The charging pump suction may be manually realigned after the volume control tank level returns to > 16%

on either channel.

Page 16 of 92 CPNPP NRC 2013 RO Written Exam Worksheet 26 to 50 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 2 K/A # 014 A2.06 Importance Rating 2.6 Rod Position Indication System: Ability to (a) predict the impacts of the following malfunctions or operations on the RPIS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Loss of LVDT Proposed Question: Common 31 Given the following conditions:

• Unit 1 is at 100% power.

• The 120 VAC Distribution System is in the Normal alignment.

• A failure of 120/208 VAC DISTRIBUTION PANEL 1C1 (CP1-ECDPNC-02) occurs.

What is the effect on the Digital Rod Position Indication (DRPI) system and the procedural action necessary?

A. DRPI is in Half-Accuracy. Energize Train C 120 VAC Distribution Panel 1C14 from CP1-ECDPNC-03, 120/208 VAC DISTRIBUTION PANEL 1C4.

B. Loss of DRPI. Energize Train C 120 VAC Distribution Panel 1C14 from CP1-ECDPNC-03, 120/208 VAC DISTRIBUTION PANEL 1C4.

C. DRPI is in Half-Accuracy. Energize Train C 120 VAC Distribution Panel 1C14 from CPX-EPDPNB-03 120/208 VAC MISCELLANEOUS POWER PANEL XC4-4.

D. Loss of DRPI. Energize Train C 120 VAC Distribution Panel 1C14 from CPX-EPDPNB-03 120/208 VAC MISCELLANEOUS POWER PANEL XC4-4.

Proposed Answer: B Page 17 of 92 CPNPP NRC 2013 RO Written Exam Worksheet 26 to 50 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible if believed that Data A was powered from 1C1 and Data B was powered from 1C14 during normal alignment, however all of DRPI is powered from 1C14 which is normally powered from 1C1.

B. Correct. As DRPI is powered from 1C14, which is normally aligned from 1C1, a loss of 1C1 would result in a loss of DRPI. In accordance with ABN-712, the power supply should be shifted to 1C4.

C. Incorrect. Plausible if believed that Data A was powered from 1C1 and Data B was powered from 1C14 during normal alignment, however all of DRPI is powered from 1C14 which is normally powered from 1C1. Further XC4-4 could be thought to be the alternate power supply to 1C14.

D. Incorrect. Plausible as this is the correct response of DRPI, however, this is the incorrect alternate power supply and XC4-4 could be thought to be the alternate power supply to 1C14.

Technical Reference(s) ABN-712, Sections 4.3 Attached w/ Revision # See SOP-608A, Section 5.1.3 & 5.1.4 Comments / Reference Proposed references to be provided during examination: None Learning Objective: COMPREHEND the normal, abnormal and emergency operation of the Rod Control Indication and Rod Insertion Limit (RIL) Monitor 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 6, 10 55.43 Page 18 of 92 CPNPP NRC 2013 RO Written Exam Worksheet 26 to 50 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SOP-608A 5.1.4 Revision # 12-1 Page 19 of 92 CPNPP NRC 2013 RO Written Exam Worksheet 26 to 50 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ABN-712, Section 4.3 Revision # 10 Page 20 of 92 CPNPP NRC 2013 RO Written Exam Worksheet 26 to 50 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SOP-608A 5.1.3 Revision # 12-1 Page 21 of 92 CPNPP NRC 2013 RO Written Exam Worksheet 26 to 50 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 2 K/A # 017 K5.02 Importance Rating 3.7 In Core Temperature Monitor System: Knowledge of the operational implications of the following concepts as they apply to the ITM system: Saturation and subcooling of water Proposed Question: Common 32 Given the following conditions on Unit 1:

• A Small Break Loss of Coolant Accident has occurred.

• Reactor Coolant Pumps have been tripped.

• Reactor Vessel Level indicating System (RVLIS) currently indicates 11 and 22 inches above core plate light LIT. All other RVLIS lights DARK.

• The Emergency Core Cooling System is injecting

• Reactor Coolant System Wide Range Pressure is 1385 psig.

• All Wide Range Hot Leg Temperatures indicate 587ºF.

• All Core Exit Thermocouples indicate between 570ºF and 575ºF.

Which of the following correctly states the fluid properties of the water cooling the top of the core?

The top of the core is being cooled by A. saturated liquid.

B. superheated steam.

C. saturated steam.

D. subcooled liquid.

Proposed Answer: D Page 22 of 92 CPNPP NRC 2013 RO Written Exam Worksheet 26 to 50 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible if calculation is performed using hot leg temperature of 587°F and believed that Quality is 0.

B. Incorrect. Plausible if thought that the saturated conditions in the hot leg were indicative of superheated conditions at the hotter core outlet.

C. Incorrect. Plausible if thought that the saturated conditions in the hot legs were indicative of saturated conditions at the core outlet, however, as the fluid level is below the hot legs only steam is being carried over to the hot leg RTDs.

D. Correct. The Core Exit Thermocouples indicate 12 to 17ºF subcooled.

Technical Reference(s) LO21.SYS.RC3 Attached w/ Revision # See Comments / Reference Proposed references to be provided during examination: None Learning Objective: LIST the input signals to the Core Cooling Monitoring System and DESCRIBE how these signals are utilized in determining the thermodynamic condition of the RCS/Reactor Vessel fluid.

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, 14 55.43 Page 23 of 92 CPNPP NRC 2013 RO Written Exam Worksheet 26 to 50 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From LO21.SYS.RC3.LN, Page 14 Revision # 02/10/04 COMPONENTS CORE COOLING MONITOR SYSTEM (CCM)

Two qualified, redundant CCM's are used for Inadequate Core Cooling (ICC) monitoring. Each CCM is designed to indicate Core Exit Thermocouple temperatures (CET function) and to monitor the RCS Subcooling Margin Monitor (SMM function).

CORE EXIT THERMOCOUPLES To provide input temperature data to the CCM microprocessor, the NSSS-supplied array of fifty CET's has been divided into two separate, redundant trains with each set having a distribution representative of all four quadrants of the reactor core exit area. The planar locations of the CETs with respect to the core fuel assembly position are illustrated on Figure 2. All CET's are axially located just above the Upper Core Plate as illustrated in Figure3.

Each CET is a type K (chromel-alumel) thermocouple contained within an aluminum-oxide insulated, stainless steel sheathed cable (1/8" OD). Each cable passes through one of four vessel head penetrations (located 90º apart and near the core periphery) which contain pressure-boundary sealing assemblies. Figure 2 includes indication of head penetration assignments for the various T/C cables, separated into groups of either twelve or thirteen cables per penetration.

Above the vessel head, the CET cables are grouped into two separate trains. Each train is routed into a separate reference junction box which contains three platinum resistance temperature detectors (RTD's): two used for reference temperature measurements plus one installed spare. These reference measurements permit the transition from chromel-alumel leads to copper conductors for signal transmission to the CCM microprocessor (Figure 4).

The CET signals are used in the CCM to monitor coolant temperatures over the entire range including normal operating conditions and extending to beyond accident extremes. Each thermocouple is constantly checked, by the CCM computer, for open or shorted conditions, and the signal is adjusted to account for the inside containment cold reference junction conditions based on the reference RTD measurements. The highest valid CET signal is displayed on the Control Board and is also employed by the microprocessor to determine the RCS saturation margin.

Page 24 of 92 CPNPP NRC 2013 RO Written Exam Worksheet 26 to 50 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From Steam Tables Page 25 of 92 CPNPP NRC 2013 RO Written Exam Worksheet 26 to 50 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 2 K/A # 027 K2.01 Importance Rating 3.1 Containment Iodine Removal System: Knowledge of bus power supplies to the following: Fans Proposed Question: Common 33 Which 480 VAC Motor Control Centers supply power to the Containment Pre-Access Filtration System Fans?

Containment Pre-Access Containment Pre-Access Filtration Fan 1-01 Filtration Fan 1-02 A. 1B1-2 1B2-1 B. XB1-2 XB2-2 C. XEB2-1 XEB2-2 D. 1EB1-2 1EB2-2 Proposed Answer: D Explanation:

A. Incorrect. Plausible if thought that fans are powered from a unit non-safeguards power supply because they are normally operated prior to a containment entry to reduce radiation exposure.

B. Incorrect. Plausible if thought that fans are powered from a common non-safeguards power supply because they are normally operated prior to a containment entry to reduce radiation exposure.

C. Incorrect. Plausible if thought that the fans are powered from a common safeguards MCC because they could be used for post accident iodine removal.

D. Correct. Train A fan is powered from MCC 1EB1-2 and Train B from MCC 1EB2-2.

Page 26 of 92 CPNPP NRC 2013 RO Written Exam Worksheet 26 to 50 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Technical Reference(s) EOP-0.0A, Attachment 8 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 Containment 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 X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7, 9 55.43 Page 27 of 92 CPNPP NRC 2013 RO Written Exam Worksheet 26 to 50 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From EOP-0.0A, Attachment 8 Revision # 8 Page 28 of 92 CPNPP NRC 2013 RO Written Exam Worksheet 26 to 50 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From EOP-0.0A, Attachment 8 Revision # 8 Page 29 of 92 CPNPP NRC 2013 RO Written Exam Worksheet 26 to 50 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 2 K/A # 029 G 2.1.32 Importance Rating 3.8 Containment Purge System: Conduct of Operations: Ability to explain and apply all system limits and precautions Proposed Question: Common 34 Given the following conditions:

• Unit 1 is in MODE 6.

• Preparations for core off-load are in progress.

• The Fuel Transfer Tube Gate Valve is open.

• The Containment Equipment Hatch is removed.

• Both doors of the Personnel Airlock are open.

• The Control Room is starting a Containment Purge per SOP-801A, Containment Ventilation System.

• 1-HV-5572, CNTMT AIR PRG SPLY DMPR 1-01, will remain closed for the Containment Purge.

Which of the following describes the reason for leaving 1-HV-5572, CNTMT AIR PRG SPLY DMPR 1-01, closed during the Containment purge per SOP-801A, Containment Ventilation System?

A. Ensures water from the Spent Fuel Pools is NOT transferred to the Refueling Cavity.

B. Ensures air flow into Containment to prevent contaminating the Safeguards Building.

C. Ensures water from the Refueling Cavity is NOT transferred to the Spent Fuel Pools.

D. Ensures air flow into the Safeguards Building so it can be monitored by Vent Stack radiation monitors.

Proposed Answer: B Page 30 of 92 CPNPP NRC 2013 RO Written Exam Worksheet 26 to 50 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible because with the Fuel Transfer Tube Gate Valve open there is a potential for transferring water, however, there is insufficient differential pressure developed because both doors of the Personnel Airlock are open.

B. Correct. As outlined in the NOTE in SOP-801A, when the Equipment Hatch is off and the Personnel Airlock is open then 1-HS-5572 is closed to ensure that air flow is from the Safeguards Building into the Containment. This is contrary to the normal pressure within the Safeguards Building during MODES 1 through 4 when it is maintained at a slight negative pressure to ensure contaminants can be monitored via the Primary Plant Ventilation System.

C. Incorrect. Plausible because with the Fuel Transfer Tube Gate Valve open there is a potential for transferring water, however, there is insufficient differential pressure developed because both doors of the Personnel Airlock are open.

D. Incorrect. Plausible because this condition could exist with the Personnel Airlock open, however, only if 1-HS-5572 was also open.

Technical Reference(s) SOP-801A, Step 5.1.6.E NOTE & CAUTION Attached w/ Revision # See SOP-816, Section 4.1 Comments / Reference Proposed references to be provided during examination: None Learning Objective: COMPREHEND the normal, abnormal and emergency operation of the Containment Purge 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 5, 9 55.43 Page 31 of 92 CPNPP NRC 2013 RO Written Exam Worksheet 26 to 50 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SOP-801A, Step 5.1.6.E NOTE & CAUTION Revision # 14 Page 32 of 92 CPNPP NRC 2013 RO Written Exam Worksheet 26 to 50 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SOP-816, Section 4.1 Revision # 13 Page 33 of 92 CPNPP NRC 2013 RO Written Exam Worksheet 26 to 50 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 2 K/A # 035 K4.05 Importance Rating 2.9 Steam Generator System: Knowledge of the SGS design feature(s) and/or interlock(s) that provide for the following: Amount of reserve water in SG Proposed Question: Common 35 Given the following conditions on Unit 2:

• Main Feedwater Regulating Valve to Steam Generator 2-01 closed and the Unit tripped on Low-Low Steam Generator level.

What is the Steam Generator level setpoint for Auxiliary Feedwater Pump start and why is Auxiliary Feedwater automatically initiated?

A. 35.4% Narrow Range Level Provide a secondary heat sink.

B. 35.4% Narrow Range Level Prevent Steam Generator dryout.

C. 38% Narrow Range Level Provide a secondary heat sink.

D. 38% Narrow Range Level Prevent Steam Generator dryout.

Proposed Answer: A Page 34 of 92 CPNPP NRC 2013 RO Written Exam Worksheet 26 to 50 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Explanation:

A. Correct. Steam Generator level below 35.4% (U-2) initiates Auxiliary Feedwater to provide secondary heat removal.

B. Incorrect. Plausible because the Steam Generator level setpoint is correct but keeping the SG components wet is why a minimum of 100 gpm to each SG is required in ECA-2.1B.

C. Incorrect. Plausible because 38% is the Unit 1 setpoint which initiates Auxiliary Feedwater to provide secondary heat removal.

D. Incorrect. Plausible because 38% is the Unit 1 setpoint is correct but keeping the SG components wet is why a minimum of 100 gpm to each SG is required in ECA-2.1B.

Technical Reference(s) LO21.SYS.AF1 Attached w/ Revision # See ECA-2.1B Comments / Reference Technical Specification 3.3.2 Bases Proposed references to be provided during examination: None Learning Objective: COMPREHEND the normal, abnormal and emergency operation of the Auxiliary Feedwater System.

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

New Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 4, 7 55.43 Page 35 of 92 CPNPP NRC 2013 RO Written Exam Worksheet 26 to 50 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From LO21.SYS.AF1.LN, Page 11 Revision 05/11/11 When Control Room switches are inaccessible, manual operation from the Remote Shutdown Panel (RSP) is provided. Local manual control from the RSP overrides all other signals. Manual control is switched from control board to the RSP with transfer switches located on the Shutdown Transfer Panel (STP) (Train "A") or on the RSP (Train "B"). When control is transferred, an alarm for local override is actuated in the Control Room.

The MDAFWPs will automatically start due to (Figure 2):

• Low-low Steam Generator narrow range level at 38% ( 35.4% for Unit 2) in two out of four detectors on any one Steam Generator,

• Trip of both main feed pumps,

• Safety injection sequence signal (SI),

• Blackout (BO) sequence signal, or

• AMSAC signal Page 36 of 92 CPNPP NRC 2013 RO Written Exam Worksheet 26 to 50 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ECA-2.1B, Attachment 4 Revision 8 Comments /

Reference:

From Tech Spec 3.3.2 Bases Revision 67 Page 37 of 92 CPNPP NRC 2013 RO Written Exam Worksheet 26 to 50 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 2 K/A # 045 A3.05 Importance Rating 2.6 Main Turbine Generator System: Ability to monitor automatic operation of the MTG System, including: Electrohydraulic control Proposed Question: Common 36 Given the following conditions:

• Unit 1 Main Turbine startup is in progress as follows:

• Turbine Speed is 1800 rpm.

• Exhaust Hood temperature is 174°F.

• Turbine Stress Evaluator (TSE) Margin is GREEN.

• No operator action has been taken since establishing 1800 rpm.

What is the status of the LP Turbine Control Valves and the HP Turbine Control Valves at this point in Main Turbine startup?

LP Turbine Control Valves HP Turbine Control Valves A. FULLY OPEN FULLY OPEN B. NOT FULLY OPEN FULLY OPEN C. FULLY OPEN NOT FULLY OPEN D. NOT FULLY OPEN NOT FULLY OPEN Proposed Answer: C Explanation:

A. Incorrect. Plausible if thought that latching the turbine and raising speed to 1800 rpm (full speed) would open all control valves.

B. Incorrect. Plausible if thought that the HP vice the LP control valves would be fully open on turbine startup and the LP control valves throttle to control speed and the throttle to control load once synchronized to the Grid.

C. Correct. Given the conditions listed and with no operator action once the Main Turbine reaches 1800 RPM the LP control valves will be fully open and HP control valves throttle to control speed and will throttle to control load once synchronized to the Grid.

D. Incorrect. Plausible if thought that both the LP and HP control valves throttle to control load.

Technical Reference(s) LO21.SYS.MT1 Attached w/ Revision # See Comments / Reference Page 38 of 92 CPNPP NRC 2013 RO Written Exam Worksheet 26 to 50 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Proposed references to be provided during examination: None Learning Objective: LIST and DESCRIBE the purpose of in-plant and Control Room System controls, indications, and alarms for the Main Turbine components.

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

New Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 5, 7 55.43 Page 39 of 92 CPNPP NRC 2013 RO Written Exam Worksheet 26 to 50 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From LO21.SYS.MT1 Revision 5/4/11 The Turbine-Generator load is controlled by stop and control valves. The stop valves admit steam to the turbine as operating conditions require, and provide extremely fast closure to isolate the turbine from the steam supply system in the event of a protective device actuation. This rapid closure of the stop valves is referred to as a turbine trip. The control valves regulate the amount of steam admitted to the Main Turbine to control the output load of the Main Generator or the speed of the turbine during startup.

Steam flow to the LP Turbines is regulated by the EHC system which uses hydraulic fluid pressure to position control valves in response to various inputs much like the HP Stop and Control valves. The LP Stop and Control valves are normally fully open with Turbine-Generator load being controlled by the HP Turbine.

During normal operation the LP Control Valves are fully open and will only throttle down upon a large loss of electrical load to help prevent overspeeding of the Main Turbine.

Page 40 of 92 CPNPP NRC 2013 RO Written Exam Worksheet 26 to 50 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 2 K/A # 055 K1.06 Importance Rating 2.6 Condenser Air Removal System: Knowledge of the physical connections and/or cause-effect relationships between the CARS and the following systems: PRM system Proposed Question: Common 37 Given the following conditions on Unit 1:

• A Steam Generator tube leak is in progress.

• 1-RE-2959 (COG-182), CONDENSER OFF GAS Radiation Monitor is in service.

• Two Condenser Exhausting Vacuum Pumps are in service.

• A rapid plant shutdown is in progress.

Which of the following describes the flow path of fission products from the condenser?

A. A portion of the fission products will be monitored by 1-RE-2959 before being transmitted to the Primary Plant Ventilation System.

B. All of the fission products will be monitored by 1-RE-2959 before being transmitted to the Primary Plant Ventilation System.

C. A portion of the fission products will be monitored by 1-RE-2959 before being transmitted to the Waste Gas System.

D. All of the fission products will be monitored by 1-RE-2959 before being transmitted to the Waste Gas System.

Proposed Answer: A Explanation:

A. Correct. A bypass line from the main 10 discharge header goes to the Process Monitor. Once sampled at the monitor the bypass line rejoins the discharge header and is routed to the Primary Plant Ventilation System for a filtered and monitored discharge to the environment.

B. Incorrect. Plausible if believed that the entire discharge from the CEVs was monitored by 1-RE-2959, however only a portion of the discharge is monitored.

C. Incorrect. Plausible as only a portion of the fission products will be monitored, however, the CEV discharge is routed to the Primary Plant Ventilation System and cannot be routed to the Waste Gas System for storage and later discharge.

D. Incorrect. Plausible if believed that the entire discharge from the CEVs was monitored by 1-RE-2959, and if believed that the CEV discharge can be routed to the Waste Gas System.

Technical Reference(s) LO21.SYS.CV1 Attached w/ Revision # See Page 41 of 92 CPNPP NRC 2013 RO Written Exam Worksheet 26 to 50 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments / Reference Proposed references to be provided during examination: None Learning Objective: STATE the physical connections and EVALUATE the cause-effect relationships between the Condenser Vacuum and Water Box Priming System and the following systems, components or events:

• Main Turbine 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, 7, 14 55.43 Comments /

Reference:

From LO21.SYS.CV1.LN, Page 8 Revision # 05/25/11 Condenser Vacuum System Flow Path (Fig. 2)

Air and non-condensable gases are drawn from the main condenser shell thru the 8-inch piping and individual isolation valves to common isolation valve u-CV-0020. Air and non-condensable gases are drawn from the auxiliary condenser shells thru the 8-inch piping and individual isolation valves to common isolation valve u-CV-0022. These lines join to form the suction of the CEV pumps. Each pump discharges through its own seal water tank (Separator) and silencer to a common, 10" discharge header .Air and non-condensable gases in the discharge header are monitored for radiation by the condenser off-gas radiation monitor (u-RE-2959), located in a bypass line, and then discharged (in the Aux Building) to the Primary Plant Ventilation System.

Page 42 of 92 CPNPP NRC 2013 RO Written Exam Worksheet 26 to 50 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 2 K/A # 068 A4.04 Importance Rating 3.8 Liquid Radwaste System: Ability to manually operate and/or monitor in the control room: Automatic isolation Proposed Question: Common 38 Which of the following will cause an AUTO closure of X-RV-5253, Liquid Waste Processing System Discharge Isolation Valve while a release is in progress?

A. PC-11 channel not responding to POLL (MAGENTA).

B. Only 2 of 4 Circulating Water Pumps running on associated Unit.

C. PC-11 channel in ALERT alarm (YELLOW).

D. Loss of counts on X-RE-5253, Liquid Effluent Radiation Monitor.

Proposed Answer: D Explanation:

A. Incorrect. Plausible because it could be thought that a monitor not responding to POLL would be INOPERABLE.

B. Incorrect. Plausible because Circulating Water Pumps must be running for the valve to remain open, however, a 2 of 4 coincidence allows release to Unit aligned for discharge.

C. Incorrect. Plausible because radiation level has increased, however, it requires a high radiation level alarm to close X-RV-5253.

D. Correct. A loss of counts on the Liquid Effluent Radiation Monitor will trip X-RV-5253 (OPERATE FAILURE).

Technical Reference(s) ALM-3200, Page 38 Attached w/ Revision # See ALM-3200, Attachment 3 Comments / Reference Proposed references to be provided during examination: None Learning Objective: STATE the functions, operation and interlocks of the Liquid Waste Processing System components.

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

New Question History: Last NRC Exam CPNPP 2011 Page 43 of 92 CPNPP NRC 2013 RO Written Exam Worksheet 26 to 50 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 11 55.43 Comments /

Reference:

From ALM-3200, Page 38 Revision # 4 Page 44 of 92 CPNPP NRC 2013 RO Written Exam Worksheet 26 to 50 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ALM-3200, Attachment 3 Revision # 4 Page 45 of 92 CPNPP NRC 2013 RO Written Exam Worksheet 26 to 50 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 007 EK2.03 Importance Rating 3.5 Reactor Trip - Stabilization - Recovery: Knowledge of the interrelations between a reactor trip and the following: Reactor trip status panel Proposed Question: Common 39 Given the following conditions:

• Unit 1 is at 75% power.

• Solid State Protection System (SSPS) Train B Actuation Logic testing is being performed.

• Train B SSPS Mode Selector switch is in the TEST position.

• Train B SSPS Input Error Inhibit switch is in the INHIBIT position.

Which of the following describes the status of the Reactor if a loss of one of the two 48 VDC instrument power supply were to occur on Train A SSPS?

A. Reactor at 75% power with a General Warning for Train A SSPS ONLY.

B. Reactor at 75% power with a General Warning for Train B SSPS ONLY.

C. Reactor Trip with a General Warning for BOTH Train A and Train B SSPS and a First Out Alarm illuminated.

D. Reactor Trip with a General Warning for BOTH Train A and Train B SSPS and NO First Out Alarm illuminated.

Proposed Answer: C Page 46 of 92 CPNPP NRC 2013 RO Written Exam Worksheet 26 to 50 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Explanation: For this question we assumed the Reactor Trip status panel is equivalent to our First Out Panel.

A. Incorrect. Plausible because a General Warning is generated for a loss of either 48 VDC power supply. If this were the only General Warning the Unit would remain at power, but performing testing on the other train generates a General Warning for both trains and the Unit trips.

B. Incorrect. Plausible because a General Warning is generated while performing testing on SSPS. If this were the only General Warning the Unit would remain at power, but a loss of either 48 VDC power supply on the other train generates a General Warning for both trains and the Unit trips.

C. Correct. Testing on one train of SSPS generates a General Warning. A loss of any of the four DC power supplies in the other train of SSPS also generates a General Warning. General Warnings in both trains of SSPS causes the Reactor Trip Breakers to open, which then causes the Turbine to trip. Since the power level is above 50%, the Turbine trip then causes a Reactor trip signal to be generated which causes the First Out annunciator. The First Out annunciator would NOT alarm if power were below 50%.

D. Incorrect. Plausible because a Reactor Trip is generated, but a First Out annunciator occurs due to the Unit being above P-9 (50%) power for RX > 50% PWR TURB TRIP.

Technical Reference(s) ALM-0064A, 1-ALB-6D, Window 1.5 Attached w/ Revision # See Comments / Reference 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 # X (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, 7 55.43 Page 47 of 92 CPNPP NRC 2013 RO Written Exam Worksheet 26 to 50 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ALM-0064A, 1-ALB-6D, Window 1.5 Revision # 6 Page 48 of 92 CPNPP NRC 2013 RO Written Exam Worksheet 26 to 50 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From CPNPP Exam Bank Revision # 03/11/09 Given the following conditions:

• Unit 1 is at 40% power.

• Solid State Protection System (SSPS) Train B Actuation Logic testing is being performed.

• Train B SSPS Mode Selector Switch is in the TEST position.

• Train B SSPS Input Error Inhibit Switch is in the INHIBIT position.

Which of the following describes the status of the Reactor if a loss of Distribution Panel 1PC1 were to occur on Train A SSPS?

A. Reactor at 40% power with a GENERAL WARNING for Train A SSPS only.

B. Reactor Trip with a GENERAL WARNING for both Train A and Train B SSPS with the First Out annunciator NOT illuminated.

C. Reactor at 40% power with a GENERAL WARNING for Train B SSPS only.

D. Reactor Trip with a GENERAL WARNING for both Train A and Train B SSPS and the First Out annunciator illuminated.

Page 49 of 92 CPNPP NRC 2013 RO Written Exam Worksheet 26 to 50 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 008 AK2.03 Importance Rating 2.5 Pressurizer Vapor Space Accident: Knowledge of the interrelations between Pressurizer Vapor Space Accident and the following: Controllers and positioners Proposed Question: Common 40 Given the following conditions:

• A Loss of Coolant Accident (LOCA) is in progress on Unit 1.

• 1-PI-455A, PRZR PRES CHAN I, is indicating 2500 psig.

• The other three Pressurizer Pressure Channels are indicating 2100 psig and lowering.

• 1EA1, Train A 6.9KV Safeguards Bus is de-energized.

• Containment pressure is 4 psig and rising.

• All Reactor Coolant Pumps are secured.

• Pressurizer valve status is as follows;

• Pressurizer Safety Valves are CLOSED.

• 1-PCV-455A, PRZR PORV, is OPEN.

• 1-PCV-456, PRZR PORV, is CLOSED.

• Pressurizer level is 100%.

• Reactor Vessel Level Indicating System (RVLIS) lights at 33, 21 and 11 are LIT and the remaining RVLIS lights are DARK.

Which of the following control channels should have initiated an automatic action to terminate the LOCA?

A. 1-PT-0455, PRZR PRESS XMTR 0455 PROT CHAN I B. 1-PT-0456, PRZR PRESS XMTR 0456 PROT CHAN II C. 1-PT-0457, PRZR PRESS XMTR 0457 PROT CHAN III D. 1-PT-0458, PRZR PRESS XMTR 0458 PROT CHAN IV Proposed Answer: D Page 50 of 92 CPNPP NRC 2013 RO Written Exam Worksheet 26 to 50 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible as the indications are consistent with those which would exists if a high failure of PT-455 were to occur with no operator action per ABN-705.

B. Incorrect. Plausible if thought that the indications are consistent with those which would exist if a failure of PCV-456 were to occur with no operator action per ABN-705.

C. Incorrect. Plausible because this answer would be correct if PORV-456 were failed in the mid position.

D. Correct. When pressure drops below 2185 psig the interlock associated with PT-0458 should have closed PORV-0455A.

Technical Reference(s) ALM-0052A, 1-ALB-5B, Windows 1.6 & 2.6 Attached w/ Revision # See ABN-705, Section 2.2 Comments / Reference Proposed references to be provided during examination: None Learning Objective: During abnormal or emergency events, ANALYZE indications to determine the cause of the abnormal or emergency event.

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Page 51 of 92 CPNPP NRC 2013 RO Written Exam Worksheet 26 to 50 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ALM-0052A, 1-ALB-5B, Window 2.6 Revision # 5 Page 52 of 92 CPNPP NRC 2013 RO Written Exam Worksheet 26 to 50 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ALM-0052A, 1-ALB-5B, Window 1.6 Revision # 5 Page 53 of 92 CPNPP NRC 2013 RO Written Exam Worksheet 26 to 50 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ABN-705, Section 2.2 Revision # 12 Page 54 of 92 CPNPP NRC 2013 RO Written Exam Worksheet 26 to 50 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 009 EA2.37 Importance Rating 4.2 Small Break LOCA: Ability to determine and interpret the following as they apply to the Small Break LOCA: Existence of adequate natural circulation Proposed Question: Common 41 Given the following conditions:

• A Small Break Loss of Coolant Accident has occurred on Unit 1.

• Conditions to start a Reactor Coolant Pump cannot be established.

• Containment pressure is 8 psig and slowly rising.

• Reactor Coolant System (RCS) pressure is 1085 psig and stable.

• All Steam Generator narrow range levels are approximately 55% and stable.

• Steam Generator pressures are as follows:

• 1-01 is 785 psig and stable.

• 1-02 is 790 psig and stable.

• 1-03 is 850 psig and stable.

• 1-04 is 780 psig and stable.

• RCS Cold Leg temperatures are as follows:

• Loop 1 is 518°F and stable.

• Loop 2 is 519°F and stable.

• Loop 3 is 360°F and lowering.

• Loop 4 is 517°F and stable.

• All RCS Hot Leg temperatures are approximately 540°F and stable.

• Core Exit Thermocouples are reading approximately 550°F and slowly rising.

What is the status of natural circulation and the expected operator action?

A. Adequate natural circulation does NOT exist and the Steam Dump Valves should be opened farther.

B. Adequate natural circulation does exist and the Atmospheric Relief Valve 1-03 should be closed farther.

C. Adequate natural circulation does NOT exist and the Atmospheric Relief Valves should be opened farther.

D. Adequate natural circulation does exist and Steam Dump Valves should be closed farther.

Page 55 of 92 CPNPP NRC 2013 RO Written Exam Worksheet 26 to 50 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Proposed Answer: C Explanation:

A. Incorrect. Plausible as Natural Circulation does NOT exist, however, the MSIVs would have closed at 6.2 psig Containment pressure and thus the Steam Dump Valves are isolated.

B. Incorrect. Plausible as three of the four Steam Generators appear to be coupled. However, adequate subcooling does not exist and one Steam Generator is uncoupled. Additionally, the CETs are increasing which indicates that adequate Natural Circulation does NOT exist. Steam Generator 3 appears to be overcooling, thus the concept to decrease steam dumping from this generator is plausible, however, the low temperature indicates lack of circulation and Cold Leg temperature lowering as a result of ECCS flow.

C. Correct. Since adequate subcooling does NOT exist, CETs are increasing and SG 1-03 is uncoupled, which indicate that Natural Circulation does NOT exist. EOS-1.2A, states to increase dumping steam to promote Natural Circulation. The Steam Dump Valves are isolated by the MSIVs which only leave the Atmospheric Relief Valves for dumping steam.

D. Incorrect. Plausible as three of the four Steam Generators appear to be coupled. However, adequate subcooling does not exist and one Steam Generator is uncoupled. Additionally, the CETs are increasing which indicates that adequate Natural Circulation does NOT exist. Steam Generator 3 appears to be overcooling, thus the concept to decrease steam dumping from this generator is plausible, however, the low temperature indicates lack of circulation and Cold Leg temperature lowering as a result of ECCS flow.

Technical Reference(s) EOS-1.2A, Attachment 3 & Step 22.a RNO Attached w/ Revision # See Comments / Reference Proposed references to be provided during examination: None Learning Objective: ANALYZE the stagnant Reactor Coolant System Loops generic issue in the Emergency Response Guideline network and proper operator 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 5, 14 55.43 Page 56 of 92 CPNPP NRC 2013 RO Written Exam Worksheet 26 to 50 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From EOS-1.2A, Attachment 3 Revision # 8 Comments /

Reference:

From EOS-1.2A, Step 22.a RNO Revision # 8 Page 57 of 92 CPNPP NRC 2013 RO Written Exam Worksheet 26 to 50 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 015/17 AK1.04 Importance Rating 2.9 RCP Malfunctions: Knowledge of the operational implications of the following concepts as they apply to the RCP Malfunctions: Basic steady-state thermodynamic relationship between RCS loops and SGs resulting from unbalanced loop flows Proposed Question: Common 42 Given the following conditions:

• Unit 2 is at 35% power.

• Reactor Coolant Pump (RCP) 2-02 trips.

In the 30 seconds following the trip of RCP 2-02, and assuming NO operator action, an automatic Reactor Trip will ...

A. ...occur, and the affected SG water level will shrink.

B. ...NOT occur, but the affected SG water level will shrink.

C. ...occur, and the affected SG water level will swell.

D. ...NOT occur, but the affected SG water level will swell.

Proposed Answer: B Explanation:

A. Incorrect. Plausible because SG level will shrink due to loss of heat input to SG 2-02; however, the Reactor will not automatically trip unless power level is greater than 48%.

B. Correct. A Reactor Trip will occur when one Reactor Coolant Pump trips with Reactor power greater than 48%. When the RCP trips, SG shrink due to loss of heat input to SG 2-02.

C. Incorrect. Plausible if thought that the P-8 permissive had been met, however, the Reactor does not trip and Steam Generator water level will initially shrink due to loss of heat input to SG 2-02.

D. Incorrect. Plausible because the Reactor will not trip, however, SG level will shrink due to loss of heat input to SG 2-02.

Technical Reference(s) ABN-101, Section 2.2 Attached w/ Revision # See ABN-101, Step 2.3.1 NOTE Comments / Reference Proposed references to be provided during examination: None Page 58 of 92 CPNPP NRC 2013 RO Written Exam Worksheet 26 to 50 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Learning Objective: COMPREHEND the normal, abnormal and emergency operation of the Reactor Coolant System.

ANALYZE the response to an RCP Trip per ABN-101, Reactor Coolant Pump Trip/Malfunction.

Question Source: Bank # X 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, 14 55.43 Page 59 of 92 CPNPP NRC 2013 RO Written Exam Worksheet 26 to 50 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ABN-101, Section 2.2 Revision # 10 Page 60 of 92 CPNPP NRC 2013 RO Written Exam Worksheet 26 to 50 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ABN-101, Step 2.3.1 NOTE Revision # 10 Page 61 of 92 CPNPP NRC 2013 RO Written Exam Worksheet 26 to 50 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 025 AK3.03 Importance Rating 3.9 Loss of RHR System: Knowledge of the reasons for the following responses as they apply to the Loss of Residual Heat Removal System: Immediate actions contained in EOP for Loss of RHRS Proposed Question: Common 43 Given the following conditions:

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

• A transition has been made to EOS-1.3A, Transfer to Cold Leg Recirculation.

• Emergency Core Cooling System (ECCS) has been transferred to Cold Leg Recirculation and the RWST level is 8% and lowering.

• The following ECCS pumps are running:

• Centrifugal Charging Pumps (CCP) 1-01 and 1-02.

• Safety Injection Pumps (SIP) 1-01 and 1-02.

• Residual Heat Removal Pumps (RHRP) 1-01 and 1-02.

• RHRP 1-01 is cavitating.

Which of the following lists the required action and reason for stopping the ECCS pumps in accordance with EOS-1.3A, Transfer to Cold Leg Recirculation?

Stop A. SIP 1-01 to improve NPSH for RHRP 1-01.

B. CCP 1-01 to improve NPSH for RHRP 1-01.

C. RHRP 1-01 to prevent pump damage.

D. CCP 1-01 to prevent pump damage.

Proposed Answer: C Page 62 of 92 CPNPP NRC 2013 RO Written Exam Worksheet 26 to 50 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible as securing SIP 1-01 would cause the greatest improvement to RHRP 1-01 NPSH, however RHRP 1-01 should be secured due to cavitation.

B. Incorrect. Plausible as securing CCP 1-01 would cause some improvement to RHRP 1-01 NPSH, however RHRP 1-01 should be secured due to cavitation.

C. Correct. Because the CAUTION in EOS-1.3A states to stop any ECCS Pump which loses suction or shows indication of cavitation should be stopped to prevent pump damage and the suction supply to the SIPs and CCPs is cross-connected so that either RHRP will provide adequate suction. The 8804 valves in the RHR system and the 8807 valves in the SI system provide this dual suction source alignment.

D. Incorrect. Plausible since CCP 1-01 is most susceptible to damage due to cavitation and if RHRP 1-01 were the only suction source to CCP 1-01 this action would be appropriate. However, there is no indication of CCP 1-01 cavitation.

Technical Reference(s) EOS-1.3A, Step 3 CAUTION Bases Attached w/ Revision # See LO21.SYS.RH1 Comments / Reference LO21.SYS.SI1 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).

Given a set of plant conditions, IDENTIFY the proper transitions through/out of EOS-1.3, 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 63 of 92 CPNPP NRC 2013 RO Written Exam Worksheet 26 to 50 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From EOS-1.3A, Step 3 CAUTION Bases Revision 8 Page 64 of 92 CPNPP NRC 2013 RO Written Exam Worksheet 26 to 50 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From LO21.SYS.RH1 Lesson Notes Revision 10/21/11 RHR PUMP TO CENTRIFUGAL CHARGING PUMP / SAFETY INJECTION PUMP SUCTION ISOLATION VALVES (U-8804A&B)

The RHR Pump to Centrifugal Charging Pump / Safety Injection Pump Suction Isolation Valves function to align one of the discharge paths of the RHR Pumps to the suction piping of the Centrifugal Charging Pumps and Safety Injection Pumps. u-8804A provides flow to the suction of the Centrifugal Charging Pump and u-8804B provides flow to the suction of the Safety Injection Pumps. This path is used when emergency operating procedures require the Emergency Core Cooling System to be placed in cold leg or hot leg recirculation mode. u-8804A&B are motor-operated valves controlled from CB-04.

Comments /

Reference:

From LO21.SYS.SI1 Lesson Notes Revision 5/2/11 Safety Injection Pump/Centrifugal Charging Pump Suction Header Cross-Tie Isolation Valves (u-8807A &

B)

The Safety Injection Pump/Centrifugal Charging Pump Suction Header Cross-Tie Valves allow either of the Residual Heat Removal Pumps to deliver flow to the suction of the Centrifugal Charging and Safety Injection Pumps during the long-term recirculation mode. The valves are arranged in parallel to assure that at least one valve can be opened to provide a flowpath between the pump suction headers. u-8807A & B are normally left in the closed position, and are opened by manual operation from the Main Control Board when transferring the Emergency Core Cooling System to Cold Leg Recirculation. The valves are located in the Train A ECCS Valve Room on the 790' level of the SFGD s Bldg.

Page 65 of 92 CPNPP NRC 2013 RO Written Exam Worksheet 26 to 50 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 027 AK2.03 Importance Rating 2.6 Pressurizer Pressure Control Malfunction: Knowledge of the interrelations between the Pressurizer Pressure Control Malfunctions and the following: Controllers and positioners.

Proposed Question: Common 44 Given the following conditions:

• Unit 2 is in MODE 1

• All Pressurizer Backup Heaters have failed ON and cannot be secured.

What is the plant response?

A. Both Pressurizer spray valves open to restore pressure between 2220 psig and 2250 psig.

B. Only one Pressurizer spray valve opens to restore pressure between 2220 psig and 2250 psig.

C. Both Pressurizer Power Operated Relief Valves open to maintain pressure below 2335 psig.

D. Only one Pressurizer Power Operated Relief Valve opens to maintain pressure below 2335 psig.

Proposed Answer: A Explanation:

A. Correct. With all backup heaters energized, both pressurizer spray valves will open. Pressurizer pressure will initially peak at approximately 2270 psig and then lower to the normal control band.

B. Incorrect. Plausible if it is believed that a single pressurizer spray valve is sufficient to overcome the energy added by all backup heaters.

C. Incorrect. Plausible if it believed that both pressurizer spray valves are not sufficient to mitigate the energy added by all backup heaters.

D. Incorrect. Plausible if it believed that both pressurizer spray valves are not sufficient to mitigate the energy added by all backup heaters and both PORVs open at the same setpoint.

Technical Reference(s) LO21.SYS.PP1 Attached w/ Revision # See DBD-ME-250 Reactor Coolant System Comments / Reference ABN-705 Page 66 of 92 CPNPP NRC 2013 RO Written Exam Worksheet 26 to 50 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Proposed references to be provided during examination: None Learning Objective: Analyze the response to a 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 Comments /

Reference:

From LO21.SYS.PP1 Revision 5/5/11 During normal plant operations, the pressurizer is filled with boiling water and steam. The temperature of the boiling (or saturated) water determines the pressure inside the entire Reactor Coolant System. Pressurizer temperature is controlled to regulate RCS pressure by energizing electric heaters in the bottom of the pressurizer to raise pressure, and by spraying the steam space (or steam bubble) in the top of the pressurizer with cooler water to reduce pressure.

Under normal operating conditions, the Pressurizer Pressure Control System will automatically maintain the plant at 2235 psig. Heaters maintain a saturated condition in the pressurizer and spray valves throttle open to hold pressure at the 2235 psig setpoint. Backup banks of heaters energize on decreasing RCS pressure. On increasing pressure, spray valves open automatically to cause partial steam bubble condensation.

Comments /

Reference:

From DBD-ME-250 Reactor Coolant System Revision 45 The pressurizer spray valves are required to pass the maximum cold leg spray flow sufficient to maintain the system pressure below the power operated relief valve setpoint for a 12 percent load rejection at 100 percent power.

The pressurizer spray line is required to deliver 900 gpm to the pressurizer.

RCS pressure is controlled by the use of the pressurizer where water and steam are maintained in equilibrium by electrical heaters and water sprays. Steam can be formed (by the heaters) or condensed (by the pressurizer spray) to reduce pressure variations due to contraction and expansion of the reactor coolant.

Page 67 of 92 CPNPP NRC 2013 RO Written Exam Worksheet 26 to 50 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ABN-705, Section 3.0 Revision 12 This reference is included to indicate that spray flow will overcome heater input if a spray valve were to fail open.

So that indicates that if heaters were to all come on that spray flow would overcome heater input and restore pressure to the control band.

Page 68 of 92 CPNPP NRC 2013 RO Written Exam Worksheet 26 to 50 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 038 G 2.4.8 Importance Rating 3.8 Steam Generator Tube Rupture: Emergency Procedures/Plan: Knowledge of how abnormal operating procedures are used in conjunction with EOPs Proposed Question: Common 45 Given the following conditions:

• EOP-3.0A, Steam Generator Tube Rupture (SGTR), is in progress.

• A Loss of Offsite Power (LOOP) occurs.

Assuming a full crew compliment is available which of the following is the anticipated response strategy?

A. Continue in EOP-3.0A, while the LOOP is addressed via the Abnormal Conditions Procedures (ABNs).

B. Suspend actions of EOP-3.0A and immediately enter ECA-0.0A, Loss of All AC Power.

C. Continue in EOP-3.0A and immediately enter ECA-0.0A, Loss of All AC Power.

D. Suspend actions of EOP-3.0A, while the LOOP is addressed via the Abnormal Conditions Procedures (ABNs).

Proposed Answer: A Explanation:

A. Correct. Crew should continue to address the tube rupture in EOP-3.0A. The crew should address the LOOP using Abnormal Conditions Procedures (ABNs) in accordance with step 33 of EOP-3.0A.

B. Incorrect. Plausible if it is believed that a LOOP is a Loss of All AC Power. However, the Emergency Diesel Generators will power the safeguards buses and the LOOP will be addressed by ABN-601 in accordance with step 33 of EOP-3.0A.

C. Incorrect. Plausible if it is believed that a LOOP is a Loss of All AC Power and that the actions of EOPs are performed concurrently with ECA-0.0A. However, the Emergency Diesel Generators will power the safeguards buses and the LOOP will be addressed by ABN-601 in accordance with step 33 of EOP-3.0A and on an actual loss of all AC, actions in accordance with EOP-3.0A would be suspended while the crew performed ECA-0.0A D. Incorrect. Plausible if it is believed that actions to restore offsite power must be performed prior to completing actions of EOP-3.0A. However, the LOOP will be addressed by ABN-601 in accordance with step 33 of EOP-3.0A.

Page 69 of 92 CPNPP NRC 2013 RO Written Exam Worksheet 26 to 50 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Technical Reference(s) EOP-3.0A, Step 33 Attached w/ Revision # See ODA-407, Attachment 8.A Comments / Reference Proposed references to be provided during examination: None Learning Objective: ANALYZE the diagnostic steps of EOP-3.0, 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 10 55.43 Page 70 of 92 CPNPP NRC 2013 RO Written Exam Worksheet 26 to 50 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From EOP-3.0A, Step 33 Revision 8 Page 71 of 92 CPNPP NRC 2013 RO Written Exam Worksheet 26 to 50 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ODA-407, Attachment 8.A, page 3 Revision 15 Page 72 of 92 CPNPP NRC 2013 RO Written Exam Worksheet 26 to 50 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ODA-407, Attachment 8.A, page 4 Revision 15 Page 73 of 92 CPNPP NRC 2013 RO Written Exam Worksheet 26 to 50 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 054 AK1.01 Importance Rating 4.1 Loss of Main Feedwater: Knowledge of the operational implications of the following concepts as they apply to the Loss of Main Feedwater: MFW line break depressurizes the S/G (similar to a steam line break)

Proposed Question: Common 46 Given the following conditions:

• Unit 1 has received an automatic Reactor Trip and Safety Injection.

• Containment pressure is 4 psig and rising.

• Containment Sump level and pump run alarms are locked in.

• Steam Generator 1-01 level is lowering.

• Steam Generators 1-02, 1-03 and 1-04 levels are rising.

• All Steam Generator pressures are 1080 psig and stable.

• 1RE-5503, CNTMT AIR GAS (CAG-197) is GREEN and stable.

Which of the following events initiated the Reactor Trip and Safety Injection?

A. Main Steam Line break on Steam Generator 1-01 B. Main Feed Line break on Steam Generator 1-01.

C. Small Break Loss of Coolant Accident on Reactor Coolant System Loop 1 Cold Leg.

D. Small Break Loss of Coolant Accident on Reactor Coolant System Loop 1 Hot Leg.

Proposed Answer: B Explanation:

A. Incorrect. Plausible because Containment pressure and dew point are both rising, however, Steam Generator pressures are stable which would differentiate a steam line from a feed line break.

B. Correct. Given the conditions listed, this is the correct diagnosis.

C. Incorrect. Plausible because Containment Sump level and pressure are both rising and Steam Generator pressures are stable, however, stable containment air gaseous activity below the alarm setpoint is not consistent with a primary coolant leak in containment.

D. Incorrect. Plausible because Containment Sump level and pressure are both rising and Steam Generator pressures are stable, however, stable containment air gaseous activity below the alarm setpoint is not consistent with a primary coolant leak in containment.

Technical Reference(s) WOG Background Document for EOP-2.0 Attached w/ Revision # See Comments / Reference Page 74 of 92 CPNPP NRC 2013 RO Written Exam Worksheet 26 to 50 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 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 3, 5 55.43 Page 75 of 92 CPNPP NRC 2013 RO Written Exam Worksheet 26 to 50 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From WOG Background Document for EOP-2.0 Revision 2 Page 76 of 92 CPNPP NRC 2013 RO Written Exam Worksheet 26 to 50 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 055 EA2.03 Importance Rating 3.9 Station Blackout: Ability to determine and interpret the following as they apply to the Station Blackout: Actions necessary to restore power Proposed Question: Common 47 A system wide grid blackout has occurred. Which of the following actions is the preferred method to restore offsite power?

Align the...

A. ...138 KV Switchyard to be powered from Stephenville.

B. ...345 KV Switchyard east bus to be powered from DeCordova.

C. ...345 KV Switchyard west bus to be powered from Wolf Hollow.

D. ...138 KV Switchyard to be powered from DeCordova.

Proposed Answer: D Explanation:

A. Incorrect. Plausible because the second option for black start is the 138 KV Stephenville corridor.

B. Incorrect. Plausible because if the 138 KV switchyard cannot be aligned, then the DeCordova 345 KV line is the next available option.

C. Incorrect. Plausible because the Wolf Hollow plant is the nearest in proximity to CPNPP and has been the black start unit in the past.

D. Correct. Aligning the 138 KV DeCordova line is the preferred Black Start corridor to restore power to Comanche Peak.

Technical Reference(s) ABN-601, Attachment 20 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 Switchyard System.

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Page 77 of 92 CPNPP NRC 2013 RO Written Exam Worksheet 26 to 50 Rev h

ES-401 CPNPP NRC 2013 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:

From ABN-601, Attachment 20 Revision # 11 Page 78 of 92 CPNPP NRC 2013 RO Written Exam Worksheet 26 to 50 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ABN-601, Attachment 20 Revision # 11 Page 79 of 92 CPNPP NRC 2013 RO Written Exam Worksheet 26 to 50 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 056 AA1.05 Importance Rating 3.8 Loss of Offsite Power: Ability to operate and/or monitor the following as they apply to a Loss of Offsite Power: Initiation (manual) of safety injection process Proposed Question: Common 48 Given the following conditions:

• A Loss of Offsite Power has occurred.

• Bus 1EA1 is de-energized.

• EOS-0.1A, Reactor Trip Response, is in progress.

• Reactor Coolant System (RCS) temperature is 561ºF and stable.

• RCS pressure is 1920 psig and trending down slowly.

• Pressurizer level is 5% and slowly lowering.

Which of the following actions is required per EOS-0.1A, Reactor Trip Response?

A. Increase Condenser Steam Dump to maintain T AVE at 557ºF per EOS-0.1A, Reactor Trip Response.

B. Attempt to restore Bus 1EA1 per ABN-602, Response to a 6900V/480V System Malfunction.

C. Manually actuate Safety Injection and return to EOP-0.0A, Reactor Trip or Safety Injection.

D. Isolate Letdown and verify Natural Circulation per EOS-0.1A, Reactor Trip Response.

Proposed Answer: C Explanation:

A. Incorrect. Plausible because this is a Step 1 RNO action of EOS-0.1A, however, Steam Dump will not be available without Circulating Water Pumps. T AVE is where it should be for the conditions.

B. Incorrect. Plausible because it would have been performed in EOP-0.0A, however, priority is Safety Injection (SI).

C. Correct. EOS-0.1A Foldout Page requires manual initiation of SI when PRZR level cannot be maintained greater than 6% and a transition back to EOP-0.0A.

D. Incorrect. Plausible because Natural Circulation would be verified if SI was not required per the Foldout Page. Additionally, letdown is isolated if Pressurizer Level is less than 17%.

Page 80 of 92 CPNPP NRC 2013 RO Written Exam Worksheet 26 to 50 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Technical Reference(s) EOS-0.1A, Attachments 1.A & 3 Attached w/ Revision # See EOS-0.1A, Steps 1 RNO & 6 RNO Comments / Reference Proposed references to be provided during examination: None Learning Objective: ANALYZE the recovery technique used and the procedure steps of EOS-0.1, Reactor Trip Response.

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

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

Reference:

From EOS-0.1A, Attachment 1.A Revision # 8 Page 81 of 92 CPNPP NRC 2013 RO Written Exam Worksheet 26 to 50 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From EOS-0.1A, Attachment 3 Revision # 8 Page 82 of 92 CPNPP NRC 2013 RO Written Exam Worksheet 26 to 50 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From EOS-0.1A, Step 1 RNO Revision # 8 Page 83 of 92 CPNPP NRC 2013 RO Written Exam Worksheet 26 to 50 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From EOS-0.1A, Step 1 RNO Revision # 8 Page 84 of 92 CPNPP NRC 2013 RO Written Exam Worksheet 26 to 50 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 057 AA2.15 Importance Rating 3.8 Loss of Vital AC Instrument Bus: Ability to determine and interpret the following as they apply to a Loss of Vital AC Instrument Bus: That a loss of ac has occurred.

Proposed Question: Common 49 Given the following conditions:

• Unit 1 is at 100% power.

• 1-ALB-10B, Window 4.16 - 118V CHAN IV INV TRBL is lit.

• Multiple indications on the Main Control Board are not indicating actual parameter values.

• Multiple Main Control Board annunciators are LIT.

Which of the following is an alternate method for determining which 118V Protection Bus is de-energized?

A. Row four lights on the Trip Status Light Boxes are LIT.

B. Row four lights on the PCIP are LIT.

C. Row four lights on the Trip Status Light Boxes are DARK.

D. Row four lights on the PCIP are DARK.

Proposed Answer: C Explanation:

A. Incorrect. Plausible because a loss of Channel 1, 2 or 3 results in the associated Trip Status Light Boxes lighting for all trips that are de-energize to actuate.

B. Incorrect. Plausible because the PCIP provides important information on Permissives and Control logics. It is plausible that an operator could attempt to assess the plant status from the PCIP Windows as a change of status does occur on the panel, however, the operator cannot determine that a loss of Channel 4 has occurred from this panel alone.

C. Correct. As identified in the NOTE box of ABN-603, Section 2.0.

D. Incorrect. Plausible because the PCIP provides important information on Permissives and Control logics. It is plausible that an operator could attempt to assess the plant status from the PCIP Windows as a change of status does occur on the panel, however, the operator cannot determine that a loss of Channel 4 has occurred from this panel alone.

Technical Reference(s) ABN-603, Section 2.0 Attached w/ Revision # See Comments / Reference Page 85 of 92 CPNPP NRC 2013 RO Written Exam Worksheet 26 to 50 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Proposed references to be provided during examination: None Learning Objective: ANALYZE the response to Loss of a Protection Bus per ABN-603, Loss of Protection or 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 X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 5 Page 86 of 92 CPNPP NRC 2013 RO Written Exam Worksheet 26 to 50 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ABN-603, Section 2.0 Revision # 8 Page 87 of 92 CPNPP NRC 2013 RO Written Exam Worksheet 26 to 50 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 058 AK3.02 Importance Rating 4.0 Loss of DC Power: Knowledge of the reasons for the following responses as they apply to the Loss of DC Power: Actions contained in EOP for loss of DC power Proposed Question: Common 50 Given the following conditions on Unit 1:

• Actions of ECA-0.0A, Loss of All AC Power, are being performed.

• Safeguards DC Bus voltage is 118 VDC and slowly lowering.

• Large non-essential loads have been shed.

When is additional load shedding required to be performed and what is the reason?

When Plant Staff has determined additional load shedding is required and DC Bus Voltage is A. less than 110 VDC. This will ensure that all control room indications remain available.

B. less than 110 VDC. This will maintain the ability to flash the Emergency Diesel Generator field.

C. less than 115 VDC. This will ensure that all control room indications remain available.

D. less than 115 VDC. This will maintain the ability to flash the Emergency Diesel Generator field.

Proposed Answer: B Explanation:

A. Incorrect. Plausible because additional loads are shed at 110 VDC, however, all indication is NOT maintained (channels 3 and 4 are load shed).

B. Correct. Additional loads are shed at 110 VDC to maintain the ability to field flash the EDG.

C. Incorrect. Plausible if believed that additional load shedding is completed to maintain all channels of control room indication at 115 VDC. However, additional load shedding is conducted at 110 VDC to maintain the ability to field flash the EDG.

D. Incorrect. Plausible because additional load shedding is conducted to maintain the ability to field flash the EDG. However, additional load shedding is conducted at 110 VDC.

Page 88 of 92 CPNPP NRC 2013 RO Written Exam Worksheet 26 to 50 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Technical Reference(s) ECA-0.0A, Attachment 7, Step 16 Bases Attached w/ Revision # See ECA-0.0A, Step 16 Comments / Reference ECA-0.0A, Attachment 2 Bases 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.

Question Source: Bank # X 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 89 of 92 CPNPP NRC 2013 RO Written Exam Worksheet 26 to 50 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ECA-0.0A, Attachment 7, Step 16 Bases Revision # 8 Page 90 of 92 CPNPP NRC 2013 RO Written Exam Worksheet 26 to 50 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ECA-0.0A, Step 16 Revision # 8 Page 91 of 92 CPNPP NRC 2013 RO Written Exam Worksheet 26 to 50 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ECA-0.0A, Attachment 2, Bases Revision 8 Page 92 of 92 CPNPP NRC 2013 RO Written Exam Worksheet 26 to 50 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 062 AA2.03 Importance Rating 2.6 Loss of Nuclear Service Water: Ability to determine and interpret the following as they apply to the Loss of Nuclear Service Water: The valve lineups necessary to restart the SWS while bypassing the portion of the system causing the abnormal condition Proposed Question: Common 51 Given the following conditions:

• Unit 1 is at 100% power.

• A body to bonnet leak has developed on the Station Service Water (SSW) system Emergency Diesel Generator 1-02 Jacket Water Cooler inlet isolation valve.

• The crew is determining how to isolate the leak that will impact the FEWEST components.

Which of the following describes what should be isolated and which of the following components will be affected by isolating the Station Service Water system leak?

1. 1-02 Centrifugal Charging Pump lube oil cooler

2. X-02 Station Service Water Screen Wash Pump

3. 1-02 Component Cooling Water Heat Exchanger

4. 1-02 Containment Spray Pump bearing cooler A. Train B Station Service Water should be isolated 1 and 4 B. Train B Station Service Water 10 header should be isolated 2 and 3 C. Train B Station Service Water should be isolated 2 and 3 D. Train B Station Service Water 10 header should be isolated 1 and 4 Proposed Answer: D Page 1 of 60 CPNPP NRC 2013 RO Written Exam Worksheet 51 to 64 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible because it could be thought that the entire train must be isolate, however only the 10 header should be isolated and the CCP and CTP are the correct components if the 10 header were isolated.

B. Incorrect. Plausible because the 10 header should be isolated however the CCW heat exchanger and screen wash pump are not supplied from the 10 header.

C. Incorrect. Plausible because it could be thought that the entire train must be isolate, however only the 10 header should be isolated and isolating the entire train would affect the CCW heat exchanger and screen was pump.

D. Correct. The CCP and CTP are loads on the 10 inch SSW header that would be affected by isolating the 10 header.

Technical Reference(s) SOP-501A, Step 5.5.2 CAUTION Attached w/ Revision # See ABN-501, Attachment 3 Comments / Reference LO21.SYS.SW1 Proposed references to be provided during examination: None Learning Objective: ANALYZE the response to a Loss Station Service Water per 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 4, 10 55.43 Page 2 of 60 CPNPP NRC 2013 RO Written Exam Worksheet 51 to 64 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ABN-501, Attachment 3 Revision # 9 Page 3 of 60 CPNPP NRC 2013 RO Written Exam Worksheet 51 to 64 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SOP-501A, Step 5.5.2 CAUTION Revision # 19 Page 4 of 60 CPNPP NRC 2013 RO Written Exam Worksheet 51 to 64 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From LO21.SYS.SW1.LN, Page 15 Revision # 04/05/05 1.1 A separate 10-inch line (Figure 2) branches from each of the main 30-inch SSW supply headers and runs through the Auxiliary and Safeguards Building. In the Auxiliary Building, the 10-inch branch supplies cooling water to the centrifugal charging pumps lube oil coolers and in the Safeguards Building it supplies cooling water to the safety injection pumps lube oil coolers, the containment spray pump bearing oil coolers and the diesel generator system. A backup water supply to the auxiliary feedwater pumps is provided from the SSW discharge piping downstream of the diesel generator cooler. The return flow joins the main SSW discharge line on the downstream side of the CCW heat exchanger. Heated service water is piped underground to the service water discharge canal which carries it back to the SSI.

Figure 2 - SSW Safety Loop (One Train) 1.2 The SSW discharge canal is common to both units. If the canal is blocked, the water will spill into the yard and drain to the SSI without impairing operation of the SSW system.

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ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From LO21.SYS.SW1.LN, Page 14 Revision # 04/05/05 FLOWPATHS 1.3 The SSW pumps take suction from the Safe Shutdown Impoundment (SSI). SSI water enters the SSW system through trash racks then traveling screens to the intake bay of the Service Water Intake Structure (SWIS) (Figure 1).

Figure 1 - Station Service Water System 1.4 The SSW pumps discharge to the CCW HX, Diesel Generator (DG) Jacket Water Coolers, Safety Injection Pump (SIP) oil cooler, Centrifugal Charging Pump (CCP) oil cooler and the Containment Spray Pump (CSP) bearing oil coolers, and then return the SSW returns water back to the SSI.

1.5 The alternate or emergency source of water to Auxiliary Feedwater (AFW) is tapped off the outlet side of the Diesel Generator Jacket Water Coolers. This is not only a convenient location but also performs some preheating should SSW be required to supply the AFW system.

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ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 065 AK3.04 Importance Rating 3.0 Loss of Instrument Air: Knowledge of the reasons for the following responses as they apply to the Loss of Instrument Air:

Crossover to backup air supplies Proposed Question: Common 52 Which of the following is the reason the Turbine Driven Auxiliary Feedwater Pump steam supply valves are equipped with air accumulators?

Following a Loss of Instrument Air the air accumulators provide air to close and maintain closed its steam supply valve for...

A. ...30 minutes to isolate a faulted Steam Generator that has a steam or feed line break.

B. ...30 minutes following a Steam Generator Tube Rupture to mitigate radiological release.

C. ...7.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> to isolate a faulted Steam Generator that has a steam or feed line break.

D. ...7.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> following a Steam Generator Tube Rupture to mitigate radiological release.

Proposed Answer: D Explanation:

A. Incorrect. Plausible because the Turbine Driven Auxiliary Feedwater Pump is isolated in EOP-2.0 response, however the 30 minutes accumulator use time is for the AFW flow control valves during a faulted SG isolation.

B. Incorrect. Plausible because the accumulators are provided for SGTR isolation, however they are sized for 7.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />.

C. Incorrect. Plausible because the accumulators are sized for 7.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />, however they are provided for SGTR isolation.

D. Correct. The accumulators are sized for 7.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> to isolate a SGTR to minimize radiological release.

Technical Reference(s) LO21.SYS.IA1 Attached w/ Revision # See LO21.SYS.MR1 Comments / Reference Proposed references to be provided during examination: None Learning Objective: DESCRIBE the basic design and flow path of the Auxiliary Feedwater System.

Page 7 of 60 CPNPP NRC 2013 RO Written Exam Worksheet 51 to 64 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Question Source: Bank #

Modified Bank # (Note changes or attach parent)

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

Reference:

From LO21.SYS.IA1.LN, Page 19 Revision 05/07/11 INDIVIDUAL VALVE ACCUMULATORS Auxiliary Feedwater Control Valves & Recirculation Valves Accumulators are provided for the Auxiliary Feedwater control valves in the individual feed lines, and for the motor driven auxiliary feedwater pumps recirculation valves. These valves share accumulators.

These accumulators are sized to allow the operator remote, manual control to regulate flow or isolate a faulted steam generator for a period of 30 minutes after loss of air. The basis of the recirculation valves is to cycle one valve, then maintain the valve in the closed position for the remainder of the 30 minute period.

Turbine Driven Auxiliary Feedwater Pump (TDAFWP) Turbine Main Steam Isolation Valves Accumulators are provided for the turbine driven auxiliary feedwater pump (TDAFWP) turbine main steam isolation valves. These accumulators are sized to have sufficient air capacity to drive the valve closed and maintain it closed for 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> plus an additional one-half hour for the TDAFWP steam supply lines to then be locally-manually isolated by operator action, including any air leakage criteria.

Comments /

Reference:

From LO21.SYS.MR1.LN, Page 8 Revision 06/09/11 Turbine Driven Auxiliary Feedwater Pump Steam Supply Redundant valves and piping provide a secure steam supply to the Turbine Driven Auxiliary Feedwater Pump (TDAFWP) over a range of steam pressures from the lowest set main steam safety valve plus 3%

accumulation (1185 psig +3% = 1221 psig or 1236 psia) to the lowest SG pressure (100 psia on the secondary side) associated with RHR cut-in conditions at 350F. The design ensures the TDAFWP can be operated even if one steam supply valve fails to open upon demand.

Normally closed isolation valves to the TDAFWP have the capability to open upon loss of all AC power. Fail open air operated valves are used for this application therefore, air accumulators tanks are provided to permit valve closure following the loss of air.

These valves are required to be closed following a SGTR to mitigate radiological consequences. Thus the air accumulators must contain sufficient air capacity to drive each valve closed and maintain it closed post accident until an operator can be dispatched to manually isolate the associated steam line.

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From LO21.SYS.AF1.LN, Page 17 Revision # 05/11/11 TDAFWP TURBINE SPEED GOVERNOR The turbine speed governor is equipped with a pneumatically operated hydraulic speed changer mechanism which is used to control the turbine governor valve. The governor actuator is also driven by the turbine rotor via spiral reduction gears driven by the same worm gear as the oil pump. The governor servomotor receives oil from the actuator and is the lowest point in the oil system. To ensure proper operation, the oil level in the governor oil sight glass should be maintained at the levels specified in the operating procedures (U1 and U2 require slightly different levels). Obviously, a low oil level would damage the regulator or prevent operation, but high oil levels are also undesirable. This is because the oil may foam causing the regulator to operate improperly.

Comments /

Reference:

From LO21.SYS.AF1.LN, Page 9 Revision # 05/11/11 The tank is provided with a diaphragm to prevent oxygenation of the stored auxiliary feedwater. The tank is also provided with two 12" vents on top of the tank above the floating diaphragm for atmospheric relief. Nitrogen connections below the diaphragm allow for inerting the tank.

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ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 077 G 2.4.1 Importance Rating 4.6 Generator Voltage and Electric Grid Disturbances: Emergency Procedures/Plan: Knowledge of EOP entry conditions and immediate action steps Proposed Question: Common 53 Given the following conditions:

• The Unit 1 Reactor was tripped following a 345 kV grid disturbance.

• Immediate Actions of EOP-0.0A, Reactor Trip or Safety Injection are being performed.

• The Balance of Plant Operator reports the following:

• Buses 1EA1 and 1EA2 are 6250 Volts and lowering.

• Both buses are powered from their Alternate Offsite source.

• All other Immediate Action, Action/Expected Responses are Normal.

What is the appropriate action?

A. Transition to ECA-0.0A, Loss of All AC Power.

B. Transition to EOS-0.1A, Reactor Trip Response.

C. Transition to ABN-601, Response to a 138/345 KV System Malfunction.

D. Transition to ABN-602, Response to a 6900/480 V System Malfunction.

Proposed Answer: A Explanation:

A. Correct. Per EOP-0.0A, Step 3 if voltage is less than 6500 V on both busses a transition to ECA-0.0A is required.

B. Incorrect. Plausible as both buses are energized, that all Action/Expected Responses are satisfied, and a transition to EOS-0.1A is appropriate.

C. Incorrect. Plausible because ABN-601 is the RNO for EOP-0.0A step 3.b.

D. Incorrect. Plausible because ABN-602 is the RNO for EOP-0.0A step 3.b.

Technical Reference(s) EOP-0.0A, Step 3 Attached w/ Revision # See Comments / Reference Proposed references to be provided during examination: None Page 10 of 60 CPNPP NRC 2013 RO Written Exam Worksheet 51 to 64 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Learning Objective: ANALYZE the immediate operator actions of 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 7, 10 55.43 Page 11 of 60 CPNPP NRC 2013 RO Written Exam Worksheet 51 to 64 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From EOP-0.0A, Step 3 Revision # 8 Page 12 of 60 CPNPP NRC 2013 RO Written Exam Worksheet 51 to 64 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # W/E04 EK2.2 Importance Rating 3.8 LOCA Outside Containment: Knowledge of the interrelations between LOCA Outside Containment and the following:

Facility's heat removal systems, including primary coolant, emergency coolant, the decay heat removal systems and relations between the proper operation of these systems to the operation of the facility Proposed Question: Common 54 Given the following conditions:

• Unit 1 is responding to a Loss of Coolant Accident (LOCA) per ECA-1.2A, LOCA Outside Containment.

• The crew believes that the leak outside of Containment has been isolated but Reactor Coolant System pressure is not rising.

Which of the following alternate indications may be used to determine if the break has been isolated per ECA-1.2A, LOCA Outside Containment?

A. Refueling Water Storage Tank level stable.

B. Reactor Vessel Level Indicating System indication rising.

C. Emergency Core Cooling System flows rising.

D. Emergency Core Cooling System alignment verification.

Proposed Answer: B Explanation:

A. Incorrect. Plausible because RWST inventory could be lost to the Safeguards Building from an interfacing system break outside Containment and RWST level stable could indicate that break flow has ceased, however, RWST level may still be lowering with the break isolated as ECCS flow refills the RCS. Therefore, RWST level change is not a good indicator of break isolation.

B. Correct. As stated in Attachment 2, Step 3 Bases, RCS pressure may not initially rise once the break is isolated, due to plant cooldown or when the RCS is saturated. RVLIS indication rising shows that ECCS flow is not leaving the RCS via the break, but rather it is refilling the RCS. This indicates that the break is isolated from the RCS.

C. Incorrect. Plausible because ECCS flows could indicate that the break is isolated if it were lowering. Rising ECCS flow is indicative of the RCS break becoming worse.

D. Incorrect. Plausible because an ECCS valve alignment is performed in Step 1 of ECA-1.2A, and this may isolate the break, however, verifying ECCS alignment alone does not ensure that the break is isolated. RCS parameters, such as pressure and RVLIS trend must be evaluated to verify break isolation.

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ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Technical Reference(s) ECA-1.2A, Attachment 2, Step 3 Bases Attached w/ Revision # See ECA-1.2A, Steps 1, 2, & 3 Comments / Reference Proposed references to be provided during examination: None Learning Objective: STATE the bases for operator actions, notes and cautions from ECA-1.2, LOCA Outside Containment.

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 2, 5, 10, 14 55.43 Comments /

Reference:

From ECA-1.2 A, Attachment 2, Step 3 Bases Revision # 8 Page 14 of 60 CPNPP NRC 2013 RO Written Exam Worksheet 51 to 64 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ECA-1.2A, Steps 1 & 2 Revision # 8 Page 15 of 60 CPNPP NRC 2013 RO Written Exam Worksheet 51 to 64 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

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From ECA-1.2A, Step 3 Revision # 8 Page 16 of 60 CPNPP NRC 2013 RO Written Exam Worksheet 51 to 64 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # W/E05 G 2.2.3 Importance Rating 3.8 Inadequate Heat Transfer - Loss of Secondary Heat Sink: Equipment Control: (multi-unit) Knowledge of the design, procedural, and operational differences between units Proposed Question: Common 55 Given the following conditions:

• Unit 2 has experienced a Loss of All Feedwater Flow following a Reactor Trip from 100% power.

• 1/2-8000A, PRZR PORV BLK VLV, was previously closed due to 1/2-PCV-455A, PRZR PORV, seat leakage.

• FRH-0.1B, Response to Loss of Secondary Heat Sink, is in progress.

• Power can NOT be restored to 480V MCC 2EB3-2 which provides power to 1/2-8000A, PRZR PORV BLK VLV.

Which of the following is required per FRH-0.1B, Response to Loss of Secondary Heat Sink?

In order for Bleed and Feed to be initiated, Steam Generator (SG) wide range levels must be less than A. 27% in at least 3 SGs. Adequate core cooling will be achieved during Bleed and Feed operations.

B. 35% in at least 3 SGs. Adequate core cooling will be achieved during Bleed and Feed operations.

C. 27% in at least 3 SGs. Inadequate core cooling will be experienced during Bleed and Feed operations without opening Reactor Head and Pressurizer vents.

D. 35% in at least 3 SGs. Inadequate core cooling will be experienced during Bleed and Feed operations without opening Reactor Head and Pressurizer vents.

Proposed Answer: C Page 17 of 60 CPNPP NRC 2013 RO Written Exam Worksheet 51 to 64 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible because 27% is the required Steam Generator level for Unit 2, however, adequate core cooling will not be achieved during Bleed and Feed operations.

B. Incorrect. Plausible if thought that 35% is the required Steam Generator level for Unit 2 and that adequate core cooling would be achieved during Bleed and Feed operations with only one PORV available.

C. Correct. This is the required Steam Generator level for Unit 2. Inadequate core cooling will be experienced without opening Reactor Head and Pressurizer vents because only one PORV is available.

D. Incorrect. Plausible because inadequate core cooling will be experienced without opening Reactor Head and Pressurizer vents, however, a Steam Generator level of 35% is required for Unit 1.

Technical Reference(s) FRH-0.1B, Steps 3, 20, & 21 Attached w/ Revision # See FRH-0.1A, Steps 3, 20, & 21 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 # X Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7, 10 55.43 Page 18 of 60 CPNPP NRC 2013 RO Written Exam Worksheet 51 to 64 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From FRH-0.1B, Step 3 Revision # 8 Page 19 of 60 CPNPP NRC 2013 RO Written Exam Worksheet 51 to 64 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From FRH-0.1B, Steps 20 & 21 Revision # 8 Page 20 of 60 CPNPP NRC 2013 RO Written Exam Worksheet 51 to 64 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From FRH-0.1A, Step 3 Revision # 8 Page 21 of 60 CPNPP NRC 2013 RO Written Exam Worksheet 51 to 64 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From FRH-0.1A, Step 20 & 21 Revision # 8 Page 22 of 60 CPNPP NRC 2013 RO Written Exam Worksheet 51 to 64 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # W/E11 EA1.3 Importance Rating 3.7 Loss of Emergency Coolant Recirculation: Ability to operate and/or monitor the following as they apply to a Loss of Emergency Coolant Recirculation: Desired operating results during abnormal and emergency situations Proposed Question: Common 56 Given the following condition:

• ECA-1.2A, LOCA Outside Containment, Step 3, directs checking Reactor Coolant System pressure to determine if the break has been isolated by previous actions.

If the break has NOT been isolated, which of the following identifies the effect that a transition to ECA-1.1A, Loss of Emergency Coolant Recirculation, has on mitigating the accident?

Actions are taken to...

A. ...transfer Safeguards Building Sump to the Refueling Water Storage Tank to extend ECCS pump availability.

B. ...decrease total injection flow to minimize Refueling Water Storage Tank depletion.

C. ...increase the injection flow rate to maintain Reactor Coolant System heat removal.

D. ...stabilize Reactor Coolant System pressure to prevent the Safety Injection Accumulators from discharging out the break.

Proposed Answer: B Explanation:

A. Incorrect. Plausible because this action could be performed to recover lost sump water, however, it is not directed by ECA-1.1A. This action is based on ECA-1.1, major action category B, increase/conserve RWST level.

B. Correct. Reducing injection flow will minimize RWST depletion.

C. Incorrect. Plausible if thought that injection flowrate is increased in accordance with ECA-1.1 major action category G, maintain RCS heat removal.

D. Incorrect. Plausible based on misconception that it is desirable to prevent SI Accumulators from discharging in accordance with ECA-1.1 major action category D, depress RCS to minimize RCS break flow.

Technical Reference(s) ECA-1.1A, Steps 17 & 22 Attached w/ Revision # See ECA-1.1A, flowchart Comments / Reference Page 23 of 60 CPNPP NRC 2013 RO Written Exam Worksheet 51 to 64 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Proposed references to be provided during examination: None Learning Objective: Given a procedural step, or sequence of steps from ECA-1.1, Loss of Emergency Coolant Recirculation, STATE the purpose/basis for the step(s).

Question Source: Bank # X 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 2, 10 55.43 Page 24 of 60 CPNPP NRC 2013 RO Written Exam Worksheet 51 to 64 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ECA-1.1A, Step 17 Revision # 8 Page 25 of 60 CPNPP NRC 2013 RO Written Exam Worksheet 51 to 64 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ECA-1.1A, Step 22 Revision # 8 Page 26 of 60 CPNPP NRC 2013 RO Written Exam Worksheet 51 to 64 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ECA-1.2A, Step 3 Revision # 8 Comments /

Reference:

From ECA-1.1A, Flowchart Revision 8 Page 27 of 60 CPNPP NRC 2013 RO Written Exam Worksheet 51 to 64 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 2 K/A # 003 AK1.03 Importance Rating 3.5 Dropped Control Rod: Knowledge of the operational implications of the following concepts as they apply Dropped Control Rod: Relationship of reactivity and reactor power to rod movement.

Proposed Question: Common 57 Given the following conditions:

• Unit 1 is operating at Middle-of-Life 100% power.

• 1/1-RBSS CONTROL ROD BANK SELECT is in MANUAL.

• T AVE -T REF Deviation is 0ºF and stable.

• Control Bank D Control Rod H-8 drops to the bottom of the core.

After the Control Rod drops the following conditions exist:

• T AVE -T REF Deviation is -7.5ºF and stable.

• Reactor Power is 99% and stable.

What is the reactivity worth of Control Rod H-8?

A. 153 pcm B. 138 pcm C. 133 pcm D. 123 pcm Proposed Answer: A Page 28 of 60 CPNPP NRC 2013 RO Written Exam Worksheet 51 to 64 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Explanation:

A. Correct. In accordance with the provided MOL Reactivity Briefing Sheet a 7.5ºF lower temperature would be equivalent to -7.5 x -18.4 = 138 pcm/ºF plus a 1% power change which would be equivalent to 1 x 15.3 = 15.3 pcm/% RTP; therefore the total reactivity worth of the Control Rod would be 138 + 15.3 = 153 pcm.

B. Incorrect. Plausible as 138 pcm would be equivalent to the reactivity change from the temperature change only.

C. Incorrect. Plausible as 133 pcm would be calculated if the power and temperature coefficients were reversed in the calculation, -7.5 x 15.3 = 115 plus 1 x 18.4 = 18; 115 + 18 = 133 pcm.

D. Incorrect. Plausible as 123 pcm would be calculated if the reactivity change from power was subtracted from the reactivity change due to temperature; 138 - 15.3 = 123 pcm.

Technical Reference(s) Reactivity Briefing Sheet for MOL Attached w/ Revision # See Comments / Reference Proposed references to be provided during examination: Reactivity Briefing Sheet for MOL Learning Objective: ANALYZE the response to a Dropped or Misaligned rod in MODE 1 or 2 in accordance with ABN-712, Rod Control 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 1 55.43 Page 29 of 60 CPNPP NRC 2013 RO Written Exam Worksheet 51 to 64 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From Reactivity Briefing Sheet for MOL Revision 2/20/12 Page 30 of 60 CPNPP NRC 2013 RO Written Exam Worksheet 51 to 64 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 2 K/A # 028 AA1.06 Importance Rating 3.3 Pressurizer Level Control Malfunction: Ability to operate and/or monitor the following as they apply to Pressurizer Level Control Malfunctions: Checking of RCS leaks Proposed Question: Common 58 Given the following conditions:

• Unit 1 is at 100%.

• Pressurizer level is stable.

• 1-TI-126, Regenerative Heat Exchanger Letdown Outlet temperature is 250ºF.

• 1-TI-127, Regenerative Heat Exchanger Charging Return temperature is 500ºF.

• 1-PI-131, Letdown Heat Exchanger Outlet pressure is 310 psig.

A Chemical and Volume Control System malfunction occurs resulting in the following conditions:

• Pressurizer level is lowering.

• 1-TI-126, Regenerative Heat Exchanger Letdown Outlet temperature is 390ºF.

• 1-TI-127, Regenerative Heat Exchanger Charging Return temperature is 540ºF.

• 1-PI-131, Letdown Heat Exchanger Outlet pressure is 310 psig.

Which of the following malfunctions would explain the observed change in indications?

A. A Charging Header leak upstream of the Regenerative Heat Exchanger.

B. A Letdown leak upstream of the Regenerative Heat Exchanger.

C. Loss of air to 1-FCV-0121, Charging Header Flow Control Valve.

D. Loss of air to 1-PCV-131, Letdown Pressure Control Valve.

Proposed Answer: A Page 31 of 60 CPNPP NRC 2013 RO Written Exam Worksheet 51 to 64 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Explanation:

A. Correct. A leak in this location would divert cooling flow (Charging) from the Regenerative Heat Exchanger resulting in higher Letdown temperatures out of the Regenerative Heat Exchanger and the Charging temperature returning to the loop would also be higher than normal and also result in lowering Pressurizer level.

B. Incorrect. Plausible because it could be thought that this leak location would result in these indications, however, a leak in Letdown upstream of the Regenerative Heat Exchanger would reduce the flow though the Heat Exchanger and result in a lower rather than higher temperature.

C. Incorrect. Plausible if thought that this failure reduced the flow of Charging and caused higher Letdown temperatures from the Regenerative Heat Exchanger and lowering Pressurizer level, however, the valve fails OPEN which would yield lowering Letdown temperatures from the Regenerative Heat Exchanger and rising Pressurizer level.

D. Incorrect. Plausible because this failure would result in the higher Letdown temperatures from the Regenerative Heat Exchanger and lowering Pressurizer level, however, on a loss of air to the Letdown Pressure Control Valve the Letdown pressure would be much lower than normal instead of at or slightly above normal pressure.

Technical Reference(s) LO21.SYS.CS1 Attached w/ Revision # See ALM-0061A, 1-ALB-6A, Window 1.4 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 # X Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 3, 5, 7 55.43 Page 32 of 60 CPNPP NRC 2013 RO Written Exam Worksheet 51 to 64 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From LO21.SYS.CS1.LN, Page 12 Revision # 04/28/11 REGENERATIVE HEAT EXCHANGER The regenerative heat exchanger is a stainless steel heat exchanger, located in containment in a shielded room on the 832' elevation. It is a counter flow design with charging return flow through the tube side and letdown flow on the shell side. Letdown enters the heat exchanger at reactor coolant system cold leg temperature and pressure and passes through baffles, which create crossflow to increase the contact between the hot letdown and the relatively cold charging flow. This allows more heat transfer between the two fluids. By utilizing a regenerative heat exchanger, the amount of heat lost from the primary system and the thermal stresses that exist at the loop charging penetrations are reduced.

During normal power operations the regenerative heat exchanger reduces letdown temperature from 560°F to approximately 260°F and raises charging water temperature to approximately that of the reactor coolant system cold loop.

Temperature of the charging flow leaving the regenerative heat exchanger is provided by thermowell-mounted RTD and indicated on CB-06 (u-TI-0126, 100-600°F) and on the plant computer.

Temperature of the letdown flow leaving the regenerative heat exchanger is provided by thermowell-mounted RTD and indicated on CB-06 (u-TI-0127, 100-600°F) and on the plant computer. This device also actuates an alarm (REGEN HX LTDN OUT TEMP HI) at >400°F.

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ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From LO21.SYS.CS1.LN, Page 18 Revision # 04/28/11 LETDOWN HEAT EXCHANGER OUTLET PRESSURE CONTROL VALVE Letdown Heat Exchanger Outlet Pressure Control Valve, u-PCV-0131 is located downstream of the letdown heat exchanger, in the letdown heat exchanger valve room. During normal power operations the valve functions to automatically maintain upstream pressure at the established setpoint, which is normally 310 psig. By maintaining a back-pressure on the system, the hot letdown fluid between the letdown orifices and the letdown heat exchanger is maintained in a subcooled condition.

The saturation temperature of water at 310 psig is approximately 425°F. The temperature of letdown leaving the regenerative heat exchanger is normally 260°F, which is approximately 165°F below saturation temperature. An alarm (REGEN HX LTDN OUT TEMP HI, set at >400°F) will warn of temperature approaching saturation temperature for this portion of the letdown piping.

Letdown Heat Exchanger Outlet Pressure Control Valve, u-PCV-0131, is air operated and fails fully open on a loss of air or control power.

Comments /

Reference:

From LO21.SYS.CS1.LN, Page 40 Revision # 04/28/11 CENTRIFUGAL CHARGING PUMP FLOW CONTROL Charging Flow Element u-FE-0121 is located on the combined charging pump discharge header in the charging pump valve room. It provides the differential pressure which is related to charging flow to Charging Pump Discharge Flow Transmitter u-FT-0121 (located, for Unit 1, in the boric acid storage tank room and, for Unit 2, in the Unit 2 CVCS valve operating room on the 822 foot elevation of the auxiliary building. The flow transmitter functions to generate a current signal which is proportional to charging header flow (from 0 to 270 gpm) for indication and control of charging flow. Charging Flow Indicators u-FI-0121A and u-FI-0121B (0 to 270 gpm) are located on CB-06 and at the remote shutdown panel, respectively. Charging flow is provided as an input to the plant computer. The CHG FLO HI/LO alarm on CB-06 also receives its inputs from the charging flow transmitter and is set to actuate at > 150 gpm and at < 55gpm.

The discharge flow from the centrifugal charging pumps to the normal charging header and to the reactor coolant pump seal injection lines is controlled by regulating the position of CCP u-01/u-02 Charging Flow Control Valve, u-FCV-0121, located in the charging pump valve room. This valve is air operated, and fails open on loss of air or control power.

Page 34 of 60 CPNPP NRC 2013 RO Written Exam Worksheet 51 to 64 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ALM-0061A, 1-ALB-6A, Window 1.4 Revision # 7 Page 35 of 60 CPNPP NRC 2013 RO Written Exam Worksheet 51 to 64 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 2 K/A # 060 G 2.4.45 Importance Rating 4.1 Accidental Gaseous Radwaste Release: Emergency Procedures/Plan: Ability to prioritize and interpret the significance of each annunciator or alarm Proposed Question: Common 59 Given the following conditions:

• Unit 2 is at 100% power.

• The Digital Radiation Monitoring System (PC-11) receives a high radiation alarm (RED) on X-RE-5701, Auxiliary Building Ventilation Duct Radiation Monitor.

Which of the following describes the expected automatic action initiated due to the high radiation alarm?

A. Any Containment Vent is terminated.

B. Any Gas Decay Tank release is terminated.

C. Containment Ventilation Isolation is actuated.

D. Control Room Emergency Recirculation is actuated.

Proposed Answer: B Explanation:

A. Incorrect. Plausible if thought that the Containment Vent release was sampled by the Auxiliary Building Ventilation Duct Radiation Monitor.

B. Correct. A high radiation signal from X-RE-5701 automatically closes HCV-014 which terminates any waste gas release that may be in progress.

C. Incorrect. Plausible because Containment Ventilation Isolation is caused by a high radiation signal, but the signal is from 2-RE-5502 or 2-RE-5503, the Containment Air Particulate and Gaseous channels.

D. Incorrect. Plausible because Control Room Emergency Recirculation is caused by a high radiation signal, but the signal is from X-RE-5895A/B or X-RE-5896A/B, the Control Room Air Supply gas channel.

Technical Reference(s) ABN-902, Section 2.2 Attached w/ Revision # See Comments / Reference Proposed references to be provided during examination: None Page 36 of 60 CPNPP NRC 2013 RO Written Exam Worksheet 51 to 64 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Learning Objective: COMPREHEND the normal, abnormal and emergency operations of the Gaseous Waste Systems.

EXPLAIN the instrumentation and controls of the Digital Radiation Monitoring System and PREDICT the system response.

Question Source: Bank # X 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 11 55.43 Page 37 of 60 CPNPP NRC 2013 RO Written Exam Worksheet 51 to 64 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ABN-902, Sections 2.1 & 2.2 Revision # 7 Page 38 of 60 CPNPP NRC 2013 RO Written Exam Worksheet 51 to 64 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ABN-902, Sections 2.1 & 2.2 Revision # 7 Page 39 of 60 CPNPP NRC 2013 RO Written Exam Worksheet 51 to 64 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 2 K/A # 067 AA2.08 Importance Rating 2.9 Plant Fire on Site: Ability to determine and interpret the following as they apply to the Plant Fire on Site: Limits of affected area Proposed Question: Common 60 Given the following conditions:

• A fire has been reported in the Nuclear Operations Support Facility (NOSF).

• The Fire Brigade has been dispatched due to the Plant Simulator being housed in the NOSF.

Which of the following describes the actions required following dispatch of the Fire Brigade to the NOSF?

The [1] shall request assistance from [2] to restore the Fire Brigade complement inside the Protected Area as soon as possible.

[1] [2]

A. Shift Manager Glen Rose Fire Department B. Fire Brigade Leader Glen Rose Fire Department C. Shift Manager Granbury Fire Department D. Fire Brigade Leader Granbury Fire Department Proposed Answer: A Explanation:

A. Correct. The Shift Manager is responsible to ensure that the Fire Brigade inside the Protected Area is restored to acceptable strength as soon as possible and Glen Rose is the CPNPP designated backup Fire Department.

B. Incorrect. Plausible because Glen Rose is the CPNPP designated backup Fire Department, however, the Shift Manager is responsible to restore the Fire Brigade.

C. Incorrect. Plausible because the Shift Manager is responsible to restore the Fire Brigade, however, the Glen Rose Fire Department is the CPNPP designated backup Fire Department.

D. Incorrect. Plausible if thought that the Fire Brigade leader requests backup and that Granbury is the CPNPP designated backup Fire Department.

Page 40 of 60 CPNPP NRC 2013 RO Written Exam Worksheet 51 to 64 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Technical Reference(s) STA-727, Step 6.3.4 Attached w/ Revision # See Comments / Reference Proposed references to be provided during examination: None Learning Objective: During unit evolutions, DIRECT shift personnel actions and ENSURE proper and effective communications are maintained.

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 41 of 60 CPNPP NRC 2013 RO Written Exam Worksheet 51 to 64 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From STA-727, Step 6.3.4 Revision # 5 Page 42 of 60 CPNPP NRC 2013 RO Written Exam Worksheet 51 to 64 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 2 K/A # 068 AA2.09 Importance Rating 4.1 Control Room Evacuation: Ability to determine and interpret the following as they apply to the Control Room Evacuation:

Saturation margin Proposed Question: Common 61 Given the following conditions:

• The Unit 1 Control Room has been evacuated per ABN-905A, Loss of Control Room Habitability.

• A plant cooldown from the Remote Shutdown Panel has been initiated with the following limits:

• Subcooling greater than 65°F.

• Actual Pressurizer Level - 25% to 50%.

• Actual Steam Generator Level - 84% to 92%.

Which of the following describes how the above limits are monitored per ABN-905A, Loss of Control Room Habitability?

Subcooling is determined by __________, Actual Pressurizer Level is determined by

__________, and Actual Steam Generator level is determined by __________.

A. reading RCS Saturation meter using temperature correction of indicated level using temperature correction of indicated level B. plotting indicated temperature and pressure using temperature correction of indicated level using temperature correction of indicated level C. reading RCS Saturation meter reading Pressurizer level meter reading Steam Generator level meter D. plotting indicated temperature and pressure reading Pressurizer level meter reading Steam Generator level meter Page 43 of 60 CPNPP NRC 2013 RO Written Exam Worksheet 51 to 64 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Proposed Answer: B Explanation:

A. Incorrect. Plausible because Pressurizer and Steam Generator levels are calculated using temperature correction curves of indicated level per Attachments 17 and 16 respectively, however, there is no RCS Saturation meter at the Remote Shutdown Panel.

B. Correct. Subcooling must be calculated using RCS temperature and pressure at the Remote Shutdown Panel per Attachment 13. Pressurizer and Steam Generator levels are calculated using temperature correction curves of indicated level per Attachments 17 and 16 respectively.

C. Incorrect. Plausible if thought that there is an RCS Saturation meter at the Remote Shutdown Panel and temperature corrections must be applied for Pressurizer and Steam Generator levels.

D. Incorrect. Plausible because Subcooling must be calculated using RCS temperature and pressure at the Remote Shutdown Panel, however, temperature corrections must be applied for Pressurizer and Steam Generator levels.

Technical Reference(s) ABN-905A, Step 2.3.58 Attached w/ Revision # See ABN-905A, Attachments 13, 16, & 17 Comments / Reference Proposed references to be provided during examination: None Learning Objective: ANALYZE the response to a Loss Of Control Room Habitability per ABN-905, Loss Of Control Room Habitability.

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7, 10 55.43 Page 44 of 60 CPNPP NRC 2013 RO Written Exam Worksheet 51 to 64 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ABN-905A, Step 2.3.58 Revision # 9 Page 45 of 60 CPNPP NRC 2013 RO Written Exam Worksheet 51 to 64 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ABN-905A, Attachment 13 Revision # 9 Page 46 of 60 CPNPP NRC 2013 RO Written Exam Worksheet 51 to 64 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ABN-905A, Attachment 16 Revision # 9 Page 47 of 60 CPNPP NRC 2013 RO Written Exam Worksheet 51 to 64 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ABN-905A, Attachment 17 Revision # 9 Page 48 of 60 CPNPP NRC 2013 RO Written Exam Worksheet 51 to 64 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 2 K/A # 076 AK2.01 Importance Rating 2.6 High Reactor Coolant Activity: Knowledge of the interrelations between High Reactor Coolant Activity and the following:

Process radiation monitors Proposed Question: Common 62 Given the following conditions:

• Unit 1 has experienced a problem with the Volume Control Tank (VCT).

• Charging Pump suction has been shifted to the Refueling Water Storage Tank (RWST) per SOP-103A, Chemical and Volume Control System.

• Chemistry has sampled the Reactor Coolant System and determined that Co-58 and Co-60 levels are increasing.

Which of the following lists the expected indication and the most probable cause for the indication?

1-RE-406 (FFL160) indication A. rising at a steady rate due to an oxygen induced CRUD burst.

B. rising at a steady rate due to oxygen induced cladding creep.

C. spiking and returning to normal due to an oxygen induced CRUD burst.

D. spiking and returning to normal due to oxygen induced cladding creep.

Proposed Answer: A Explanation:

A. Correct. An increase in Co-58 and Co-60 are the result of oxygen induced CRUD burst from shifting to the Refueling Water Storage Tank.

B. Incorrect. Plausible because a FFL160 indication would be rising but cobalt is not an indication of cladding damage.

C. Incorrect. Plausible because FFL160 indication would rise not spike and return to normal and cobalt comes from a CRUD burst not clad failure.

D. Incorrect. Plausible because FFL160 indication would rise not spike and return to normal and cobalt comes from a CRUD burst not clad failure.

Technical Reference(s) ABN-102, Steps 1, 6 & 7 NOTES Attached w/ Revision # See SOP-103A, Step 5.5.15.A CAUTION Comments / Reference Page 49 of 60 CPNPP NRC 2013 RO Written Exam Worksheet 51 to 64 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Proposed references to be provided during examination: None Learning Objective: ANALYZE the response to High Reactor Coolant Activity per 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 10, 12 55.43 Comments /

Reference:

From SOP-103A, Step 5.5.15.A CAUTION Revision # 17 Page 50 of 60 CPNPP NRC 2013 RO Written Exam Worksheet 51 to 64 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ABN-102, Step 7 NOTE Revision # 7 Page 51 of 60 CPNPP NRC 2013 RO Written Exam Worksheet 51 to 64 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ABN-102, Step 1 & 6 NOTES Revision # 7 Page 52 of 60 CPNPP NRC 2013 RO Written Exam Worksheet 51 to 64 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 2 K/A # W/E08 EK3.1 Importance Rating 3.4 RCS Overcooling - PTS: Knowledge of the reasons for the following responses as they apply to the Pressurized Thermal Shock: Facility operating characteristics during transient conditions, including coolant chemistry and the effects of temperature, pressure, and reactivity changes and operating limitations and reasons for these operating characteristics Proposed Question: Common 63 What is the reason for terminating Safety Injection flow during the performance of FRP-0.1A, Response to Imminent Pressurized Thermal Shock Conditions?

A. Prevent entry into a condition to the right of Limit A on the Integrity Status Tree.

B. Correct cause of event to allow exit from FRP-0.1A prior to completion.

C. Maintain RWST inventory in the event of flaw propagation.

D. Stop RCS cooldown and minimize RCS pressure increase.

Proposed Answer: D Explanation:

A. Incorrect. Plausible because to the right of Limit A is a branch on the Integrity Safety Function Status Tree, however, this is a desirable position during a PTS event.

B. Incorrect. Plausible because correcting cause would be desired but once entered FRP-0.1A must be completed.

C. Incorrect. Plausible because terminating SI would conserve inventory but it is not the reason stated in the bases for FRP-0.1A, Step 7.

D. Correct. Following SI actuation, RCS conditions may be restored to within acceptable limits for SI termination to be allowed. ECCS flow may have contributed to the RCS cooldown or may prevent a subsequent reduction in RCS pressure.

Technical Reference(s) FRP-0.1A, Attachment 4, Step 7 Bases Attached w/ Revision # See FRP-0.1A, CSFST Comments / Reference Proposed references to be provided during examination: None Page 53 of 60 CPNPP NRC 2013 RO Written Exam Worksheet 51 to 64 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 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 Conditions.

Question Source: Bank #

Modified Bank # X (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 54 of 60 CPNPP NRC 2013 RO Written Exam Worksheet 51 to 64 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From FRP-0.1A, Attachment 4, Step 7 Bases Revision # 8 Page 55 of 60 CPNPP NRC 2013 RO Written Exam Worksheet 51 to 64 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From FRP-0.1A, CSFST Revision # 8 Page 56 of 60 CPNPP NRC 2013 RO Written Exam Worksheet 51 to 64 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Original Bank Question Page 57 of 60 CPNPP NRC 2013 RO Written Exam Worksheet 51 to 64 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 2 K/A # W/E10 EK3.3 Importance Rating 3.4 Natural Circulation with Steam Void in Vessel with/without RVLIS: Knowledge of the reasons for the following responses as they apply to the Natural Circulation with Steam Void in Vessel with/without RVLIS: Manipulation of controls required to obtain desired operating results during abnormal and emergency situations Proposed Question: Common 64 During the performance of EOS-0.4B, Natural Circulation with Steam Void in Vessel (without RVLIS), when starting the first Reactor Coolant Pump, ensure pressurizer level is A. ...between 30% and 40% to compensate for the level increase due to the void formation.

B. ...between 30% and 40% to compensate for the level increase due to pump start.

C. ...above 90% to compensate for the level decrease due to the void collapsing.

D. ...above 90% to compensate for the level decrease due to eliminating a non-condensable void.

Proposed Answer: C Explanation:

A. Incorrect. Plausible because IF a Reactor Coolant Pump CANNOT be started, EOS-0.4B contains guidance to maintain Pressurizer level between 30% and 40%. However, level will decrease due to void collapse when a Reactor Coolant Pump is started, and must be established at greater than 90%.

B. Incorrect. Plausible because IF a Reactor Coolant Pump CANNOT be started, EOS-0.4B contains guidance to maintain Pressurizer level between 30% and 40%. However, level will decrease due to void collapse when a Reactor Coolant Pump is started, and must be established at greater than 90%.

C. Correct. Pressurizer level is maintained above 90% to accommodate void collapse when starting the first Reactor Coolant Pump.

D. Incorrect. Plausible because the Pressurizer level value is correct, however, if non-condensable gases were indicated, they would be addressed via FRI-0.3B, Response to Voids in Reactor Vessel.

Technical Reference(s) EOS-0.4B, Step 1 & 2 Attached w/ Revision # See FRI-0.3B, Attachment 7, Bases Comments / Reference Proposed references to be provided during examination: None Page 58 of 60 CPNPP NRC 2013 RO Written Exam Worksheet 51 to 64 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Learning Objective: ANALYZE the recovery technique used and the procedure steps of EOS-0.3, Natural Circulation Cooldown with Steam Void in Vessel (with RVLIS).

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

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

Reference:

From EOS-0.4B, Step 1 & 2 Revision 8 Page 59 of 60 CPNPP NRC 2013 RO Written Exam Worksheet 51 to 64 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From FRI-0.3B, Attachment 7, Bases Revision 8 Page 60 of 60 CPNPP NRC 2013 RO Written Exam Worksheet 51 to 64 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 2 K/A # W/E15 EA1.1 Importance Rating 2.9 Containment Flooding: Ability to operate and/or monitor the following as they apply to Containment Flooding: Components and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes and automatic and manual features Proposed Question: Common 65 Given the following conditions:

• Unit 1 has experienced a Loss of Coolant Accident.

• Containment pressure is 8 psig and rising.

• While responding in EOP-1.0A, Loss of Reactor or Secondary Coolant, the Control Room observed Containment Sump level greater than 816' and entered FRZ-0.2A, Response to Containment Flooding.

• The Control Room is attempting to identify water volumes other than the Refueling Water Storage Tank or Safety Injection Accumulators that may be the source of the additional water in Containment.

Which of the following systems may still be contributing to the Containment Flooding?

A. Main Feedwater B. Fire Protection Water C. Ventilation Chilled Water D. Component Cooling Water Proposed Answer: D Explanation:

A. Incorrect. Plausible because Main Feedwater could be a contributor if not isolated on reactor trip with low Tave.

B. Incorrect. Plausible because Fire Protection Water could be a contributor if not isolated by Phase A containment isolation.

C. Incorrect. Plausible because Ventilation Chilled Water could be a contributor if not isolated by Phase A containment isolation.

D. Correct. Containment pressure has not reached 18 psig yet so Component Cooling water is not completely isolated to containment and could be the source.

Technical Reference(s) FRZ-0.2A, Step 1 Attached w/ Revision # See Comments / Reference Page 1 of 38 CPNPP NRC 2013 RO Written Exam Worksheet 65 to 75 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Proposed references to be provided during examination: None Learning Objective: DISCUSS the symptoms, or Entry Conditions for FRZ-0.2, Response to Containment Flooding.

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 9, 10 55.43 Page 2 of 38 CPNPP NRC 2013 RO Written Exam Worksheet 65 to 75 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From FRZ-0.2A, Step 1 Revision # 8 Page 3 of 38 CPNPP NRC 2013 RO Written Exam Worksheet 65 to 75 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 3 Category # 1 K/A # G 2.1.2 Importance Rating 4.1 Conduct of Operations: Knowledge of operator responsibilities during all modes of plant operation Proposed Question: Common 66 Given the following conditions:

• A non-licensed operator successfully completed the Generic Fundamentals Examination, and is enrolled in the Initial Licensed Operator Training Program.

• Unit 2 is at End-of-Life.

• BTRS Demineralizers will be placed in service during the Shift to dilute the Reactor Coolant System.

• The trainee requests to perform the dilution for training.

Who can approve the trainee performing the dilution and what requirement must be met by the trainee?

A. Shift Manager The trainee has completed Initial Non-Licensed Operator training classroom instruction for BTRS.

B. Shift Operations Manager The trainee has completed Initial Licensed Operator training classroom instruction for BTRS.

C. Shift Operations Manager The trainee has completed Initial Non-Licensed Operator training classroom instruction for BTRS.

D. Shift Manager The trainee has completed Initial Licensed Operator training classroom instruction for BTRS.

Proposed Answer: D Page 4 of 38 CPNPP NRC 2013 RO Written Exam Worksheet 65 to 75 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible because the Shift Manager must approve the use of a trainee for a reactivity manipulation, however, a trainee must have completed the classroom phase of license training covering the evolution in question prior to manipulating controls.

B. Incorrect. Plausible because one requirement for a trainee to manipulate controls is to be enrolled in a replacement licensed operator training program. However, the Shift Operations Manager is required to approve the use of a trainee for a reactor startup only and a trainee must have completed the classroom phase of license training covering the evolution in question prior to manipulating controls.

C. Incorrect. Plausible because one requirement for a trainee to manipulate controls is to be enrolled in a replacement licensed operator training program. However, the Shift Operations Manager is required to approve the use of a trainee for a reactor startup only and a trainee must have completed the classroom phase of license training covering the evolution in question prior to manipulating controls.

D. Correct. The Shift Manager must approve the use of a trainee for a reactivity manipulation.

Additionally, this evolution is permissible only if a person is enrolled in the Replacement License Program and has successfully completed classroom instruction.

Technical Reference(s) ODA-102, Step 6.24 Attached w/ Revision # See 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 # X 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 5 of 38 CPNPP NRC 2013 RO Written Exam Worksheet 65 to 75 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ODA-102, Step 6.24 Revision # 26 Page 6 of 38 CPNPP NRC 2013 RO Written Exam Worksheet 65 to 75 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 3 Category # 1 K/A # G 2.1.36 Importance Rating 3.0 Conduct of Operations: Knowledge of procedures and limitations involved in core alterations Proposed Question: Common 67 Given the following conditions:

• Unit 2 is in MODE 6.

• N-31 and N-32 are the OPERABLE Source Range Nuclear Instruments.

• STA-617, High Voltage Switching and Clearance, is about to be performed in the Switchyard.

Which of the following must be performed prior to implementing STA-617, High Voltage Switching and Clearance?

A. Place the High Flux at Shutdown Switch in BLOCK on both N-31 and N-32 to prevent loss of the Source Range Nuclear Instrumentation.

B. Place the High Flux at Shutdown Switch in BLOCK on both N-31 and N-32 to prevent a Containment evacuation.

C. Suspend CORE ALTERATIONS and positive reactivity additions due to the potential for spiking of the Source Range Nuclear Instrumentation.

D. Suspend CORE ALTERATIONS and positive reactivity additions due to the potential for loss of power to Refueling equipment.

Proposed Answer: C Explanation:

A. Incorrect. Plausible because spiking of the Source Range Nuclear Instrumentation will occur, however, placing the High Flux at Shutdown Switch in BLOCK would violate Technical Specifications.

B. Incorrect. Plausible because spiking could activate the Containment Evacuation Alarm, however, the switch would not be placed in BLOCK.

C. Correct. Per the Precaution outlined in RFO-102.

D. Incorrect. Plausible because CORE ALTERATIONS would be suspended, however, not for the reasons listed.

Technical Reference(s) RFO-102, Steps 3.13 & 3.17 Attached w/ Revision # See Comments / Reference Page 7 of 38 CPNPP NRC 2013 RO Written Exam Worksheet 65 to 75 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Proposed references to be provided during examination: None Learning Objective: EXPLAIN indication and Control/Trips for Source Range High Flux at Shutdown and Containment Evacuation Alarms.

Question Source: Bank # X 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 6, 10 55.43 Page 8 of 38 CPNPP NRC 2013 RO Written Exam Worksheet 65 to 75 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From RFO-102, Step 3.13 Revision # 13 Page 9 of 38 CPNPP NRC 2013 RO Written Exam Worksheet 65 to 75 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From RFO-102, Step 3.17 Revision # 13 Page 10 of 38 CPNPP NRC 2013 RO Written Exam Worksheet 65 to 75 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 3 Category # 2 K/A # G 2.2.18 Importance Rating 2.6 Equipment Control: Knowledge of the process for managing maintenance activities during shutdown operations, such as risk assessments, work prioritization, etc.

Proposed Question: Common 68 Given the following conditions with Unit 1 in Reduced Inventory:

• Spent fuel is still in the Reactor Vessel.

• IPO-010A, Reactor Coolant System Reduced Inventory Operations, is in progress and Attachment 1, Shiftly Checklist, was completed by the previous shift.

• The oncoming crew has been on watch for 30 minutes.

• Maintenance wants to open the Instrument Air to Containment penetration for repairs.

• No other penetrations are opened for maintenance.

Which of the following describes the appropriate response to the request to open the Instrument Air to Containment penetration?

A. The penetration SHALL NOT be opened until Reactor Vessel level is raised above 80.

B. The penetration can be opened for maintenance with NO administrative controls.

C. The penetration can be opened for maintenance if tracked per Technical Specifications.

D. The penetration SHALL NOT be opened until a new IPO-010A, Attachment 1, Shiftly Checklist is completed.

Proposed Answer: C Page 11 of 38 CPNPP NRC 2013 RO Written Exam Worksheet 65 to 75 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible if thought that no penetrations may be opened in Reduced Inventory conditions.

B. Incorrect. Plausible if thought that since no other penetrations are impaired one does not have any administrative controls.

C. Correct. Up to ten penetrations may be impaired per IPO-010A, Attachment 1 if tracked and can be sealed if needed.

D. Incorrect. Plausible if thought that the Shiftly Checklist from the previous shift is not adequate.

Technical Reference(s) IPO-010A, Attachment 1, 2.0.C Attached w/ Revision # See Comments / Reference Proposed references to be provided during examination: None Learning Objective: DISCUSS the Precautions, Limitations and key Attachments of WCI-401, Outreach Safety Function Guide.

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7, 10 55.43 Page 12 of 38 CPNPP NRC 2013 RO Written Exam Worksheet 65 to 75 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From IPO-010A, Attachment 1, 2.0.C Revision # 18 Page 13 of 38 CPNPP NRC 2013 RO Written Exam Worksheet 65 to 75 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 3 Category # 2 K/A # G 2.2.38 Importance Rating 3.6 Equipment Control: Knowledge of conditions and limitations in the facility license Proposed Question: Common 69 Given the following conditions:

• Unit 2 is in MODE 5.

• RCS temperature is currently 110ºF.

• Maintenance reports that Component Cooling Water (CCW) Pump 2-02 has a failed motor bearing.

Under these conditions, which of the following is the HIGHEST Reactor Coolant System temperature that Unit 2 can be increased to WITHOUT violating Technical Specifications?

A. 139ºF B. 199ºF C. 319ºF D. 349ºF Proposed Answer: B Explanation:

A. Incorrect. Plausible because Refueling outages are typically considered to be conducted after the RCS temperature is reduced below 140ºF, but both CCW Trains are only required to be OPERABLE in MODES 1-4.

B. Correct. Both CCW Trains are required to be OPERABLE prior to entry into MODE 4. The RCS temperature defining MODE 4 operations is 200ºF, so this is the highest temperature that can be achieved. There are no provisions of Technical Specification LCO 3.0.4 that would allow entry into MODE 4.

C. Incorrect. Plausible because a temperature of 320ºF appears throughout Tech Specs, primarily associated with LTOP, but both trains of CCW are required to be OPERABLE in MODES 1-4.

D. Incorrect. Plausible because a temperature of 350ºF would prevent entering MODE 3 from MODE 4, however, both trains of CCW are required to be OPERABLE in MODES 1-4.

Technical Reference(s) Technical Specification LCO 3.7.7 Attached w/ Revision # See Technical Specification Definitions 1.1 Comments / Reference Page 14 of 38 CPNPP NRC 2013 RO Written Exam Worksheet 65 to 75 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Proposed references to be provided during examination: None Learning Objective: APPLY the administrative requirements of the Component Cooling Water System including Technical Specifications, TRM and ODCM.

Question Source: Bank # X 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:

From Technical Specification LCO 3.7.7 Amendment # 156 Page 15 of 38 CPNPP NRC 2013 RO Written Exam Worksheet 65 to 75 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From Technical Specification Definitions 1.1 Amendment # 150 Page 16 of 38 CPNPP NRC 2013 RO Written Exam Worksheet 65 to 75 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 3 Category # 3 K/A # G 2.3.4 Importance Rating 3.2 Radiation Control: Knowledge of radiation exposure limits under normal or emergency condition Proposed Question: Common 70 Given the following condition:

• A 20 year old CPNPP employee has 1000 mrem of TEDE exposure for 2013.

Which of the following describes the MAXIMUM amount of additional TEDE exposure that may be received without additional CPNPP authorization (i.e., Plant Manager approval, Radiation and Industrial Safety Manager, employee supervisor, etc.), and what is the MAXIMUM amount of additional TEDE exposure that may be received prior to exceeding 10CFR20 (NRC) exposure limits?

A. 1000 mrem; 4000 mrem B. 3000 mrem; 4000 mrem C. 1000 mrem; 5000 mrem D. 3000 mrem; 5000 mrem Proposed Answer: A Explanation:

A. Correct. Admin limit is 2000 mrem per year; 10CFR20 limit is 5000 mrem per year.

B. Incorrect. Plausible because the Admin limit previously was 4000 mrem and a Planned Special Exposure is also 4000 mrem; 10CFR20 limit is 5000 mrem per year.

C. Incorrect. Plausible because the Admin limit is 2000 mrem, however, 5000 mrem is the legal limit.

Therefore, the person may only receive 4000 mrem additional.

D. Incorrect. Plausible because the Admin limit previously was 4000 mrem and a Planned Special Exposure is also 4000 mrem, however, 1000 mrem is the admin limit. Additionally, 5000 mrem is the legal limit and the person may only receive 2000 mrem additional.

Technical Reference(s) STA-655, Attachment 8.A Attached w/ Revision # See Comments / Reference Proposed references to be provided during examination: None Page 17 of 38 CPNPP NRC 2013 RO Written Exam Worksheet 65 to 75 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Learning Objective: INTERPRET and ENSURE compliance with plant administrative and operational procedures, guidelines, and policies.

Question Source: Bank # X 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 12 55.43 Comments /

Reference:

From STA-655, Attachment 8.A Revision # 20 Page 18 of 38 CPNPP NRC 2013 RO Written Exam Worksheet 65 to 75 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From STA-655, Attachment 8.A Revision # 20 Page 19 of 38 CPNPP NRC 2013 RO Written Exam Worksheet 65 to 75 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 3 Category # 3 K/A # G 2.3.12 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: Common 71 Given the following:

• Radiography is in progress in the Unit 2 Safeguards Building.

• When the source is exposed the dose rates in the area are rising to 1440 mrem/hr.

• The radiography is being performed in an open area of the Safeguards Building with NO enclosure.

Which of the following describes the type of radiological area and type of radiation monitoring required for entry?

A. Very High Radiation Area (VHRA)

Electronic Dosimeter and TLD B. Locked High Radiation Area (LHRA)

Electronic Dosimeter and TLD C. Very High Radiation Area (VHRA)

TLD only D. Locked High Radiation Area (LHRA)

TLD only Proposed Answer: B Explanation:

A. Incorrect. Plausible if thought that greater than 1000mr/hr is VHRA.

B. Correct. Area meets requirements for LHRA and Electronic dosimeter may be used for entry dose monitoring along with the TLD.

C. Incorrect. Plausible if thought that greater than 1000mr/hr is VHRA but area is LHRA due to dose rate and TLD only is not sufficient for monitoring.

D. Incorrect. Plausible LHRA is correct but TLD only is not sufficient for monitoring.

Technical Reference(s) STA-660 Attached w/ Revision # See Comments / Reference Proposed references to be provided during examination: None Page 20 of 38 CPNPP NRC 2013 RO Written Exam Worksheet 65 to 75 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Learning Objective: STATE the inverse square law.

CALCULATE the dose rate at varying distances from point sources, line sources, plane sources, and tank sources.

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 21 of 38 CPNPP NRC 2013 RO Written Exam Worksheet 65 to 75 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From STA-660 Revision 15 PCN 1 Page 22 of 38 CPNPP NRC 2013 RO Written Exam Worksheet 65 to 75 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From STA-660 Revision 15 PCN 1 Page 23 of 38 CPNPP NRC 2013 RO Written Exam Worksheet 65 to 75 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 3 Category # 3 K/A # G 2.3.13 Importance 3.4 Rating Radiation Control: Knowledge of radiological safety procedures pertaining to licensed operator duties, such as response to radiation monitor alarms, containment entry requirements, fuel handling responsibilities, access to locked high radiation areas, aligning filters, etc.

Proposed Question: Common 72 Given the following conditions:

• Maintenance Services has requested entry into the Incore Instrumentation Room, Elev. 832 to clean up debris around the Seal Table.

Which of the following identifies the condition that must be met prior to allowing access per STA-620, Containment Entry?

The Incore Detectors...

A. must be fully withdrawn from the core.

B. shall be fully inserted into the core.

C. shall be stored and tagged out-of-service.

D. Drive System must be disconnected.

Proposed Answer: C Explanation:

A. Incorrect. Plausible if thought that withdrawing incore detectors from the core places detectors in storage. However, when incore detectors are fully withdrawn, they are located in the incore instrumentation room. Additionally, the system must be tagged out to prevent inadvertent movement.

B. Incorrect. Plausible because the incore detectors may be fully inserted in the core, however, the incore detectors may also be inserted in their storage locations and the system is required to be tagged out to prevent inadvertent movement.

C. Correct. Per STA-620 the Incore Detectors System should be stored and tagged out of service to prevent possible movement and radiation overexposure of personnel.

D. Incorrect. Plausible if thought that disconnecting the drive system would prevent inadvertent movement of the Incore Detectors, however, it is the tagout that ultimately protects personnel.

Page 24 of 38 CPNPP NRC 2013 RO Written Exam Worksheet 65 to 75 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Technical Reference(s) STA-620, Step 6.1.2 Attached w/ Revision # See Comments / Reference Proposed references to be provided during examination: None Learning Objective: LIST the prerequisites that must be met prior to a Containment Entry.

Question Source: Bank # X 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 25 of 38 CPNPP NRC 2013 RO Written Exam Worksheet 65 to 75 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From STA-620, Step 6.1.2 Revision # 13 Page 26 of 38 CPNPP NRC 2013 RO Written Exam Worksheet 65 to 75 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 3 Category # 4 K/A # G 2.4.49 Importance Rating 4.6 Emergency Procedures / Plan: Ability to perform without reference to procedures those actions that require immediate operation of system components and controls.

Proposed Question: Common 73 Which of the following applies to the performance of Abnormal Conditions Procedure (ABN)

Initial Operator Actions?

ABN Initial Operator Actions SHALL be performed A. without verbalization and without a brief pause for SRO intervention.

B. with verbalization and without a brief pause for SRO intervention.

C. with verbalization and with a brief pause for SRO intervention.

D. without verbalization and with a brief pause for SRO intervention.

Proposed Answer: C Explanation:

A. Incorrect. Plausible because this is the expectation for ERG Immediate Action performance.

B. Incorrect. Plausible because actions shall be verbalized, however, the brief pause is required.

C. Correct. This is the expectation for ABN Initial Operator Action performance per Operations Guideline 3.

D. Incorrect. Plausible because verbalization is NOT required for ERG Immediate Action performance; however, the brief pause is required for ABN Initial Operator Action performance.

Technical Reference(s) Operations Guideline 3, Attachment 6 Attached w/ Revision # See A ODA-102, section 6.9 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 Page 27 of 38 CPNPP NRC 2013 RO Written Exam Worksheet 65 to 75 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 2 Comments /

Reference:

From Operations Guideline 3, Attachment 6 Revision 110 Page 28 of 38 CPNPP NRC 2013 RO Written Exam Worksheet 65 to 75 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ODA-102, section 6.9 Revision 26 Page 29 of 38 CPNPP NRC 2013 RO Written Exam Worksheet 65 to 75 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 3 Category # 4 K/A # G 2.4.31 Importance Rating 4.2 Emergency Procedures / Plan: Knowledge of annunciator alarms, indications or response procedures Proposed Question: Common 74 Given the following conditions:

• Unit 1 is responding to a Steam Generator Tube Rupture and is currently in EOP-0.0A, Reactor Trip or Safety Injection.

Which of the following is correct concerning Transient Annunciator Response?

A. The master silence button may be used repeatedly at the discretion of the Reactor Operator after verbalizing Silencing.

B. The Reactor Operator shall inform the Unit Supervisor of ALL Orange or Yellow annunciators in alarm.

C. The Reactor Operator shall inform the Unit Supervisor of any alarms that indicate failure of an ESF component.

D. Annunciators which cannot be expeditiously addressed should continue to Flash until they can be addressed.

Proposed Answer: C Explanation:

A. Incorrect. Plausible as the master silence is to be used to allow clear communications, however, repeated use of the master silence requires Unit Supervisor approval.

B. Incorrect. Plausible as all Orange and Yellow annunciators show a level of importance beyond the standard white annunciator, however, the operating guidance is that only those annunciators which indicate that another ERG or ABN should be entered or that indicate ESF component failures or ESF actuations that may have occurred or need to be manually performed should be verbalized reported to the Unit Supervisor.

C. Correct. The failure of an ESF component should be reported to the Unit Supervisor within normal communication standards.

D. Incorrect. Plausible as not all annunciators can be expeditiously addressed, however, the guidance is that the annunciators should be acknowledged and cleared when conditions stabilize in order to identify those that may indicate a subsequent fault or transient.

Page 30 of 38 CPNPP NRC 2013 RO Written Exam Worksheet 65 to 75 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Technical Reference(s) A Operations Guideline 3, Attachment 6 Attached w/ Revision # See Comments / Reference Proposed references to be provided during examination: None Learning Objective: STATE the performance and design attributes of the following Reactor Vessel and Internals System components, flowpaths, and features:

• Reactor Vessel Head

• Penetrations and O Rings 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 31 of 38 CPNPP NRC 2013 RO Written Exam Worksheet 65 to 75 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From Operations Guideline 3, Attachment 6 Revision 110 Page 32 of 38 CPNPP NRC 2013 RO Written Exam Worksheet 65 to 75 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 3 Category # 4 K/A # G 2.4.46 Importance Rating 4.2 Emergency Procedures/Plan: Ability to verify that the alarms are consistent with the plant conditions Proposed Question: Common 75 Given the following conditions:

• Unit 1 is operating at 100% power.

• ABN-302, Feedwater, Condensate, Heater Drain System Malfunction, is being performed due to a trip of Main Feedwater Pump 1-01.

Which of the following alarms is NOT CONSISTENT with given plant conditions?

A. 1-ALB-8A, Window 2.12 - SG 2 LEV DEV B. 1-ALB-6D, Window 1.10 - AVG TAVG - TREF DEV C. 1-PCIP Window 3.4 - TURB LOAD REJ STM DMP ARMED C-7 D. 1-ALB-6D, Window 4.14 - CONTROL ROD BANK D FULL WTHDRWL Proposed Answer: D Explanation:

A. Incorrect. Plausible if believed that the automatic runback coupled with feedwater control valve response will maintain steam generator water level within 5% of program. However, the large feedwater flow transient caused by the trip of a Main Feedwater Pump will cause steam generator level deviations of sufficient magnitude to cause level deviation alarms.

B. Incorrect. Plausible if believed that automatic rod control will maintain T AVE within 2.5°F of T REF .

However, the rate of turbine power decrease due to the automatic runback (35%/minute) will cause T AVE -T REF deviation to go above the alarm setpoint.

C. Incorrect. Plausible if believed that the automatic runback following a Main Feed Pump trip occurs at the normal runback rate of 10 MW/minute. However, the automatic runback occurs at 35%/minute, which is greater than the 10% power per 120 second arming setpoint for C-7.

D. Correct. When a Main Feedwater Pump trips, a turbine runback is automatically initiated. This will cause control rods to step in due to a mismatch in turbine and reactor power. Control Bank D will not move out to 223 steps and 1-ALB-6D, Window 4.14 should not alarm.

Page 33 of 38 CPNPP NRC 2013 RO Written Exam Worksheet 65 to 75 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Technical Reference(s) ABN-302, Sections 2.1 and 2.2 Attached w/ Revision # See ALM-0064A, 1-ALB-6D, Window 4.14 Comments / Reference ALM-0064A, 1-ALB-06D, Window 1.10 ALM-0065A, PCIP, Window 3.4 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 # X 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 38 CPNPP NRC 2013 RO Written Exam Worksheet 65 to 75 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ABN-302, Sections 2.1 and 2.2 Revision 14 Page 35 of 38 CPNPP NRC 2013 RO Written Exam Worksheet 65 to 75 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ALM-0064A, 1-ALB-06D, Window 4.14 Revision 6 Page 36 of 38 CPNPP NRC 2013 RO Written Exam Worksheet 65 to 75 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ALM-0064A, 1-ALB-06D, Window 1.10 Revision 6 Page 37 of 38 CPNPP NRC 2013 RO Written Exam Worksheet 65 to 75 Rev h

ES-401 CPNPP NRC 2013 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ALM-0065A, PCIP, Window 3.4 Revision 4 Page 38 of 38 CPNPP NRC 2013 RO Written Exam Worksheet 65 to 75 Rev h

ES-401 CPNPP NRC 2013 SRO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 011 G 2.2.4 Importance Rating 3.6 Large Break LOCA: Emergency Procedures/Plan: (multi-unit) Ability to explain the variations in control board layouts, systems, instrumentation and procedural actions between units at a facility Proposed Question: SRO 76 Given the following conditions:

• Unit 2 is responding to Large Break Loss of Coolant Accident (LOCA).

• Containment pressure is 30 psig and lowering.

• Reactor Coolant System pressure is 30 psig and lowering.

• Wide range T HOT indications in three loops are 280°F and lowering.

• The other wide range T HOT indication is 210°F and lowering.

• Refueling Water Storage Tank level is 63% and lowering.

• All safety systems functioned as designed.

• While checking intact Steam Generators (SG) the following levels are observed:

• SG 2-01 is 45% and stable.

• SG 2-02 is 8% and stable.

• SG 2-03 is 15% and stable.

• SG 2-04 is 35% and stable.

Which of the following actions is required to ensure the Steam Generator level control band is maintained per EOP-1.0B, Loss of Reactor or Secondary Coolant?

Establish a level band in...

A. ...Steam Generator 2-02 of 10% to 50%.

B. ...Steam Generators 2-02 & 2-03 of 18% to 50%.

C. ...Steam Generators 2-02, 2-03 & 2-04 of 43% to 60%.

D. ...Steam Generators 2-01, 2-02, 2-03 & 2-04 of 50% to 60%.

Proposed Answer: B Page 1 of 60 CPNPP NRC 2013 SRO Written Exam Worksheet 76 to 88 Rev h

ES-401 CPNPP NRC 2013 SRO Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible because the control band on Unit 2 without Adverse Containment is 10% to 50%.

B. Correct. The control band on Unit 2 with Adverse Containment is 18% to 50%.

C. Incorrect. Plausible because the control band without Adverse Containment on Unit 1 is 43% to 60%.

D. Incorrect. Plausible because the control band on Unit 1 with Adverse Containment is 50% to 60%.

Technical Reference(s) EOP-1.0A, Step 3 Attached w/ Revision # See EOP-1.0B, Step 3 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 EOP-1.0, Loss of Reactor or Secondary Coolant.

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 2 of 60 CPNPP NRC 2013 SRO Written Exam Worksheet 76 to 88 Rev h

ES-401 CPNPP NRC 2013 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From EOP-1.0A, Step 3 Revision # 8 Page 3 of 60 CPNPP NRC 2013 SRO Written Exam Worksheet 76 to 88 Rev h

ES-401 CPNPP NRC 2013 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From EOP-1.0B, Step 3 Revision # 8 Page 4 of 60 CPNPP NRC 2013 SRO Written Exam Worksheet 76 to 88 Rev h

ES-401 CPNPP NRC 2013 SRO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 022 AA2.02 Importance Rating 3.7 Loss of Reactor Coolant Makeup: Ability to determine and interpret the following as they apply to the Loss of Reactor Coolant Makeup: Charging pump problems Proposed Question: SRO 77 Given the following conditions:

• Unit 1 is at 100% power.

• Centrifugal Charging Pump (CCP) 1-01 is in service.

• Volume Control Tank (VCT) level is 50%.

• 1-FI-121A, CHRG FLO is stable at 130 gpm.

• 1-FI-132, LTDN FLO is stable at 120 gpm.

• 1/1-LCV-112B, VCT TO CHRG PMP SUCT VLV spuriously closes.

• The following alarms are received;

• 1-ALB-6A, Window 1.4, REGEN HX LTDN OUT TEMP HI

• 1-ALB-6A, Window 3.4, CHRG FLO HI/LO What action must be taken and the procedure used?

A. Stop Centrifugal Charging Pump 1-01 and then isolate letdown per SOP-103A, Chemical and Volume Control System.

B. Stop Centrifugal Charging Pump 1-01 and then isolate letdown per ABN-105, Chemical and Volume Control System Malfunction.

C. Isolate letdown and then stop Centrifugal Charging Pump 1-01 per SOP-103A, Chemical and Volume Control System.

D. Isolate letdown and then stop Centrifugal Charging Pump 1-01 per ABN-105, Chemical and Volume Control System Malfunctions.

Proposed Answer: B Page 5 of 60 CPNPP NRC 2013 SRO Written Exam Worksheet 76 to 88 Rev h

ES-401 CPNPP NRC 2013 SRO Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible because the actions stated are correct, however, ABN-105 is the correct procedure to use.

B. Correct. This is a symptom of pump cavitation due to suction valve being closed. The RNO actions of ABN-105 direct the operator to stop the CCP and isolate Letdown.

C. Incorrect. Plausible because actions stated are correct, however, they must be performed in the opposite order in accordance with ABN-105.

D. Incorrect. Plausible because actions stated are correct, however, they must be performed in the opposite order.

Technical Reference(s) ABN-105, Step 7.3.1 RNO Attached w/ Revision # See 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 Chemical and Volume Control 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 55.43 5 Page 6 of 60 CPNPP NRC 2013 SRO Written Exam Worksheet 76 to 88 Rev h

ES-401 CPNPP NRC 2013 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ABN-105, Step 7.3.1 RNO Revision # 7 Page 7 of 60 CPNPP NRC 2013 SRO Written Exam Worksheet 76 to 88 Rev h

ES-401 CPNPP NRC 2013 SRO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 029 G 2.4.6 Importance Rating 4.7 ATWS: Knowledge of EOP mitigation strategies.

Proposed Question: SRO 78 Given the following conditions:

• An ATWT has occurred on Unit 1.

• FRS-0.1A, Response to Nuclear Power Generation/ATWT is in progress.

• Emergency Boration is in progress.

• Safety Injection has actuated.

• All Steam Generator pressures are 800 psig and lowering.

• All Reactor Coolant System Cold Leg temperatures are 521°F and lowering.

• Reactor Power is 9% and lowering.

Which of the following mitigation strategies is appropriate for the event?

A. Remain in FRS-0.1A, Response to Nuclear Power Generation/ATWT and isolate faulted steam generators. Transition to EOP-0.0A, Reactor Trip or Safety Injection when FRS-0.1A, Response to Nuclear Power Generation/ATWT is complete.

B. Remain in FRS-0.1A, Response to Nuclear Power Generation/ATWT and isolate faulted steam generators. Transition to EOP-0.0A, Reactor Trip or Safety Injection when faulted SG isolation is complete.

C. Transition to EOP-2.0A, Faulted Steam Generator Isolation while performing FRS-0.1A, Response to Nuclear Power Generation/ATWT in parallel. Transition to EOP-0.0A, Reactor Trip or Safety Injection when faulted SG isolation is complete.

D. Transition to EOP-2.0A, Faulted Steam Generator Isolation while performing FRS-0.1A, Response to Nuclear Power Generation/ATWT in parallel. Transition to EOP-0.0A, Reactor Trip or Safety Injection when FRS-0.1A, Response to Nuclear Power Generation/ATWT is complete.

Proposed Answer: A Page 8 of 60 CPNPP NRC 2013 SRO Written Exam Worksheet 76 to 88 Rev h

ES-401 CPNPP NRC 2013 SRO Written Exam Worksheet Form ES-401-5 Explanation:

A. Correct. The faulted SG will be isolated in FRS-0.1A. Transition to EOP-0.0A will be made when FRS-0.1A is complete.

B. Incorrect. Plausible because remaining in FRS-0.1A is correct but transition to EOP-0.0A after fault isolation is not correct. FRS-0.1A must be completed to transition to EOP-0.0A.

C. Incorrect. Plausible because EOP-2.0A can be performed in parallel with other ORGs, however FRS-0.1A must remain as the procedure in affect. FRS-0.1A must be completed to transition to EOP-0.0A.

D. Incorrect. Plausible because EOP-2.0A can be performed in parallel with other ORGs, however FRS-0.1A must remain as the procedure in affect. Transition to EOP-0.0A will be made when FRS-0.1A is complete.

Technical Reference(s) ODA-407, Attachment 8.A Attached w/ Revision # See FRS-0.1A, step 15 Comments / Reference FRS-0.1A, Step 6 EOP-2.0A, Symptoms and Entry Conditions Proposed references to be provided during examination: None Learning Objective: Given a procedure step, note, or caution discuss the reason for the step, note, or caution in FRS-0.1 Question Source: Bank #

Modified Bank # X (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 5 Page 9 of 60 CPNPP NRC 2013 SRO Written Exam Worksheet 76 to 88 Rev h

ES-401 CPNPP NRC 2013 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ODA-407, Attachment 8.A Revision 8 Comments /

Reference:

From FRS-0.1A, step 15 Revision 8 Page 10 of 60 CPNPP NRC 2013 SRO Written Exam Worksheet 76 to 88 Rev h

ES-401 CPNPP NRC 2013 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From FRS-0.1A, Step 6 Revision # 8 Page 11 of 60 CPNPP NRC 2013 SRO Written Exam Worksheet 76 to 88 Rev h

ES-401 CPNPP NRC 2013 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From EOP-2.0A, Symptoms and Entry Conditions Revision 8 Page 12 of 60 CPNPP NRC 2013 SRO Written Exam Worksheet 76 to 88 Rev h

ES-401 CPNPP NRC 2013 SRO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 026 AA2.04 Importance Rating 2.9 Loss of Component Cooling Water: Ability to determine and interpret the following as they apply to the Loss of Component Cooling Water: The normal values and upper limits for the temperatures of the components cooled by CCW Proposed Question: SRO 79 Given the following conditions on Unit 1:

• High temperature alarms are received on all 4 Reactor Coolant Pumps (RCP).

• Thermal Barrier Return temperatures indicate 185ºF and rising at 5ºF per minute.

• Lower Seal Bearing temperature indicates 200ºF and rising at 5ºF per minute.

Which of the following describes the status of RCP thermal barrier cooling, and the action required for this condition?

A. 1-HV-4696, THBR CLR CCW RET ISOL VLV (IRC), is closed.

RCP temperatures exceed the operating limits and must be immediately tripped per ABN-502, Component Cooling Water System Malfunctions.

B. 1-HV-4709, THBR CLR CCW RET ISOL VLV (ORC), is closed.

RCPs must be tripped within 5 minutes due to high temperature per ABN-101, Reactor Coolant Pump Trip/Malfunction.

C. 1-HV-4709, THBR CLR CCW RET ISOL VLV (ORC), is closed.

RCP temperatures exceed the operating limits and must be immediately tripped per ABN-502, Component Cooling Water System Malfunctions.

D. 1-HV-4696, THBR CLR CCW RET ISOL VLV (IRC), is closed.

RCPs must be tripped within 5 minutes due to high temperature per ABN-101, Reactor Coolant Pump Trip/Malfunction.

Proposed Answer: B Page 13 of 60 CPNPP NRC 2013 SRO Written Exam Worksheet 76 to 88 Rev h

ES-401 CPNPP NRC 2013 SRO Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible because closure of 1-HV-4696 could result in a rise in temperatures as shown, however, this condition is addressed by ABN-101 and RCP temperature limits have not yet been exceeded.

B. Correct. 1-HV-4709 closes when temperatures exceed 182.5°F. Per ABN-101, RCPs must be tripped when lower seal bearing temperature reaches 225°F and this will occur within 5 min. per the conditions in the Stem.

C. Incorrect. Plausible because 1-HV-4709 is closed, however, RCP temperatures have not yet exceeded the limits and the procedure reference is incorrect.

D. Incorrect. Plausible because the RCPs must be tripped within 5 min. and the procedure reference is correct, however, 1-HV-4696 closes when flow is greater than or equal to 64 gpm and this condition has not been identified.

Technical Reference(s) ABN-101, Section 8.1 & 8.2 Attached w/ Revision # See ABN-101, Step 8.3.2 Comments / Reference ABN-101, Attachment 1 Proposed references to be provided during examination: None Learning Objective: ANALYZE the response to an RCP High Temperature or Loss of CCW to any RCP in accordance with ABN-101, Reactor Coolant Pump Trip/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 55.43 5 Page 14 of 60 CPNPP NRC 2013 SRO Written Exam Worksheet 76 to 88 Rev h

ES-401 CPNPP NRC 2013 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ABN-101, Section 8.1 & 8.2 Revision # 10 Page 15 of 60 CPNPP NRC 2013 SRO Written Exam Worksheet 76 to 88 Rev h

ES-401 CPNPP NRC 2013 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ABN-101, Step 8.3.2 Revision # 10 Page 16 of 60 CPNPP NRC 2013 SRO Written Exam Worksheet 76 to 88 Rev h

ES-401 CPNPP NRC 2013 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ABN-101, Attachment 1 Revision # 10 Page 17 of 60 CPNPP NRC 2013 SRO Written Exam Worksheet 76 to 88 Rev h

ES-401 CPNPP NRC 2013 SRO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 040 G 2.4.41 Importance Rating 4.6 Steam Line Rupture: Emergency Procedures/Plan: Knowledge of the emergency action level thresholds and classifications Proposed Question: SRO 80 Given the following conditions:

• Unit 2 has experienced a steam line fault in the Main Steam Line.

• Security has reported significant damage to the Safeguards Building.

• Two of the four Main Steam Line Isolation Valves have failed to close.

• Personnel cannot enter the Main Steam Header room due to steam and building damage.

Which of the following Emergency Action Level category and classification applies?

A. Hazards, Natural or Destructive Phenomena - Unusual Event B. Hazards, Fire or Explosion - Unusual Event C. Hazards, Natural or Destructive Phenomena - Alert D. Hazards, Fire or Explosion - Alert Proposed Answer: D Explanation:

A. Incorrect. Plausible if it believed that a Main Steam Line Break is classified as a Destructive Phenomena. However, Note 9 in the EAL chart directs that a steam line break be classified under the Fire or Explosion EAL. Additionally, this fault would be classified as an Alert due to visible damage to the safeguards building.

B. Incorrect. Plausible because Note 9 in the EAL chart directs that a steam line break be classified under the Fire or Explosion EAL. However, this fault would be classified as an Alert due to visible damage to the safeguards building.

C. Incorrect. Plausible because this fault would be classified as an Alert due to visible damage to the safeguards building. However, Note 9 in the EAL chart directs that a steam line break be classified under the Fire or Explosion EAL.

D. Correct. Explosion is defined as a rapid, violent, unconfined combustion, or catastrophic failure of pressurized/energized equipment that imparts energy of sufficient force to potentially damage permanent structures. Additionally, this fault would be classified as an Alert due to visible damage to the safeguards building.

Page 18 of 60 CPNPP NRC 2013 SRO Written Exam Worksheet 76 to 88 Rev h

ES-401 CPNPP NRC 2013 SRO Written Exam Worksheet Form ES-401-5 Technical Reference(s) EPP-201, Common & Hot EALs Table Attached w/ Revision # See EPP-201, EAL Technical Bases Comments / Reference Proposed references to be provided during examination: Emergency Action Level Charts Emergency Action Level Technical Bases 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 Comments /

Reference:

From EPP-201, Common EALs Table Revision # 12 Comments /

Reference:

From EPP-201, Common EALs Table Revision # 12 Page 19 of 60 CPNPP NRC 2013 SRO Written Exam Worksheet 76 to 88 Rev h

ES-401 CPNPP NRC 2013 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From EPP-201, Common EALs Table Revision # 12 Page 20 of 60 CPNPP NRC 2013 SRO Written Exam Worksheet 76 to 88 Rev h

ES-401 CPNPP NRC 2013 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From EPP-201, EAL Technical Bases Revision # 12 Category: H - Hazards Subcategory: 2 - Fire or Explosion Initiating Condition: Fire or explosion affecting the operability of plant safety systems required to establish or maintain safe shutdown EAL:

HA2.1 Alert Fire or explosion resulting in EITHER:

• Visible damage to any Table H-1 structures

• Control Room indication of degraded performance of systems required to establish or maintain safe shutdown Mode Applicability:

All Basis:

Generic VISIBLE DAMAGE is used to identify the magnitude of the FIRE or EXPLOSION and to discriminate against minor FIRES and EXPLOSIONS.

The reference to structures containing safety systems or components is included to discriminate against FIRES or EXPLOSIONS in areas having a low probability of affecting safe operation. The significance here is not that a safety system was degraded but the fact that the FIRE or EXPLOSION was large enough to cause damage to these systems.

The use of VISIBLE DAMAGE should not be interpreted as mandating a lengthy damage assessment prior to classification. The declaration of an Alert and the activation of the Technical Support Center will provide the Emergency Coordinator with the resources needed to perform detailed damage assessments.

The Emergency Coordinator also needs to consider any security aspects of the EXPLOSION.

Page 21 of 60 CPNPP NRC 2013 SRO Written Exam Worksheet 76 to 88 Rev h

ES-401 CPNPP NRC 2013 SRO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # W/E12 EA2.2 Importance Rating 3.9 Uncontrolled Depressurization of All Steam Generators: Ability to determine and interpret the following as they apply to the Uncontrolled Depressurization of All Steam Generators: Adherence to appropriate procedures and operation within the limitations in the facility's license and amendments Proposed Question: SRO 81 Given the following conditions with the Unit in MODE 1:

• The Reactor tripped and Safety Injection occurred due to low Pressurizer pressure.

• All Steam Generator pressures were decreasing in an uncontrolled manner.

• Main Steam Isolation Valves would NOT close from the Control Room.

• After entering ECA-2.1A, Uncontrolled Depressurization of All Steam Generators, local operator actions restored the pressure boundary for all Steam Generators.

• Steam Generator pressures are stabilizing but an increase in pressure has not been seen in any Steam Generator.

• After Steam Generator 1-01 pressure stabilizes, level continues to increase in an uncontrolled manner.

Which of the following procedures should be implemented?

A. Transition to ECA-3.1A, SGTR with Loss of Reactor Coolant - Subcooled Recovery Desired.

B. Continue with ECA-2.1A, Uncontrolled Depressurization of All Steam Generators.

C. Transition to EOP-3.0A, Steam Generator Tube Rupture.

D. Transition to EOP-1.0A, Loss of Reactor or Secondary Coolant.

Proposed Answer: C Page 22 of 60 CPNPP NRC 2013 SRO Written Exam Worksheet 76 to 88 Rev h

ES-401 CPNPP NRC 2013 SRO Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible if believed that the ruptured steam generator is not isolated due to steam generator pressure not rising. Additionally, transition to ECA-3.1A would be correct following actions in EOP-3.0A if the steam generator remained faulted. However, the fact that steam generator pressures are stabilizing indicates steam generator isolation. Additionally, ECA-3.1A cannot be entered directly from ECA-2.1A.

B. Incorrect. Plausible because this is the procedure in effect, however, the ECA-2.1A Foldout Page requires a transition to EOP-3.0A on any Steam Generator with an increasing level.

C. Correct. ECA-2.1A Foldout Page requires a transition to EOP-3.0A on any Steam Generator with an increasing level.

D. Incorrect. Plausible because with the Pressurizer empty it could be thought that a transition to EOP-1.0A is appropriate due to SI Reinitiation Criteria, however, EOP-3.0A entry is required.

Technical Reference(s) ECA-2.1A, Attachment 1A, Foldout Page Attached w/ Revision # See ECA-2.1A, Step 5 RNO Comments / Reference ECA-2.1A, Flowchart FRI-0.2A, CSFST Proposed references to be provided during examination: None Learning Objective: IDENTIFY the proper transitions out of ECA-2.1, Uncontrolled Depressurization of All Steam Generators.

Question Source: Bank # X 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 23 of 60 CPNPP NRC 2013 SRO Written Exam Worksheet 76 to 88 Rev h

ES-401 CPNPP NRC 2013 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ECA-2.1A, Attachment 1A, Foldout Page Revision # 8 Page 24 of 60 CPNPP NRC 2013 SRO Written Exam Worksheet 76 to 88 Rev h

ES-401 CPNPP NRC 2013 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ECA-2.1A, Step 5 RNO Revision # 8 Page 25 of 60 CPNPP NRC 2013 SRO Written Exam Worksheet 76 to 88 Rev h

ES-401 CPNPP NRC 2013 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ECA-2.1A, Flowchart Revision # 8 Page 26 of 60 CPNPP NRC 2013 SRO Written Exam Worksheet 76 to 88 Rev h

ES-401 CPNPP NRC 2013 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From FRI-0.2A, CSFST Revision # 8 Page 27 of 60 CPNPP NRC 2013 SRO Written Exam Worksheet 76 to 88 Rev h

ES-401 CPNPP NRC 2013 SRO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 2 K/A # 037 AA2.02 Importance Rating 3.9 Steam Generator Tube Leak: Ability to determine and interpret the following as they apply to the Steam Generator Tube Leak: Agreement/disagreement among redundant radiation monitors Proposed Question: SRO 82 Given the following conditions:

• Unit 1 is at 100% power.

• Chemistry reports an abnormal increase in Condenser Off-Gas specific activity.

• Chemistry suspects that a Steam Generator tube leak exists.

• Chemistry calculates the leak rate is 10 gpd on Steam Generator 1-02.

Which of the following lists the two independent radiation monitors that should be used to confirm the Steam Generator tube leakage?

A. COG-182 Condenser Off-Gas Radiation Monitor, and N16-175 # 2 Main Steam Line N16 Radiation Monitor B. COG-182 Condenser Off-Gas Radiation Monitor, and MSL-179 # 2 Main Steam Line Radiation Monitor C. SGS-164 Steam Generator Blowdown Sample Radiation Monitor, and N16-175 # 2 Main Steam Line N16 Radiation Monitor D. SGS-164 Steam Generator Blowdown Sample Radiation Monitor, and MSL-179 # 2 Main Steam Line Radiation Monitor Proposed Answer: A Page 28 of 60 CPNPP NRC 2013 SRO Written Exam Worksheet 76 to 88 Rev h

ES-401 CPNPP NRC 2013 SRO Written Exam Worksheet Form ES-401-5 Explanation:

A. Correct. Per the conditions in the Stem and the guidance provided in ABN-106, these are the correct Radiation Monitors to confirm the Steam Generator Tube Leakage.

B. Incorrect. Plausible because COG-182, Condenser Off-Gas Radiation Monitor is correct, however, MSL-179, #2 Main Steam Line Radiation Monitor does not meet the minimum sensitivity setpoint required per ABN-106.

C. Incorrect. Plausible because the N16 Main Steam Line Radiation Monitor is correct and Steam Generator Blowdown will isolate on high radiation, however, this is a symptom associated with leakage greater than or equal to 75 gpd.

D. Incorrect. Plausible because both of these instruments can be used, however, MSL-179, #2 Main Steam Line Radiation Monitor does not meet the minimum sensitivity setpoint for the leak rate involved (10 gpd) and neither does the Steam Generator Blowdown Radiation Monitor.

Technical Reference(s) ABN-106, Section 2.0 Attached w/ Revision # See ABN-106, Step 2.3.1 & 2.3.2 NOTES Comments / Reference ABN-106, Step 3.3.1 NOTE ABN-106, Section 3.0 Proposed references to be provided during examination: None Learning Objective: ANALYZE the response to a Steam Generator Tube Leakage less than 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 Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 5 Page 29 of 60 CPNPP NRC 2013 SRO Written Exam Worksheet 76 to 88 Rev h

ES-401 CPNPP NRC 2013 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ABN-106, Section 2.0 Revision # 10 Page 30 of 60 CPNPP NRC 2013 SRO Written Exam Worksheet 76 to 88 Rev h

ES-401 CPNPP NRC 2013 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ABN-106, Step 2.3.1 & 2.3.2 NOTES Revision # 10 Page 31 of 60 CPNPP NRC 2013 SRO Written Exam Worksheet 76 to 88 Rev h

ES-401 CPNPP NRC 2013 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ABN-106, Step 3.3.1 NOTE Revision # 10 Page 32 of 60 CPNPP NRC 2013 SRO Written Exam Worksheet 76 to 88 Rev h

ES-401 CPNPP NRC 2013 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ABN-106, Section 3.0 Revision # 10 Page 33 of 60 CPNPP NRC 2013 SRO Written Exam Worksheet 76 to 88 Rev h

ES-401 CPNPP NRC 2013 SRO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 2 K/A # 051 G 2.2.44 Importance Rating 4.4 Loss of Condenser Vacuum: Equipment Control: Ability to interpret control room indications to verify the status and operation of the system, and understand how operator actions and directives affect plant and system conditions Proposed Question: SRO 83 Given the following conditions:

• Unit 1 is at 9% power.

• The Main Turbine is latched at 1800 RPM in preparation for synchronization.

• Main Condenser Vacuum is 21 and lowering.

• All Condenser Vacuum Pumps are running.

• Operators have been dispatched to determine the cause for the loss of vacuum.

Which of the following should be performed per ABN-304, Main Condenser and Circulating Water System Malfunction?

Trip the A. Reactor, enter EOP-0.0A, Reactor Trip or Safety Injection, and continue actions of ABN-304, Main Condenser and Circulating Water System Malfunction.

B. Main Turbine, enter ABN-403, Turbine Trip Response, and continue actions of ABN-304, Main Condenser and Circulating Water System Malfunction.

C. Reactor, enter EOP-0.0A, Reactor Trip or Safety Injection, and exit ABN-304, Main Condenser and Circulating Water System Malfunction.

D. Main Turbine, enter ABN-403, Turbine Trip Response, and exit ABN-304, Main Condenser and Circulating Water System Malfunction.

Proposed Answer: B Page 34 of 60 CPNPP NRC 2013 SRO Written Exam Worksheet 76 to 88 Rev h

ES-401 CPNPP NRC 2013 SRO Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible because this answer is correct if Reactor power was greater than or equal to 10%, however, at this power level only a Main Turbine trip is required.

B. Correct. As outlined in ABN-304, Step 3.3.3 RNO.

C. Incorrect. Plausible because the Reactor would be tripped if Reactor power was greater than or equal to 10%, however, the loss of vacuum must still be addressed because other plant systems such as Main Feedwater Pumps and Steam Dump System would be affected.

D. Incorrect. Plausible because a Main Turbine trip is required and ABN-403 would be entered, however, the loss of vacuum must still be addressed because other plant systems such as Main Feedwater Pumps would be affected.

Technical Reference(s) ABN-304, Section 3.2 Attached w/ Revision # See ABN-304, Step 3.3.3 RNO Comments / Reference Proposed references to be provided during examination: None Learning Objective: ANALYZE the response to a lowering Main or Auxiliary Condenser Vacuum per ABN-304, Main Condenser and Circulating 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 55.43 5 Comments /

Reference:

From ABN-304, Section 3.2 Revision # 8 Page 35 of 60 CPNPP NRC 2013 SRO Written Exam Worksheet 76 to 88 Rev h

ES-401 CPNPP NRC 2013 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ABN-304, Step 3.3.3 RNO Revision # 8 Page 36 of 60 CPNPP NRC 2013 SRO Written Exam Worksheet 76 to 88 Rev h

ES-401 CPNPP NRC 2013 SRO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 2 K/A # W/E03 EA2.1 Importance Rating 4.2 LOCA Cooldown - Depressurization: Ability to determine and interpret the following as they apply to the LOCA Cooldown and Depressurization: Facility conditions and selection of appropriate procedures during abnormal and emergency operations Proposed Question: SRO 84 Given the following conditions:

• A Steam Line Break has occurred.

• Normal Charging is being established per EOS-1.1A, Safety Injection Termination.

• The faulted Steam Generator has stopped depressurizing.

• While attempting to control Pressurizer level with normal Charging, Pressurizer level continues to lower.

Which of the following describes the action required?

A. Re-actuate Safety Injection and transition to EOP-1.0A, Loss of Reactor or Secondary Coolant.

B. Re-actuate Safety Injection and transition to EOP-0.0A, Reactor Trip or Safety Injection.

C. Realign the Centrifugal Charging Pump Injection flowpath and transition to EOS-1.2A, Post LOCA Cooldown and Depressurization.

D. Realign the Centrifugal Charging Pump Injection flowpath and remain in EOS-1.1A, Safety Injection Termination.

Proposed Answer: C Explanation:

A. Incorrect. Plausible because once SI is terminated the Foldout Page requires transition to EOP-1.0 after starting ECCS Pumps.

B. Incorrect. Plausible because this would be correct if events occurred in EOS-0.1A, Reactor Trip Response. EOS-1.1A requires manual operation and transition to correct procedure for LOCA.

C. Correct. Per EOS-1.1A, Step 11 RNO the Safety Injection Pumps are not stopped, Letdown is not established, and crew is either in the wrong procedure or another event has occurred. Equipment alignment at this point would require transition to EOS-1.2A for more appropriate recovery.

D. Incorrect. Plausible because the Charging Pump injection flowpath should be realigned, however, the RNO action of Step 11 requires entry into EOS-1.2A, Post LOCA Cooldown and Depressurization.

Page 37 of 60 CPNPP NRC 2013 SRO Written Exam Worksheet 76 to 88 Rev h

ES-401 CPNPP NRC 2013 SRO Written Exam Worksheet Form ES-401-5 Technical Reference(s) EOS-1.1A, Step 11 RNO Attached w/ Revision # See EOS-1.1A, Steps 2 & 15 Comments / Reference Proposed references to be provided during examination: None Learning Objective: Given a set of plant conditions, IDENTIFY the proper transitions through/out of EOS-1.1, Safety Injection Termination.

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

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

Reference:

From EOS-1.1A, Step 11 RNO Revision # 8 Page 38 of 60 CPNPP NRC 2013 SRO Written Exam Worksheet 76 to 88 Rev h

ES-401 CPNPP NRC 2013 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From EOS-1.1A, Step 2 Revision # 8 Comments /

Reference:

From EOS-1.1A, Step 15 Revision # 8 Page 39 of 60 CPNPP NRC 2013 SRO Written Exam Worksheet 76 to 88 Rev h

ES-401 CPNPP NRC 2013 SRO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 2 K/A # W/E14 G 2.4.49 Importance Rating 4.4 High Containment Pressure: Emergency Procedures/Plan: Ability to perform without reference to procedures those actions that require immediate operation of system components and controls Proposed Question: SRO 85 Given the following conditions:

• Unit 2 entered EOP-0.0B, Reactor Trip or Safety Injection due to a Reactor Trip and Safety Injection with Containment Spray actuation.

• The actions of EOP-0.0B have been completed and a diagnosis has been made that a loss of Reactor Coolant inside Containment exists.

• At the time of entry into EOP-1.0B, Loss of Reactor or Secondary Coolant, an ORANGE Path condition is noted on the Containment Status tree when 2-ALB-4B Window 1.8 - RWST 2 OF 4 LVL LO-LO annunciates and level is confirmed to be at 33%.

Which of the following is the proper course of actions to implement?

A. Transition to FRZ-0.1B, Response to High Containment Pressure and complete the actions without delay so that EOS-1.3B, Transfer to Cold Leg Recirculation can be implemented to realign Emergency Core Cooling System injection.

B. Complete Steps 1, 2 and 3 of EOS-1.3B, Transfer to Cold Leg Recirculation and EOP-1.0B, Loss of Reactor or Secondary Coolant. Performance of FRZ-0.1B, Response to High Containment Pressure is not required as EOP-0.0B verified conditions for Containment Spray.

C. Transition to FRZ-0.1B, Response to High Containment Pressure and concurrently implement actions of EOS-1.3B, Transfer to Cold Leg Recirculation. When completed, return to EOP-1.0B, Loss of Reactor or Secondary Coolant.

D. Complete Steps 1, 2 and 3 of EOS-1.3B, Transfer to Cold Leg Recirculation then review Critical Safety Function Status Trees. If ORANGE condition on Containment still exists, then transition to FRZ-0.1B, Response to High Containment Pressure.

Proposed Answer: D Page 40 of 60 CPNPP NRC 2013 SRO Written Exam Worksheet 76 to 88 Rev h

ES-401 CPNPP NRC 2013 SRO Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible because the Critical Safety Function ORANGE Path is normally immediately addressed, however, entry and completion of EOS-1.3B takes priority.

B. Incorrect. Plausible because completing the actions of EOS-1.3B is correct, however, the Critical Safety Function ORANGE Path cannot be ignored.

C. Incorrect. Plausible because the last two actions listed are correct, however, EOP-1.0B is not implemented at this time.

D. Correct. Given the conditions listed, Steps 1 through 3 of EOS-1.3B must be performed without delay and prior to responding to any FRG which would normally have priority.

Technical Reference(s) FRZ-0.1B, CSFST Flowchart Attached w/ Revision # See EOP-1.0B, Attachment 1.A, Foldout Page Comments / Reference ODA-407 Attachment 8A EOS-1.3B, 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-1.3, Transfer to Cold Leg Recirculation.

Question Source: Bank # X 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 41 of 60 CPNPP NRC 2013 SRO Written Exam Worksheet 76 to 88 Rev h

ES-401 CPNPP NRC 2013 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From FRZ-0.1B, CSFST Flowchart Revision # 8 Page 42 of 60 CPNPP NRC 2013 SRO Written Exam Worksheet 76 to 88 Rev h

ES-401 CPNPP NRC 2013 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From EOP-1.0B, Attachment 1.A, Foldout Page Revision # 8 Page 43 of 60 CPNPP NRC 2013 SRO Written Exam Worksheet 76 to 88 Rev h

ES-401 CPNPP NRC 2013 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ODA-407 Attachment 8.A Revision # 14 Page 44 of 60 CPNPP NRC 2013 SRO Written Exam Worksheet 76 to 88 Rev h

ES-401 CPNPP NRC 2013 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ODA-407 Attachment 8.A Revision # 14 Comments /

Reference:

From EOS-1.3B Step 1 CAUTION Revision # 8 Page 45 of 60 CPNPP NRC 2013 SRO Written Exam Worksheet 76 to 88 Rev h

ES-401 CPNPP NRC 2013 SRO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 012 A2.01 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: Faulty bistable operation Proposed Question: SRO 86 Given the following conditions:

• Annunciator 1-ALB-6D, Window 1.3 - 1 OF 4 HI SETPT PR FLUX HI, is in alarm.

• All Power Range Nuclear Instruments read approximately 100%.

• OVERPOWER TRIP HIGH RANGE light for Power Range Nuclear Instrument N-42 is LIT on the NIS Panel due to failure of the High Setpoint High Flux Trip Bistable.

• ABN-703, Power Range Instrument Malfunction, is being performed to place the channel out-of-service for repairs.

Which of the following identifies the impact on the Reactor Protection System and what action should be taken to mitigate the situation?

A. An OP HI FLUX ROD STOP C-2 is generated and cannot be bypassed.

Reactor Trip bistables for Loop 2 must be placed in TRIP within one hour.

B. A Power Range High Flux Trip will be generated but can be blocked.

If the Reactor is to remain at 100% RTP, the QUADRANT POWER TILT RATIO must be determined using Core Power Distribution Measurement information.

C. A Power Range High Flux Trip will be generated but cannot be blocked.

Reactor Trip bistables for Loop 2 must be placed in TRIP within one hour.

D. An OP HI FLUX ROD STOP C-2 is generated and can be bypassed.

If the Reactor is to remain at 100% RTP, the QUADRANT POWER TILT RATIO must be determined using Core Power Distribution Measurement information.

Proposed Answer: D Page 46 of 60 CPNPP NRC 2013 SRO Written Exam Worksheet 76 to 88 Rev h

ES-401 CPNPP NRC 2013 SRO Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible because C-2 Rod Stop is generated, however, N16 Power Channel Defeat Switch must be aligned to Loop 2 but the bistables need not be placed in trip and the Rod Stop can be bypassed.

B. Incorrect. Plausible because a Power Range High Flux Trip will be generated and the performance action is correct, however, the trip cannot be blocked.

C. Incorrect. Plausible because the Power Range High Flux Trip will be generated and cannot be blocked, however, the bistables need not be placed in trip.

D. Correct. The C-2 Rod Stop is generated and QPTR must be monitored once every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

Technical Reference(s) ABN-703, Sections 2.1, 2.2, & Step 2.3.2 Attached w/ Revision # See Technical Specification SR 3.2.4.2 Comments / Reference Proposed references to be provided during examination: None Learning Objective: COMPREHEND the normal, abnormal and emergency operation of the Nuclear Instrumentation System.

APPLY the administrative requirements of the Nuclear Instrumentation System including Technical Specifications, TRM and ODCM.

Question Source: Bank # X 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, 5 Page 47 of 60 CPNPP NRC 2013 SRO Written Exam Worksheet 76 to 88 Rev h

ES-401 CPNPP NRC 2013 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ABN-703, Section 2.1 Revision # 8 Page 48 of 60 CPNPP NRC 2013 SRO Written Exam Worksheet 76 to 88 Rev h

ES-401 CPNPP NRC 2013 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ABN-703, Section 2.2 Revision # 8 Page 49 of 60 CPNPP NRC 2013 SRO Written Exam Worksheet 76 to 88 Rev h

ES-401 CPNPP NRC 2013 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ABN-703, Step 2.3.2 Revision # 8 Page 50 of 60 CPNPP NRC 2013 SRO Written Exam Worksheet 76 to 88 Rev h

ES-401 CPNPP NRC 2013 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From Tech Spec SR 3.2.4.2 Amendment # 156 Page 51 of 60 CPNPP NRC 2013 SRO Written Exam Worksheet 76 to 88 Rev h

ES-401 CPNPP NRC 2013 SRO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 039 A2.03 Importance Rating 3.7 Main and Reheat Steam System: Ability to (a) predict the impacts of the following malfunctions or operations on the MRSS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Indications and alarms for main steam and area radiation monitors (during SGTR)

Proposed Question: SRO 87 Given the following conditions:

• Unit 2 is at 35% power and ramping.

• Chemistry reports Steam Generator 2-01 primary to secondary leakage indicates a 120 gpd tube leak.

What indication is used to confirm the Chemistry sample and what is the required action?

A. 2-RE-2959 (COG-282) and the unit should be in MODE 3 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

B. 2-RE-2325A (N16-274) and the unit should be in MODE 3 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

C. 2-RE-2959 (COG-282) and the unit should be in MODE 3 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

D. 2-RE-2325A (N16-274) and the unit should be in MODE 3 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

Proposed Answer: C Page 52 of 60 CPNPP NRC 2013 SRO Written Exam Worksheet 76 to 88 Rev h

ES-401 CPNPP NRC 2013 SRO Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible as COG-282 is correct for confirmation that a Steam Generator Tube Leak exists, however, in accordance with ABN-106, the unit should be placed in MODE 3 in 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

B. Incorrect. Plausible as N16-274 is normally the most accurate indications, however, below 40% power the N16 radiation monitors are removed from poll. Further in accordance with ABN-106, the unit should be placed in MODE 3 in 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

C. Correct. Plausible as COG-282 is correct for confirmation that a Steam Generator Tube Leak exists and in accordance with ABN-106, the unit should be placed in MODE 3 in 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

D. Incorrect. Plausible as N16-274 is normally the most accurate indications, however, below 40% power the N16 radiation monitors are removed from poll. In accordance with ABN-106, the unit should be placed in MODE 3 in 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

Technical Reference(s) IPO-003A, step 5.6.16 Attached w/ Revision # See ABN-106, Steps 3.3.2 & 3.3.3 Comments / Reference 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.

ANALYZE the response to a Steam Generator Tube Leakage greater than or equal to 75 gpd per 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 Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 4, 5 Page 53 of 60 CPNPP NRC 2013 SRO Written Exam Worksheet 76 to 88 Rev h

ES-401 CPNPP NRC 2013 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From IPO-003A, step 5.6.16 Revision 28 Page 54 of 60 CPNPP NRC 2013 SRO Written Exam Worksheet 76 to 88 Rev h

ES-401 CPNPP NRC 2013 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ABN-106, Steps 3.3.2 & 3.3.3 Revision 10 Page 55 of 60 CPNPP NRC 2013 SRO Written Exam Worksheet 76 to 88 Rev h

ES-401 CPNPP NRC 2013 SRO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 004 G 2.2.25 Importance Rating 4.2 Chemical and Volume Control System: Equipment Control: Knowledge of the bases in Technical Specifications for limiting conditions for operations and safety limits Proposed Question: SRO 88 The limitation for a maximum of two Charging Pumps to be OPERABLE when less than 350°F is based on...

A. ...the reduced ECCS flow requirements when less than 350°F.

B. ...a mass addition pressure transient being relieved by a single PORV.

C. ...not exceeding the maximum flow rate to the RCS when less than 350°F.

D. ...preventing excessive cooldown of the Reactor Vessel Cold Leg nozzle.

Proposed Answer: B Explanation:

A. Incorrect. Plausible because there is a reduced ECCS flow requirement in MODE 4 when only one ECCS train shall be OPERABLE per Technical Specification LCO 3.5.3, however, limitations for the Charging Pumps are based on a single PORV in service.

B. Correct. As outlined in Technical Specification Surveillance Requirement (SR) 3.4.12.1.

Additionally, analyses demonstrate that either one RCS relief valve or the depressurized RCS and RCS vent can maintain RCS pressure below limits when only two charging pumps are actuated.

Thus, the LCO allows only two charging pumps OPERABLE during the LTOP MODES.

C. Incorrect. Plausible because there are Low Temperature Overpressure Protection System requirements while in MODE 4, however, limitations in the LCO are not applicable when >320°F with one RCP and operation, Pressurizer level 92%, and heat up rate limited to 60°F per hour.

D. Incorrect. Plausible because The Reactor Vessel Cold Leg Nozzles are referenced in Technical Specification LCO 3.5.3 Bases, however, limitations for the Charging Pumps are based on a single PORV in service.

Technical Reference(s) Technical Specification SR 3.4.12.1 Attached w/ Revision # See Technical Specification SR 3.4.12.1 Bases Comments / Reference Proposed references to be provided during examination: None Learning Objective: APPLY the administrative requirements of the Chemical and Volume Control System including Technical Specifications, TRM and ODCM.

Question Source: Bank # X Page 56 of 60 CPNPP NRC 2013 SRO Written Exam Worksheet 76 to 88 Rev h

ES-401 CPNPP NRC 2013 SRO Written Exam Worksheet Form ES-401-5 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 2 Page 57 of 60 CPNPP NRC 2013 SRO Written Exam Worksheet 76 to 88 Rev h

ES-401 CPNPP NRC 2013 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From Technical Specification SR 3.4.12.1 Amendment 156 Page 58 of 60 CPNPP NRC 2013 SRO Written Exam Worksheet 76 to 88 Rev h

ES-401 CPNPP NRC 2013 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From Technical Specification SR 3.4.12.1 Bases Revision 67 Page 59 of 60 CPNPP NRC 2013 SRO Written Exam Worksheet 76 to 88 Rev h

ES-401 CPNPP NRC 2013 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From Technical Specification SR 3.4.12.1 Bases Revision 67 Page 60 of 60 CPNPP NRC 2013 SRO Written Exam Worksheet 76 to 88 Rev h

ES-401 CPNPP NRC 2013 SRO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 2 K/A # 061 G2.2.40 Importance Rating 4.7 Auxiliary Feedwater: Ability to apply Technical Specifications for a system.

Proposed Question: SRO 89 Given the following conditions:

• Unit 1 at 2% power holding for Main Feedwater Pump 1A to be available.

• Twenty-four hours ago, the Turbine Driven Auxiliary Feedwater Pump failed Surveillance Requirement 3.7.5.2 and is being disassembled.

• Motor Driven Auxiliary Feedwater Pump (MDAFWP) 1-02 trips on overcurrent.

• The Unit Supervisor directs cross-connecting MDAFWP 1-01 to feed all four Steam Generators per ABN-305, Auxilary Feedwater System Malfunctions.

What is(are) the Technical Specification Required Actions?

A. Place Unit 1 in MODE 3 within 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> and MODE 4 within 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br />.

B. Place Unit 1 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 MODE 4 within 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br />.

C. Place Unit 1 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 restore one AFW train to OPERABLE within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

D. Immediately initiate action to restore one AFW train to OPERABLE.

Proposed Answer: D Page 1 of 50 CPNPP NRC 2013 SRO Written Exam Worksheet 89 to 100 Rev h

ES-401 CPNPP NRC 2013 SRO Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible with the TDAFWP inoperable and one MDAFWP feeding all four Steam Generators, all three AFW pumps are inoperable. This would be the correct action if believed the LCO 3.0.3 were entered for this Condition. However, LCO 3.7.5 Condition D NOTE states that LCO 3.0.3 should not be entered.

B. Incorrect. Plausible if believed that the TDAFWP and MDAFWP 1-02 were the only inoperable pumps as this is the correct actions for two trains of AFW inoperable.

However, cross-connecting the MDAFWP 1-01 to feed all four Steam Generators makes MDAFWP 1-01 inoperable as well. This would be the correct action for LCO 3.7.5 Condition C.

C. Incorrect. Plausible as this action would be indicative of a misunderstanding that the TDAFWP is not required OPERABLE in MODE 3 and thus placing the unit in MODE 3 exits the two trains inoperable LCO 3.7.5 Condition C and therefore 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> would remain of the 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> for Condition B for only one train inoperable. This is not correct, as the allowance for testing the TDAFWP in MODE 3 is for testing purposes only and does not exclude the pump from OPERABILITY requirements in MODE 3.

D. Correct. With the TDAFWP inoperable, MDAFWP 1-02 tripped and MDAFWP 1-01 feeding all four Steam Generators, all three AFW trains are inoperable in accordance with ABN-305 Note. LCO 3.7.5 Condition D is correct for all three trains inoperable.

Technical Reference(s) ABN-305 Attached w/ Revision # See Technical Specification 3.0.3 Comments / Reference Technical Specification 3.7.5 Proposed references to be provided during examination: LCO 3.7.5 Learning Objective: APPLY the administrative requirements of the Chemical and Volume 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 7, 10 55.43 2 Page 2 of 50 CPNPP NRC 2013 SRO Written Exam Worksheet 89 to 100 Rev h

ES-401 CPNPP NRC 2013 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ABN-305, Attachment 3 note Revision 7 Comments /

Reference:

From ABN-305, step 3.3.3 Revision 7 Page 3 of 50 CPNPP NRC 2013 SRO Written Exam Worksheet 89 to 100 Rev h

ES-401 CPNPP NRC 2013 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From Technical Specification 3.0.3 Revision 158 Page 4 of 50 CPNPP NRC 2013 SRO Written Exam Worksheet 89 to 100 Rev h

ES-401 CPNPP NRC 2013 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From Technical Specification 3.7.5, LCO B Revision 158 Page 5 of 50 CPNPP NRC 2013 SRO Written Exam Worksheet 89 to 100 Rev h

ES-401 CPNPP NRC 2013 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From Technical Specification 3.7.5, LCO C & D Revision 158 Page 6 of 50 CPNPP NRC 2013 SRO Written Exam Worksheet 89 to 100 Rev h

ES-401 CPNPP NRC 2013 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From Technical Specification 3.7.5, SR 3.7.5.2 Revision 158 Page 7 of 50 CPNPP NRC 2013 SRO Written Exam Worksheet 89 to 100 Rev h

ES-401 CPNPP NRC 2013 SRO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 103 G 2.4.4 Importance Rating 4.7 Containment System: Emergency Procedures/Plan: Ability to recognize abnormal indications for system operating parameters which are entry-level conditions for emergency and abnormal operating procedures Proposed Question: SRO 90 Given the following conditions:

• Unit 1 experienced an accident 20 minutes ago.

• All systems responded as expected with the following indications:

• Containment pressure is 16 psig and lowering.

• Neutron Source Range flux is on scale and stable.

• Reactor Coolant System (RCS) Subcooling is 0°F.

• Reactor Vessel Level Indication System lights from 11 through 49 above flange are LIT.

• All Steam Generator narrow range levels are 50% to 60%.

• All RCS Cold Leg temperatures are 270°F to 275°F.

• Containment Sump level is 817 ft.

• Pressurizer level is 0%.

• EOP-0.0A, Reactor Trip or Safety Injection, has been exited and the Shift Technical Advisor (STA) is reviewing Critical Safety Function Status Trees.

Which of the following Critical Safety Function Status Trees should the STA recommend entering?

A. FRS-0.2A, Response to Loss of Core Shutdown.

B. FRC-0.3A, Response to Saturated Core Cooling.

C. FRP-0.2A, Response to Anticipated Pressurized Thermal Shock.

D. FRZ-0.2A, Response to Containment Flooding.

Proposed Answer: D Page 8 of 50 CPNPP NRC 2013 SRO Written Exam Worksheet 89 to 100 Rev h

ES-401 CPNPP NRC 2013 SRO Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible because a review of the CSFST indicates that FRS-0.2A conditions are met, however, the ORANGE path for CONTAINMENT overrides the YELLOW path for SUBCRITICALITY.

B. Incorrect. Plausible because a review of the CSFST indicates that FRC-0.2A conditions are met, however, the ORANGE path for CONTAINMENT overrides the YELLOW path for CORE COOLING.

C. Incorrect. Plausible because a review of the CSFST indicates that FRP-0.2A conditions are met, however, the ORANGE path for CONTAINMENT overrides the YELLOW path for INTEGRITY.

D. Correct. Given the conditions listed, entry into FRZ-0.2A is the correct Functional Restoration Procedure.

Technical Reference(s) FRZ-0.2A, CSFST Attached w/ Revision # See FRP-0.2A, CSFST Comments / Reference FRC-0.3A, CSFST FRS-0.2A, CSFST Proposed references to be provided during examination: None Learning Objective: DISCUSS the symptoms or Entry Conditions for FRZ-0.2, Response to Containment Flooding.

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 9 of 50 CPNPP NRC 2013 SRO Written Exam Worksheet 89 to 100 Rev h

ES-401 CPNPP NRC 2013 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From FRZ-0.2A, CSFST Revision # 8 Page 10 of 50 CPNPP NRC 2013 SRO Written Exam Worksheet 89 to 100 Rev h

ES-401 CPNPP NRC 2013 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From FRP-0.2A, CSFST Revision # 8 Page 11 of 50 CPNPP NRC 2013 SRO Written Exam Worksheet 89 to 100 Rev h

ES-401 CPNPP NRC 2013 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From FRC-0.3A, CSFST Revision # 8 Page 12 of 50 CPNPP NRC 2013 SRO Written Exam Worksheet 89 to 100 Rev h

ES-401 CPNPP NRC 2013 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From FRS-0.2A, CSFST Revision # 8 Page 13 of 50 CPNPP NRC 2013 SRO Written Exam Worksheet 89 to 100 Rev h

ES-401 CPNPP NRC 2013 SRO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 2 K/A # 015 A2.05 Importance Rating 3.8 Nuclear Instrumentation: 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: Core void formation.

Proposed Question: SRO 91 Given the following conditions:

• Unit 1 has experienced a Large Break Loss of Coolant Accident during a reactor startup.

• All equipment functioned as designed.

• The crew has transitioned from EOP-0.0A, Reactor Trip or Safety Injection, to EOP-1.0A, Loss of Reactor or Secondary Coolant.

• EOP-0.0A, Attachment 2, Safety Injection Actuation Alignment, has been completed.

Which of the following describes the IMMEDIATE result that voiding in the downcomer region would have on source range instrumentation and the procedure used to mitigate these plant conditions?

A. The displacement of downcomer water would increase the neutron leakage and result in a higher source range count rate.

Continue in EOP-1.0A rather than transition to FRS-0.2A, Response to Loss of Core Shutdown.

B. A decrease in downcomer water density would reduce fission and result in a lower source range count rate.

Continue in EOP-1.0A rather than transition to FRS-0.2A, Response to Loss of Core Shutdown.

C. The displacement of downcomer water would increase the neutron leakage and result in a higher source range count rate.

Transition to FRS-0.2A, Response to Loss of Core Shutdown, rather than continue in EOP-1.0A.

D. A decrease in downcomer water density would reduce fission and result in a lower source range count rate.

Transition to FRS-0.2A, Response to Loss of Core Shutdown, rather than continue in EOP-1.0A.

Proposed Answer: A Page 14 of 50 CPNPP NRC 2013 SRO Written Exam Worksheet 89 to 100 Rev h

ES-401 CPNPP NRC 2013 SRO Written Exam Worksheet Form ES-401-5 Explanation:

A. Correct. Void formation in the downcomer will cause more neutron leakage outside the vessel and neutron flux SR indication will increase. Continuing with EOP-1.0A is the correct action.

B. Incorrect. Plausible because continuing with EOP-1.0A is the correct action, however, void formation in the downcomer will cause more neutron leakage outside the vessel and neutron flux SR indication will increase.

C. Incorrect. Plausible because void formation in the downcomer will cause more neutron leakage outside the vessel and neutron flux SR indication will increase. However, with the conditions listed, the highest transition to FRS-0.2A would be via a yellow path which does not take precedence over EOP-1.0A actions.

D. Incorrect. Plausible if believed that void formation would cause source range neutron instrumentation level to increase due to an increased fission rate, however, void formation in the downcomer will cause more neutron leakage outside the vessel and neutron flux SR indication will increase. Additionally, with the conditions listed, the highest transition to FRS-0.2A would be via a yellow path which does not take precedence over EOP-1.0A actions.

Technical Reference(s) LO21.MCO.MI8 Attached w/ Revision # See Comments / Reference Proposed references to be provided during examination: None Learning Objective: Describe the effects of core uncover on core kinetics and relate the Excore Nuclear System response to voiding.

Question Source: Bank # X 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 5 Page 15 of 50 CPNPP NRC 2013 SRO Written Exam Worksheet 89 to 100 Rev h

ES-401 CPNPP NRC 2013 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From LO21.MCO.MI8 Revision 3/30/04 Page 16 of 50 CPNPP NRC 2013 SRO Written Exam Worksheet 89 to 100 Rev h

ES-401 CPNPP NRC 2013 SRO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 2 K/A # 086 A2.02 Importance Rating 3.3 Fire Protection System: Ability to (a) predict the impacts of the following malfunctions or operations on the Fire Protection System; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Low FPS header pressure Proposed Question: SRO 92 Given the following conditions:

• X-07, Jockey Fire Pump, was placed in RUN to avoid problems with the pump coupling.

• Fire water header pressure has slowly lowered to 141 psig over several days.

• X-04, Electric Motor Driven Fire Pump, failed to automatically start.

• When X-HS-4091B, ELEC FIRE PMP START pushbutton is depressed, the Electric Motor Driven Fire Pump STARTS and then immediately STOPS when pressure reaches 155 psig.

• X-04, Electric Motor Driven Fire Pump was disabled from AUTO starting and stopped per SOP-904, Fire Protection Main Water Supply and Fire Pumps System.

• X-05, Diesel Fire Pump, has been placed in RUN.

Which of the following describes the impact on the Fire Suppression Water System and what Compensatory Measures that should be implemented?

A. One Fire Pump is inoperable.

Establish a backup fire suppression water supply within 14 days per STA-738, Fire Protection Systems/Equipment Impairments.

B. Two Fire Pumps are inoperable.

Establish a backup fire suppression water supply within 14 days per STA-738, Fire Protection Systems/Equipment Impairments.

C. One Fire Pump is inoperable.

Restore the Electric Fire Pump to OPERABLE status within 7 days or provide an alternate backup pump or supply per STA-738, Fire Protection Systems/Equipment Impairments.

D. Two Fire Pumps are inoperable.

Restore the Electric Fire Pump to OPERABLE status within 7 days or provide an alternate backup pump or supply per STA-738, Fire Protection Systems/Equipment Impairments.

Page 17 of 50 CPNPP NRC 2013 SRO Written Exam Worksheet 89 to 100 Rev h

ES-401 CPNPP NRC 2013 SRO Written Exam Worksheet Form ES-401-5 Proposed Answer: C Explanation:

A. Incorrect. Plausible because one fire pump is inoperable and a backup fire suppression water supply would satisfy STA-738 requirements, however, this must be done within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

B. Incorrect. Plausible because establishing a backup fire suppression water supply would satisfy STA-738 requirements, however, placing a diesel driven fire pump in run does not render the pump inoperable. Only one fire pump is inoperable.

C. Correct. As outlined in SOP-904 and STA-738.

D. Incorrect. Plausible if thought that the Jockey Fire Pump was also considered inoperable or that running the Diesel Fire Pump rendered it inoperable. Otherwise, the action is correct.

Technical Reference(s) STA-738, Attachment 8.A Attached w/ Revision # See Comments / Reference Proposed references to be provided during examination: None Learning Objective: COMPREHEND the administrative requirements of abnormal operations of the Fire Protection system.

Question Source: Bank # X 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 5 Page 18 of 50 CPNPP NRC 2013 SRO Written Exam Worksheet 89 to 100 Rev h

ES-401 CPNPP NRC 2013 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From STA-738, Attachment 8.A Revision 6 Page 19 of 50 CPNPP NRC 2013 SRO Written Exam Worksheet 89 to 100 Rev h

ES-401 CPNPP NRC 2013 SRO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 2 K/A # 002 G 2.4.18 Importance Rating 4.0 Reactor Coolant System: Knowledge of the specific bases EOPs.

Proposed Question: SRO 93 Given the following conditions on Unit 1:

• 1-PCV-455B, RC LOOP 1 PRZR SPR VLV has failed opened.

• An automatic Reactor Trip occurred on Pressurizer Pressure Low prior to performing ABN-705, Pressurizer Pressure Malfunction.

• Safety Injection actuated on Pressurizer Pressure Low at 1820 psig.

• EOP-0.0A, Reactor Trip or Safety Injection is in progress.

• Pressurizer Level is 25% and stable.

What action is required per EOP-0.0A Step 10 when 1-PCV-455B cannot be closed?

A. Stop Reactor Coolant Pumps 1-01 and 1-04.

B. Stop Reactor Coolant Pump 1-01 only.

C. Stop Reactor Coolant Pumps 1-04 and 1-02.

D. Stop Reactor Coolant Pump 1-04 only.

Proposed Answer: A Page 20 of 50 CPNPP NRC 2013 SRO Written Exam Worksheet 89 to 100 Rev h

ES-401 CPNPP NRC 2013 SRO Written Exam Worksheet Form ES-401-5 Explanation:

A. Correct. In accordance with EOP-0.0A Bases Table 1 stopping RCPs 1-01 and 1-04 will essentially stop spray flow through loop 1 spray valve. With pressurizer level at 25% spray flow will be negligible from RCPs 1-02 and 1-03.

B. Incorrect. Plausible as RCP 1-01 is the primary driving head for spray flow through 1-PCV-455B; however the EOP-0.0A Bases Table 1 shows that RCP 1-04 will still supply spray flow and therefore RCP 1-04 should also be stopped.

C. Incorrect. Plausible as RCP 1-04 is known to produce the greatest spray flow and therefore is logical to stop. Additionally stopping RCP 1-02 would lower the driving head farther. However, since 1-PCV-455B is primarily driven from the discharge of RCP 1-01, spray flow is maintained with RCPs 1-01 and 1-03 running.

D. Incorrect. Plausible as RCP 1-04 is known to produce the greatest spray flow and therefore is logical to stop. However, since 1-PCV-455B is primarily driven from the discharge of RCP 1-01, spray flow is maintained with RCPs 1-01, 1-02 and 1-03 running.

Technical Reference(s) EOP-0.0A, step 10 Attached w/ Revision # See 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 #

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 21 of 50 CPNPP NRC 2013 SRO Written Exam Worksheet 89 to 100 Rev h

ES-401 CPNPP NRC 2013 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From EOP-0.0A, step 10 Revision 8 Page 22 of 50 CPNPP NRC 2013 SRO Written Exam Worksheet 89 to 100 Rev h

ES-401 CPNPP NRC 2013 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From EOP-0.0A, Attachment 10 Bases Revision 8 Page 23 of 50 CPNPP NRC 2013 SRO Written Exam Worksheet 89 to 100 Rev h

ES-401 CPNPP NRC 2013 SRO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 3 Category # 1 K/A # G 2.1.4 Importance Rating 3.8 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: SRO 94 Given the following conditions:

• A Senior Reactor Operator (SRO) is completing the Shift Operations Watch Bill for the oncoming crew.

• The SRO is reviewing four Staff Reactor Operators (RO) work history to determine if each has met the MINIMUM requirements for maintaining an Active License status in accordance with ODA-315, Licensed Operator Maintenance Tracking.

Which of the following meets the MINIMUM requirements for ensuring that the ROs have satisfied the ODA-315 requirements?

The operator completed five 12-hour shifts...

A. ...during the previous quarter including turnovers, with one four hour absence for makeup of a missed training simulator scenario.

B. ...during the previous quarter including turnovers, during each shift the operator had to utilize Short Term Relief to attend a daily meeting.

C. ...with the fifth shift beginning at 1800 on the last day of the calendar quarter including turnovers.

D. ...during the previous quarter, during one shift the operator had a family emergency and was excused from end of shift turnover.

Proposed Answer: B Page 24 of 50 CPNPP NRC 2013 SRO Written Exam Worksheet 89 to 100 Rev h

ES-401 CPNPP NRC 2013 SRO Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible as this is 5 12-hour shifts including turnover. However, the four hour absence is in excess of the allowed Short Term Relief as allowed per OWI-107 and thus that shift would not count.

B. Correct. Per ODA-315, 5 12-hour shifts including turnover are the minimum to maintain an Active License status. Allowances for Short Term Relief per OWI-107 are allowed when completing the shifts per ODA-315.

C. Incorrect. Plausible as this is 5 12-hour shifts. However, in accordance with ODA-315, they must all be completed in the previous quarter.

D. Incorrect. Plausible as this is 5 12-hour shifts. However, both turnovers must be included per ODA-315 in order to be counted.

Technical Reference(s) ODA-315, section 6.2 Attached w/ Revision # See OWI-107, section 6.1.4 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 CPNPP 2012 Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 2 Page 25 of 50 CPNPP NRC 2013 SRO Written Exam Worksheet 89 to 100 Rev h

ES-401 CPNPP NRC 2013 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ODA-315, section 6.2 Revision 6 Page 26 of 50 CPNPP NRC 2013 SRO Written Exam Worksheet 89 to 100 Rev h

ES-401 CPNPP NRC 2013 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From OWI-107, section 6.1.4 Revision 8 Page 27 of 50 CPNPP NRC 2013 SRO Written Exam Worksheet 89 to 100 Rev h

ES-401 CPNPP NRC 2013 SRO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 3 Category # 2 K/A # G 2.2.7 Importance Rating 3.6 Equipment Control: Knowledge of the process for conducting special or infrequent tests Proposed Question: SRO 95 Which of the following individuals is expected to be in charge of High Risk, Infrequent Evolution, or Heightened Level of Awareness activities conducted on watch?

A. Plant Manager B. Director, Nuclear Operations C. Shift Operations Manager D. Unit Supervisor Proposed Answer: D Explanation:

A. Incorrect. Plausible because consideration is given to assigning this individual with the authority and experience to exercise continuous responsibility for the oversight of a particular test or evolution, however, it is the Unit Supervisor who acts as the SRO in charge.

B. Incorrect. Plausible because the Director, Nuclear Operations normally holds a SRO license, however, they are not in charge of these activities.

C. Incorrect. Plausible because the Shift Operations Manager has management responsibilities at the Station, however, it is the Unit Supervisor who is the SRO in charge.

D. Correct. As prescribed in OWI-107.

Technical Reference(s) OWI-107, Step 6.2.3.F Attached w/ Revision # See Comments / Reference Proposed references to be provided during examination: None Learning Objective: DESCRIBE the responsibilities of and ASSUME Control Room command function.

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

New Question History: Last NRC Exam Page 28 of 50 CPNPP NRC 2013 SRO Written Exam Worksheet 89 to 100 Rev h

ES-401 CPNPP NRC 2013 SRO Written Exam Worksheet Form ES-401-5 Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 5 Comments /

Reference:

From OWI-107, Step 6.2.3.F Revision # 8 Page 29 of 50 CPNPP NRC 2013 SRO Written Exam Worksheet 89 to 100 Rev h

ES-401 CPNPP NRC 2013 SRO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 3 Category # 2 K/A # G 2.2.35 Importance Rating 4.5 Equipment Control: Ability to determine Technical Specification Mode of Operation Proposed Question: SRO 96 Unit 1 is commencing a core reload from a full offload.

Which of the following describes the time at which core offload to MODE 6 is declared per RFO-102, Refueling Operation?

A. When Containment Closure is established in preparation for core reload.

B. When the first new fuel assembly is placed in the Containment Storage Rack.

C. When the first fuel assembly is engaged by the Refueling Machine Main Hoist in preparation for placement in the core.

D. When the first fuel assembly is engaged by the Refueling Machine Main Hoist is placed in its designated position in the core.

Proposed Answer: C Explanation:

A. Incorrect. Plausible because Containment Closure is established per OPT-408A prior to MODE 6 entry in RFO-102.

B. Incorrect. Plausible because it is specifically excluded as a MODE 6 entry criteria by RFO-102.

C. Correct. Per Step 5.5.4.A of RFO-102 when the first assembly is engaged in the Main Hoist MODE 6 is declared.

D. Incorrect. Plausible if thought that MODE 6 is not entered until an assembly is placed in the core.

Technical Reference(s) RFO-102 Attached w/ Revision # See Comments / Reference Proposed references to be provided during examination: None Learning Objective:

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

New Question History: Last NRC Exam Page 30 of 50 CPNPP NRC 2013 SRO Written Exam Worksheet 89 to 100 Rev h

ES-401 CPNPP NRC 2013 SRO Written Exam Worksheet Form ES-401-5 Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 2 Comments /

Reference:

From RFO-102 Revision 13 Page 31 of 50 CPNPP NRC 2013 SRO Written Exam Worksheet 89 to 100 Rev h

ES-401 CPNPP NRC 2013 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From Technical Specification LCO 3.0.4 Amendment # 150 Page 32 of 50 CPNPP NRC 2013 SRO Written Exam Worksheet 89 to 100 Rev h

ES-401 CPNPP NRC 2013 SRO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 3 Category # 3 K/A # G 2.3.5 Importance Rating 2.9 Radiation Control: Ability to use radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc.

Proposed Question: SRO 97 Page 33 of 50 CPNPP NRC 2013 SRO Written Exam Worksheet 89 to 100 Rev h

ES-401 CPNPP NRC 2013 SRO Written Exam Worksheet Form ES-401-5 Given the following conditions:

• Unit 1 has experienced a manual Reactor Trip and Safety Injection initiation based on lowering Reactor Coolant System (RCS) pressure and lowering Pressurizer level.

• EOP-0.0A, Reactor Trip and Safety Injection, is in progress.

• The following parameters are noted during diagnostics:

• All three Pressurizer Safety Valves are CLOSED.

• Both Pressurizer Spray Valves are CLOSED.

• Both Pressurizer PORVs are CLOSED.

• RCS subcooling is 15°F and stable.

• RCS pressure is 1530 psig and stable.

• All Reactor Coolant Pumps have been STOPPED.

• All Steam Generator Pressures are 1080 psig and slowly rising.

• COG-182, Condenser Off-Gas Radiation Monitor, is Normal.

• MSL-178 through MSL-181, Main Steam Line Radiation Monitors, are GREEN and stable.

• SGS-164, SG Blowdown Sample Radiation Monitor, is GREEN and stable.

• All Steam Generator levels are 52% and slowly rising under operator control.

• Containment Pressure is 0.2 psig and stable.

• Containment Recirc Sump level is 808' and stable.

• Containment Radiation on Grid 4 is Normal.

• The following Area Radiation Monitors on Grid 4 indicate RED and rising.

• 1-RE-6259A, PENET AREA RM 77S.

• 1-RE-6259B, PENET AREA RM 77N.

• X-RE-5570A, S WRGM EFFLUENT.

Which of the following indicates the proper action for optimal recovery?

A. Continue with EOP-0.0A, Reactor Trip or Safety Injection, as no procedure transitions have been identified.

B. Transition to EOS-1.1A, Safety Injection Termination, as ECCS flow is NOT required.

C. Transition to ECA-1.2A, LOCA Outside Containment, based on probable leakage into the Safeguards Building.

D. Transition to EOS-0.0, Rediagnosis, to identify what indication was misinterpreted.

Proposed Answer: C Page 34 of 50 CPNPP NRC 2013 SRO Written Exam Worksheet 89 to 100 Rev h

ES-401 CPNPP NRC 2013 SRO Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible because it could be evaluated that no procedural transitions have yet been identified, however, indications are correct for a transition to ECA-1.2A.

B. Incorrect. Plausible because with none of the major accident (SGTR, Faulted SG or LOCA) transitions identified, it could be interpreted that SI Termination should take place, however, subcooling would not allow SI Termination and a transition to ECA-1.2A is correct.

C. Correct. Interpretation of the readings on Grid 4 Radiation Monitors indicates that a transition to ECA-1.2A is required.

D. Incorrect. Plausible because the major diagnostic steps of EOP-0.0A have been completed, however, transition from EOP-0.0A to EOS-0.0A is not proper when a transition has not been identified.

Technical Reference(s) EOP-0.0A, Step 19 Attached w/ Revision # See ECA-1.2A, Entry Conditions Comments / Reference EOS-0.0A, Entry Conditions EOS-1.1A, Attachment 1.A Proposed references to be provided during examination: None Learning Objective: ANALYZE the diagnostic steps of EOP-0.0, Reactor Trip or Safety Injection.

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

New Question History: Last NRC Exam CPNPP 2012 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 5 Page 35 of 50 CPNPP NRC 2013 SRO Written Exam Worksheet 89 to 100 Rev h

ES-401 CPNPP NRC 2013 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From EOP-0.0A, Step 19 Revision # 8 Page 36 of 50 CPNPP NRC 2013 SRO Written Exam Worksheet 89 to 100 Rev h

ES-401 CPNPP NRC 2013 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ECA-1.2A, Entry Conditions Revision # 8 Page 37 of 50 CPNPP NRC 2013 SRO Written Exam Worksheet 89 to 100 Rev h

ES-401 CPNPP NRC 2013 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From EOS-0.0A, Entry Conditions Revision # 8 Comments /

Reference:

From EOS-1.1A, Attachment 1.A Revision # 8 Page 38 of 50 CPNPP NRC 2013 SRO Written Exam Worksheet 89 to 100 Rev h

ES-401 CPNPP NRC 2013 SRO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 3 Category # 3 K/A # G 2.3.15 Importance Rating 3.1 Radiation Control: Knowledge of radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc.

Proposed Question: SRO 98 Given the following conditions:

• Both Units are in MODE 1.

• Radiation Monitor X-RE-5895A, North Control Room Air intake fails LOW.

• Radiation Monitor X-RE-5895B, North Control Room Air intake is operating normally.

• Radiation Monitor X-RE-5896A, South Control Room Air Intake is operating normally.

• Radiation Monitor X-RE-5896B, South Control Room Air Intake is operating normally.

Which of the following identifies the Technical Specification requirements placed on the Control Room Heating, Ventilation, and Air Conditioning (HVAC) System?

A. Place both Control Room HVAC Trains in the Emergency Recirculation Mode within 30 days.

B. Secure the Control Room Makeup Air Supply Fan from the North Air Intake within 7 days.

C. Restore the affected Control Room Emergency Filtration/Pressurization System Train to OPERABLE status within 7 days.

D. Restore the affected Control Room Air Conditioning System Train to OPERABLE status within 30 days.

Proposed Answer: B Page 39 of 50 CPNPP NRC 2013 SRO Written Exam Worksheet 89 to 100 Rev h

ES-401 CPNPP NRC 2013 SRO Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible because this action would meet the Required Actions, however, the time frame of 30 days would not satisfy Technical Specification 3.3.7.A. This is the Required Action if both Trains of Control Room Emergency Filtration System (CREFS) Actuation instrumentation were INOPERABLE with the incorrect Completion Time.

B. Correct. With one Air Intake Radiation Monitor INOPERABLE, place the associated CREFS Train in the Emergency Recirculation Mode or secure the affected Intake Makeup Air Supply Fan within 7 days per Technical Specification LCO 3.3.7.A.

C. Incorrect. Plausible because it could be thought that the Radiation Monitor failure affected the CREFS per Technical Specification LCO 3.7.10.

D. Incorrect. Plausible because it could be thought that the Radiation Monitor failure affected the Control Room Air Conditioning System per Technical Specification LCO 3.7.11.

Technical Reference(s) Technical Specification LCO 3.3.7.A & B Attached w/ Revision # See Technical Specification LCO Table 3.3.7-1 Comments / Reference Technical Specification LCO 3.7.10.A Technical Specification LCO 3.7.11 Proposed references to be provided during examination: None Learning Objective: APPLY the administrative requirements of the Control Room Emergency Filtration System including Technical Specifications, TRM and ODCM.

Question Source: Bank # X 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 40 of 50 CPNPP NRC 2013 SRO Written Exam Worksheet 89 to 100 Rev h

ES-401 CPNPP NRC 2013 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From Technical Specification LCO 3.3.7.A Amendment # 156 Page 41 of 50 CPNPP NRC 2013 SRO Written Exam Worksheet 89 to 100 Rev h

ES-401 CPNPP NRC 2013 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From Technical Specification LCO 3.3.7.B Amendment # 156 Page 42 of 50 CPNPP NRC 2013 SRO Written Exam Worksheet 89 to 100 Rev h

ES-401 CPNPP NRC 2013 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From Technical Specification LCO Table 3.3.7-1 Amendment # 156 Page 43 of 50 CPNPP NRC 2013 SRO Written Exam Worksheet 89 to 100 Rev h

ES-401 CPNPP NRC 2013 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From Technical Specification LCO 3.7.10.A Amendment # 156 Page 44 of 50 CPNPP NRC 2013 SRO Written Exam Worksheet 89 to 100 Rev h

ES-401 CPNPP NRC 2013 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From Technical Specification LCO 3.7.11.A Amendment # 156 Page 45 of 50 CPNPP NRC 2013 SRO Written Exam Worksheet 89 to 100 Rev h

ES-401 CPNPP NRC 2013 SRO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 3 Category # 4 K/A # G 2.4.29 Importance Rating 4.4 Emergency Procedures / Plan: Knowledge of the emergency plan Proposed Question: SRO 99 Given the following conditions:

• The Emergency Response Organization has been activated.

• A Site Area Emergency has been declared and a Site Evacuation is in progress.

• The Emergency Coordinator is in the Emergency Operations Facility (EOF).

Which of the following actions may be delegated by the Emergency Coordinator?

A. Authorizing re-entry into evacuated areas.

B. Making Protective Action Recommendations to off-site authorities.

C. Approving shift schedules that support long-term emergency response.

D. Approving Notification Message Forms prior to sending.

Proposed Answer: C Explanation:

A. Incorrect. Plausible if thought that the Operations Support Center Manager can authorize re-entry as the position controls ERDC Teams.

B. Incorrect. Plausible because PARS are reviewed by Radiation Protection prior to sending, however, this function cannot be delegated.

C. Correct. As listed in EPP-109, Step 4.1.1 and is a responsibility of the Recovery Manager when the Recovery Organization is formed.

D. Incorrect. Plausible because the EOF Communicator sends the messages, however, the Emergency Coordinator must approve Notification Message Forms.

Technical Reference(s) EPP-109, Steps 4.1.1 & 4.2 Attached w/ Revision # See 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.

Page 46 of 50 CPNPP NRC 2013 SRO Written Exam Worksheet 89 to 100 Rev h

ES-401 CPNPP NRC 2013 SRO Written Exam Worksheet Form ES-401-5 Question Source: Bank # X 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 Comments /

Reference:

From EPP-109, Step 4.2 Revision # 14 4.2 Non-delegatable Duties The Emergency Coordinator shall not delegate decision making authority for:

• recommending use of potassium iodide.

• authorizing re-entry into evacuated onsite areas.

• authorizing personnel exposures in excess of 10CFR, Part 20 limits. [C-06380]

• making protective action recommendations to offsite authorities.

• approving notification messages. [C-05325]

Comments /

Reference:

From EPP-109, Step 4.1.1 Revision # 14 Page 47 of 50 CPNPP NRC 2013 SRO Written Exam Worksheet 89 to 100 Rev h

ES-401 CPNPP NRC 2013 SRO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 3 Category # 4 K/A # G 2.4.40 Importance Rating 4.5 Emergency Procedures / Plan: Knowledge of SRO responsibilities in emergency plan implementation Proposed Question: SRO 100 Given the following conditions:

• At 0200, the Shift Manager (acting as Emergency Coordinator) declared an ALERT based on RCS leakage of 200 gpm.

• At 0211, initial notifications to Offsite Agencies were completed.

• At 0300, the Unit 1 Reactor was tripped and Safety Injection actuated due to increased RCS leakage.

• At 0309, the RCS became saturated and all RVLIS lights indicated DARK.

• At 0327, the Shift Manager (acting as the Emergency Coordinator) declared an escalation of the event to a SITE AREA EMERGENCY.

• At 0339, notification of Emergency Classification escalation to Offsite Agencies was completed.

Which of the following statements is correct regarding the escalation and escalation notification?

A. The escalation was NOT timely.

The notification was timely.

B. The escalation was timely.

The notification was NOT timely.

C. The escalation was timely.

The notification was timely.

D. The escalation was NOT timely.

The notification was NOT timely.

Proposed Answer: A Explanation:

A. Correct. The escalation was made in 18 minutes which is NOT timely and the notification was made in 12 minutes which is timely.

B. Incorrect. Plausible if thought that escalation of an event has more time available than initial classification and that notification needs to be immediate as is required for an attack against the station (NRC Operations Center).

Page 48 of 50 CPNPP NRC 2013 SRO Written Exam Worksheet 89 to 100 Rev h

ES-401 CPNPP NRC 2013 SRO Written Exam Worksheet Form ES-401-5 C. Incorrect. Plausible if thought that escalation of an event has more time available than initial classification and the notification was made in less than 15 minutes and if thought that at notification needs to be immediate as is required for an attack against the station (NRC Operations Center).

D. Incorrect. Plausible because the escalation was not timely (18 minutes) and if thought that notification needs to be immediate as is required for an attack against the station (NRC Operations Center).

Technical Reference(s) EPP-203, Step 4.1.2.1 & 4.1.5 Attached w/ Revision # See EPP-201, Step 4.3 Comments / Reference Proposed references to be provided during examination: None Learning Objective: IDENTIFY time requirements for emergency notifications.

Question Source: Bank # X 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 Comments /

Reference:

From EPP-203, Step 4.1.2.1 Revision 16 Page 49 of 50 CPNPP NRC 2013 SRO Written Exam Worksheet 89 to 100 Rev h

ES-401 CPNPP NRC 2013 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From EPP-203, Step 4.1.5 Revision 16 Comments /

Reference:

From EPP-201, Step 4.3 Revision 12 Page 50 of 50 CPNPP NRC 2013 SRO Written Exam Worksheet 89 to 100 Rev h