ML101240979

2010-04 Draft Written Exam

ML101240979
Person / Time
Site:	Comanche Peak
Issue date:	03/29/2010
From:	Clyde Osterholtz Operations Branch IV
To:	Luminant Generation Co
References
50-445/10-301, 50-446/10-301
Download: ML101240979 (442)

Text

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 003 A1.09 Importance Rating 2.8 Reactor Coolant Pump System: Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the RCPS controls including: Seal flow and D/P Proposed Question: Common 1 Given the following conditions:

• Unit 1 is operating at 100% power.

• Reactor Coolant Pump 1-03 is experiencing seal degradation on #1 Seal.

• Reactor Coolant Pump 1-03 indicated #1 Seal leakoff flow is steady at 7.1 gpm.

• The following Annunciators are in alarm:

• 1-ALB-5A, Window 1.2 - ANY RCP SEAL 1 LKOFF FLO HI.

• 1-ALB-5A, Window 2.2 - ANY RCP SEAL 1 P LO.

• 1-ALB-5A, Window 3.2 - ANY RCP SEAL 2 LKOFF FLO HI.

• RCP 1-03 seal and bearing temperatures are stable.

Which ONE (1) of the following describes the required operator action based on the above indications and alarms?

A. Indicated #1 Seal leakoff flow exceeds 6 gpm. Trip the Reactor and then stop Reactor Coolant Pump 1-03.

B. Total #1 Seal leakoff flow exceeds 8 gpm. Trip the Reactor and then stop Reactor Coolant Pump 1-03.

C. Indicated #1 Seal leakoff flow is greater than 6 gpm but less than 8 gpm with stable seal inlet and pump bearing temperatures. Perform an orderly shutdown to secure Reactor Coolant Pump 1-03 within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

D. Total #1 Seal leakoff flow is greater than 6 gpm but less than 8 gpm with stable seal inlet and pump bearing temperatures. Raise seal injection flow to greater than 8 gpm to ensure proper Thermal Barrier cooling.

Proposed Answer: B Page 1 of 1 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible because if leakoff flow was greater than 6 gpm with rising temperatures than the RCP is immediately removed from service after tripping the Reactor, however, in this case seal and bearing temperatures are stable.

B. Correct. Total #1 Seal leakoff flow equals indicated #1 Seal leakoff flow plus the #2 Seal leakoff flow. #2 Seal leakoff flow is > 1 gpm if the annunciator is in alarm. The total flow is greater than 8 gpm based on the leakoff flow indication for #1 Seal (7.1 gpm) + > 1 gpm for #2 Seal leakoff flow, requiring a Reactor trip and stopping the RCP.

C. Incorrect. Plausible because it could be thought that the 8 gpm total leakoff flow limit was not exceeded which would then make these actions correct with the exception of stopping the RCP which must be done within eight hours.

D. Incorrect. Plausible because it could be thought that the 8 gpm total leakoff flow limit was not exceeded and this action is required to ensure RCP Thermal Barrier cooling.

Technical Reference(s) ABN-101, Steps 4.3.1 to 4.3.3 Attached w/ Revision # See ALM-0051A, 1-ALB-5A, Window 1.2 Comments / Reference ALM-0051A, 1-ALB-5A, Window 2.2 ALM-0051A, 1-ALB-5A, Window 3.2 Proposed references to be provided during examination: None Learning Objective: ANALYZE the indications and DESCRIBE the mitigation strategy and major steps taken relative to the Reactor Coolant System for:

• ABN-101, Reactor Coolant Pump Trip/Malfunction (OP51.SYS.RC1.OB17)

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, 10 55.43 Page 2 of 2 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ABN-101, Step 4.3.1 Revision # 10 Page 3 of 3 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ABN-101, Step 4.3.2 Revision # 10 Page 4 of 4 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ABN-101, Step 4.3.3 Revision # 10 Page 5 of 5 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ALM-0051A, 1-ALB-5A, Window 1.2 Revision # 5 Page 6 of 6 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ALM-0051A, 1-ALB-5A, Window 2.2 Revision # 5 Page 7 of 7 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ALM-0051A, 1-ALB-5A, Window 3.2 Revision # 5 Page 8 of 8 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 004 K6.20 Importance Rating 2.5 Chemical and Volume Control System: Knowledge of the effect that a loss or malfunction of the following will have on the CVCS components: Function of demineralizer, including boron loading and temperature limits Proposed Question: Common 2 The purpose of the Chemical and Volume Control System mixed bed demineralizers is to remove ___________ from the Reactor Coolant System, and during the process of alternating demineralizers, the operator should be alert for the possibility of the insertion of positive reactivity from the removal of boron from the ___________.

A. ionic impurities Reactor Coolant System B. ionic impurities demineralizer beds C. boron Reactor Coolant System D. boron demineralizer beds Proposed Answer: A Explanation:

A. Correct. This is the function of a mixed bed demineralizer. Alternating demineralizers can result in a positive reactivity addition due to absorption of boron from the Reactor Coolant System.

B. Incorrect. Plausible because removal of ionic impurities is correct, however, boron is not removed from the demineralizer beds rather it is absorbed.

C. Incorrect. Plausible because a positive reactivity addition due to absorption of boron from the Reactor Coolant System is correct, however, the purpose of a mixed bed demineralizer is to remove ionic impurities.

D. Incorrect. Plausible because a mixed bed demineralizer could be used to remove boron at end-of-life, however, boron is not removed from the demineralizer beds rather it is absorbed.

Technical Reference(s) SOP-103, Section 3.0 and Step 5.3.7 Attached w/ Revision # See OP51.SYS.CS1, Page 18 Comments / Reference Proposed references to be provided during examination: None Page 9 of 9 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Learning Objective: STATE the performance and design attributes of the following CVCS components, flowpaths, and features:

• Mixed Bed Demineralizers including boron loading and temperature limits (OP51.SYS.CS1.OB02)

Question Source: Bank # SYS.CS1.OB02-76 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, 6 55.43 Comments /

Reference:

From SOP-103, Section 3.0 Revision # 17 Page 10 of 10 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SOP-103, Step 5.3.7 Revision # 17 Comments /

Reference:

From OP51.SYS.CS1, Page 18 Revision # 2 FLOWPATH Reactor coolant enters the chemical and volume control system as letdown from the drain connection of Reactor Coolant Loop 3 Crossover Leg (on the suction side of Reactor Coolant Pump u-03). Hot letdown water from the reactor coolant system passes through a regenerative heat exchanger where it is cooled by the returning charging water. It then flows through one or more parallel orifices, which reduce the pressure and regulate the flow rate. The letdown flow rate is normally about 120 gpm and can be as high as 140 gpm. Piping length and system flow rate ensure the letdown water remains in the containment for a sufficient time to allow most of the N16 radioactivity in the water to decay.

The letdown, having been cooled, reduced in pressure and reduced in N16 activity level, passes through a containment penetration into the safeguards building. The non-regenerative letdown heat exchanger further reduces letdown temperature to near ambient. A second pressure reduction occurs across the letdown backpressure regulating valve.

Chemistry is controlled by passing the flow through demineralizers that remove ionic impurities and through a filter that removes solids. Flow can be routed to the boron thermal regeneration system for further chemistry adjustment if desired. Letdown then passes through a three-way valve, which either directs flow into the volume control tank for reuse or to the recycle holdup tank for processing.

Page 11 of 11 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 005 K2.01 Importance Rating 3.0 Residual Heat Removal System: Knowledge of bus power supplies to the following: RHR pumps Proposed Question: Common 3 Given the following conditions:

• Unit 2 is in MODE 4.

• Both Residual Heat Removal (RHR) Pumps are in operation.

• Safeguards Bus 2EA2 has just experienced an 86-1 lockout.

• Emergency Diesel Generator 2-02 has started.

Which ONE (1) of the following is the status of the Residual Heat Removal Pumps?

A. RHR Pump 2-01 is running and RHR Pump 2-02 is stopped.

B. RHR Pump 2-01 is stopped and RHR Pump 2-02 is running.

C. Both RHR Pumps are stopped.

D. Both RHR Pumps are running.

Proposed Answer: A Explanation:

A. Correct. Given the conditions listed, RHR Pump 2-01 will continue to run and RHR Pump 2-02 will be secured.

B. Incorrect. Plausible if thought that the power supplies were reversed, however, Safeguards Bus 2EA2 powers Residual Heat Removal Pump 2-02.

C. Incorrect. Plausible if thought that the 86-1 lockout affected both Safeguards Buses.

D. Incorrect. Plausible because it could be thought that because the Diesel Generator had started and the breaker had closed; however, the breaker remains open on an 86-1 lockout. Additionally, if an 86-2 lockout had occurred, the breaker would be closed and both RHR Pumps would be running.

Technical Reference(s) ABN-602, Step 2.3.3 Note Attached w/ Revision # See OP51.SYS.RH1.LN, Page 20 Comments / Reference OP51.SYS.AC2.LN, Pages 46, 47 & 48 Proposed references to be provided during examination: None Page 12 of 12 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Learning Objective: STATE the specific power supply including source of control power voltage for the Residual Heat Removal Pumps. (OP51.SYS.RH1.OB06)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

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

Reference:

From ABN-602, Step 2.3.3 Note Revision # 7 Page 13 of 13 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From OP51.SYS.RH1.LN, Page 20 Revision # 4 RESIDUAL HEAT REMOVAL PUMPS Each train of the RHR System contains a single stage centrifugal pump. Their 450 HP motors are powered from the Safeguards 6.9 KV buses uEA1 and uEA2. Each pump is designed to circulate 3800 gpm of flow at a discharge pressure of 167.2 psig. The suction parameters of the pump should not exceed 350ºF and 400 psig. Minimum flow through the pump should never fall below 500 gpm to ensure the mechanical seals receive enough flow to prevent overheating.

Comments /

Reference:

From OP51.SYS.AC2.LN, Page 46 Revision # 1 LOCKOUTS SAFEGUARDS BUSES There are two different distinct types of lockouts on the safeguards buses, an 86-1 and an 86-2 type lockout (Figure 20). The biggest difference is that for the 86-2 lockout the operator is afforded the capability of bypassing the lockout to allow the EDG to supply the bus for safe shutdown of the reactor. This is done by placing the EDG in the Emergency Start mode either manually or automatically. Either type lockout, if not bypassed, will open the supply breakers to the bus to de-energize the bus to isolate the fault. Also, breaker operation is blocked by open contacts in the closing circuitry to prevent the operator from reclosing the breaker with a fault on the bus.

Comments /

Reference:

From OP51.SYS.AC2.LN, Pages 47 & 48 Revision # 1 86-1 Lockout An 86-1 Lockout is caused by an overcurrent condition due to a phase-to-phase fault existing on the component (i.e., the bus, EDG or transformer).

An 86-1 lockout on a safeguards bus will trip open all input supply breakers to the bus. Also, relay contacts in the closing in the closing circuitry prevent the operator from reclosing the breaker with the fault present.

For the normal supply breaker, an 86-1 lockout relay actuation for the incoming transformer will also trip open the normal breaker and prevent the operator from closing the breaker with a fault on the transformer.

In the trip circuitry for the breakers supplied via XST2 (1EA1-1, 1EA2-1, 2EA1-2 & 2EA2-2), 94 relay contacts (94-1AX, 94-2AX) for E6 and W6 will also send a trip signal to the breakers on an 86-1 or 86-2 transformer fault either XST2 or on 1ST or a trip signal from either 7970 or 7980 switchyard breaker.

The alternate supply breaker also receives trip signals on the receipt of a 86-1 lockout relay for either the bus or the incoming transformer. Again, the operator is prevented from closing the breaker with a fault present with open contacts in the closing circuitry for the breaker.

The Emergency Diesel Generator output breaker receives trip signals 86-1 lockout relays on the bus or from the EDG itself. Another contact shown on Figure 20 (1-PS-9A) trips the output breaker when the EDG is tripped. The operator is prevented from bypassing the 86-1 lockouts for either the bus of the EDG and is prevented from reclosing the breaker until the condition is cleared.

Page 14 of 14 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 006 K5.05 Importance Rating 3.4 Emergency Core Cooling System: Knowledge of the operational implications of the following concepts as they apply to ECCS: Effects of pressure on a solid system Proposed Question: Common 4 Given the following conditions:

• An inadvertent Safety Injection Actuation has occurred.

Which ONE (1) of the following describes the adverse affect of allowing Safety Injection to continue without performing Safety Injection Termination?

A. ECCS Pumps will be running for extended time periods at minimum flow.

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

C. Centrifugal Charging Pumps running in the Injection Mode will collapse the Pressurizer bubble and pressurize the RCS to the PORV setpoint.

D. Reactor Coolant Pumps will be running without adequate pump seal cooling.

Proposed Answer: C Explanation:

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

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

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

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

Technical Reference(s) OP51.SYS.SI1.LN, Pages 19 & 22 Attached w/ Revision # See EOP-0.0A, Attachment 1A, Step 1 Note Comments / Reference Proposed references to be provided during examination: None Page 15 of 15 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Learning Objective: ANALYZE the indications and DESCRIBE the mitigation strategy and major steps taken, both initial and subsequent, for:

• EOS-1.1, Safety Injection Termination (LO21.SYS.SI1.OB29)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

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

Reference:

From OP51.SYS.SI1.LN, Page 19 Revision # 8 Centrifugal Charging Pumps The Centrifugal Charging Pumps are eleven stage centrifugal pumps manufactured by Pacific Pumps Company (section of the Dresser Corporation). These pumps provide flow to the Reactor Coolant System at any design pressure (RCS design pressure is 2485 psig). Shutoff head for each pump is ~

2590 psid. Each pump is rated for 150 gpm at a differential pressure of 2510 psid, with flow increasing to 550 gpm as system backpressure decreases to approximately 625 psid. 600 hp motors, supplied by the safeguards 6.9 kV busses, uEA1 & uEA2 power the pumps. The Centrifugal Charging Pumps are located in individual rooms on the 810' level of the Auxiliary Bldg.

Comments /

Reference:

From OP51.SYS.SI1.LN, Page 22 Revision # 8 Centrifugal Charging Pump Safety Injection Isolation Valves (u-8801A & B)

The Centrifugal Charging Pump Safety Injection Isolation Valves allow the Centrifugal Charging Pumps to inject borated water directly into the Reactor Coolant System Cold Legs. During normal power operations, these valves are closed to allow the operating Centrifugal Charging Pump to discharge through the normal charging line to maintain Pressurizer level. Upon receipt of a Safety Injection signal, the normal charging flow path is isolated, and u-8801A and u-8801B automatically open. These valves are located in the Containment Penetration Rooms (north on room Unit 2 and south room on Unit 1) on the 810' level of the SFGDs Bldg.

Page 16 of 16 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From EOP-0.0A, Attachment 1A, Step 1 Note Revision # 8 Page 17 of 17 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 007 K4.01 Importance Rating 2.6 Pressurizer Relief/Quench Tank System: Knowledge of the PRTS design feature(s) and/or interlock(s) that provide for the following: Quench tank cooling Proposed Question: Common 5 Given the following condition:

• Annunciator 1-ALB-05B, Window 2.3 - PRT HI TEMP has just alarmed.

Which ONE (1) of the following describes how the Pressurizer Relief Tank (PRT) is normally cooled per SOP-110A, Reactor Coolant Drain Tank System?

A. Recirculate the PRT through the Reactor Coolant Drain Tank Heat Exchanger, using Component Cooling Water to cool the Heat Exchanger.

B. Recirculate the PRT through the Reactor Coolant Drain Tank Heat Exchanger, using Reactor Makeup Water to cool the Heat Exchanger.

C. Drain the PRT to the Reactor Coolant Drain Tank while making up to the PRT from the Demineralized Water Storage Tank.

A. Drain the PRT to the Reactor Coolant Drain Tank while making up to the PRT from the Reactor Makeup Water Storage Tank.

Proposed Answer: A Explanation:

A. Correct. The procedure used to cool the PRT is SOP-110A, Reactor Coolant Drain Tank System.

The reference includes the steps required to cool the PRT which encompasses the NOTE at the end of Step 5.4.I stating that The PRT is now recirculating in the cooldown mode.

B. Incorrect. Plausible because the Reactor Coolant Drain Tank (RCDT) Heat Exchanger is used, however, use of Reactor Makeup Water would generate radioactive waste which is undesirable.

C. Incorrect. Plausible because there is a flowpath from the PRT to the RCDT, however, this method would generate waste which is undesirable.

D. Incorrect. Plausible because there is a flowpath from the PRT to the RCDT, however, this method would generate waste which is undesirable.

Technical Reference(s) SOP-110A, Section 5.4 Attached w/ Revision # See PO21.SYS.RC4, Figure 1 Comments / Reference ALM-0052A, 1-ALB-05B, Window 2.3 Page 18 of 18 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Proposed references to be provided during examination: None Learning Objective: STATE the function and operation of the following Reactor Coolant System components, flowpaths and features:

• Pressurizer Relief Tank (OP51.SYS.RC1.OB02)

Question Source: Bank # SYS.RC1.OB02-25 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam CPNPP 2009 Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 3, 10 55.43 Page 19 of 19 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SOP-110A, Section 5.4 Revision # 9 Page 20 of 20 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SOP-110A, Section 5.4 Revision # 9 Page 21 of 21 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From PO21.SYS.RC4, Figure 1 Revision # 1 Page 22 of 22 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ALM-0052A, 1-ALB-05B, Window 2.3 Revision # 5 Page 23 of 23 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ALM-0052A, 1-ALB-05B, Window 2.3 Revision # 5 Page 24 of 24 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 008 A4.07 Importance Rating 2.9 Component Cooling Water System: Ability to manually operate and/or monitor in the control room: Control of minimum level in the CCWS surge tank Proposed Question: Common 6 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 ONE (1) 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 if RED flagged on the CCW Pump Switch escutcheon.

B. Train A CCW Pump trips; Train B CCW Pump remains in standby if RED flagged on the CCW Pump Switch escutcheon.

C. Train A Safeguards Loop Isolation Valves close; Train B CCW Pump AUTO starts if GREEN flagged on the CCW Pump Switch escutcheon.

D. Train A Safeguards Loop Isolation Valves close; Train B CCW Pump remains in the standby if GREEN flagged on the CCW Pump Switch escutcheon.

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.

Page 25 of 25 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Technical Reference(s) ALM-0032A, 1-ALB-3B, Window 2.2 Attached w/ Revision # See ABN-502, Steps 2.1.a, 2.2, & 3.2.b Comments / Reference 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.

(OP51.SYS.CC1.OB09)

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 26 of 26 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ALM-0032A, 1-ALB-3B, Window 2.2 Revision # 7 Page 27 of 27 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ABN-502, Steps 2.1.a & 2.2 Revision # 6 Page 28 of 28 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ABN-502, Step 3.2.b Revision # 6 Page 29 of 29 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 010 K5.01 Importance Rating 3.5 Pressurizer Pressure Control System: Knowledge of the operational implications of the following concepts as they apply to the PZR PCS: Determination of condition of fluid in PZR, using steam tables Proposed Question: Common 7 Given the following conditions:

• A Reactor trip occurred 10 minutes ago due to a Loss of Coolant Accident.

• ECCS Cold Leg Injection is in progress.

Current conditions are as follows:

• Pressurizer liquid temperature is 604ºF.

• Pressurizer vapor temperature is 613ºF.

• RCS pressure is 1650 psig and stable.

• Pressurizer level is 25% and slowly rising.

Given these conditions, the Pressurizer liquid is __________ and the Pressurizer vapor is A. subcooled; saturated B. saturated; superheated C. saturated; saturated D. subcooled; superheated Proposed Answer: D Explanation:

A. Incorrect. Plausible because the liquid is subcooled, however, if the steam tables are misread the vapor could be read as saturated.

B. Incorrect. Plausible because the vapor is superheated, however, if the steam tables are misread the liquid could be read as saturated.

C. Incorrect. Plausible if the steam tables are misread the liquid and vapor could be read as saturated.

D. Correct. Saturation temperature for 1650 psig (1665 psia) is ~610.3ºF, therefore, the liquid is subcooled and the vapor is superheated.

Technical Reference(s) Steam Tables Attached w/ Revision # See Comments / Reference Page 30 of 30 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Proposed references to be provided during examination: Steam Tables Learning Objective: DESCRIBE or STATE how the following concepts or conditions apply to the Pressurizer Pressure and Level Control System:

• Determination of condition of fluid in PRZR, using steam tables.

(OP51.SYS.PP1.OB08)

Question Source: Bank # SYS.PP1.OB08-63 Modified Bank # (Note changes or attach parent)

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

Reference:

From Steam Tables Revision # N/A Saturation temperature for 1650 psia is 609.05ºF and saturation temperature for 1700 psia =

613.13ºF, therefore, saturation temperature for 1665 psia (1650 psig) = ~610.3ºF Page 31 of 31 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 010 G 2.1.32 Importance Rating 3.8 Pressurizer Pressure Control System: Conduct of Operations: Ability to explain and apply system limits and precautions Proposed Question: Common 8 Given the following conditions:

• The controlling Pressurizer Pressure instrument has failed low.

• The FIRST action performed by the Reactor Operator was to SELECT the Alternate Channel for control.

Which ONE (1) of the following describes the adverse effect of selecting the Alternate Channel for control PRIOR to performing the Initial Operator Actions of ABN-705, Pressurizer Pressure Malfunction?

A. Pressurizer Heaters that were energized will Trip and Lockout due to the breakers anti-pump feature.

B. Pressurizer Spray Valves could open with actual pressure low.

C. A Pressurizer Power Operated Relief Valve could open.

D. A momentary Low Pressure Trip signal on two out of four Pressurizer Pressure Channels could be initiated.

Proposed Answer: C Explanation:

A. Incorrect. Plausible because it is a misconception that a coincident TRIP and CLOSE signals would be present and cause the anti-pumping circuit to lockout the feeder breaker.

B. Incorrect. Plausible because it is a misconception that the improper sequence could result in improper spray flow due to the proportional/integral function of Master Pressure Controller.

C. Correct. ABN-705 requires that the PZR Pressure Controller be placed in MANUAL prior to selecting the Alternate Channel due to the possibility of inadvertently opening a PORV due to the proportional/integral function of Master Pressure Controller.

D. Incorrect. Plausible because it is a misconception that the improper sequence could momentarily cause a second low pressure signal and result in an inadvertent Reactor trip.

Technical Reference(s) ABN-705, Steps 2.2.b & 2.3 Attached w/ Revision # See Comments / Reference Proposed references to be provided during examination: None Page 32 of 32 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Learning Objective: ANALYZE the indications and DESCRIBE the mitigation strategy for the following procedures as they affect the Pressurizer Pressure and Level Control system:

• ABN-705, Pressurizer Pressure Malfunction (OP51.SYS.PP1.OB14)

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 33 of 33 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ABN-705, Step 2.2.b Revision # 12 Page 34 of 34 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ABN-705, Step 2.3 Revision # 12 Page 35 of 35 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 012 K3.01 Importance Rating 3.9 Reactor Protection System: Knowledge of the effect that a loss or malfunction of the RPS will have on the following: CRDS Proposed Question: Common 9 Given the following conditions:

• Unit 1 is performing OPT-447A, Mode 1, 3 and 4 Train A SSPS Actuation Logic Testing.

• Reactor Trip Breaker, RTA, and Reactor Trip Bypass Breaker, BYA are both CLOSED per OPT-447A, Mode 1, 3 and 4 Train A SSPS Actuation Logic Testing.

• During the test, Annunciator 1-ALB-6D, Window 2.5 - SSPS GENERAL WARNING TRAIN B is received due to failure of an SSPS 15 VDC power supply.

.Which ONE (1) of the following describes the expected plant response for this condition?

A. Reactor Trip and Bypass Breakers, RTA, RTB, and BYA will trip.

B. Only Reactor Trip Breakers, RTA and RTB will trip.

C. Only Reactor Trip Bypass Breaker, BYA will trip.

D. No automatic responses other than a General Warning Alarm.

Proposed Answer: A Explanation:

A. Correct. The Train A Bypass Breaker being connected and closed creates a General Warning Condition (as well as the test lineup) on the Train A and when a General Warning Alarm (GWA) condition occurs on the Train B an undervoltage trip condition is created to trip all Reactor Trip and Bypass Breakers.

B. Incorrect. Plausible because it is a misconception that an auto shunt trip signal was generated which would trip only the Reactor Trip Breakers.

C. Incorrect. Plausible because it is a misconception that a direct shunt trip signal was generated which would trip only the Reactor Trip Bypass Breaker.

D. Incorrect. Plausible because it is a misconception that only the one degraded condition (failure of an auctioneered power supply) is present; however, with a second GWA condition on the Train A an undervoltage trip condition is generated.

Technical Reference(s) OP5.1.SYS.ES1.LN, Pages 35, 40, & 42 Attached w/ Revision # See ALM-0064A, 1-ALB-6D, Window 2.5 Comments / Reference Page 36 of 36 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Proposed references to be provided during examination: None Learning Objective: For the Reactor Trip, Block, TSLBs and Permissive Status Indications, DESCRIBE the meaning of any given window. (OP51.SYS.ES1.OB14)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 6 55.43 Page 37 of 37 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ALM-0064A, 1-ALB-6D, Window 2.5 Revision # 6 Page 38 of 38 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From OP5.1.SYS.ES1.LN, Page 35 Revision # 1 OUTPUTS OF SSPS An automatic reactor trip is accomplished by de-energizing the undervoltage coil in the reactor trip breakers and bypass breakers, and energizing the shunt trip coil in the reactor trip breakers. Output relays which are called Slave Relays, located in the Output Relay Cabinets, actuate field equipment when a safeguard actuation is required.

As mentioned earlier, the functions available from a train of SSPS are either a reactor trip or actuation of one or more Engineering Safeguards Features (ESF). Reactor trips are initiated by each train of SSPS tripping open its associated reactor trip breaker RTA (RTB) and the bypass breaker BYB (BYA) in the opposite train (See OP51.SYS.ES1.FG01). In this way if the opposite train is in test with its reactor trip breaker disabled, the reactor would still trip when the required conditions are present. This allows the testability at power requirement to be met.

Page 39 of 39 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From OP5.1.SYS.ES1.LN, Page 40 Revision # 1 SSPS General Warning Alarms Annunciators are provided in the Control Room to indicate trouble with either protection system train.

The protection system train trouble alarm, "SSPS General Warning Train A or B" circuit is shown in OP51.SYS.ES2.FG24 below.

OP51.SYS.ES2.FG24 - SSPS General Warning Alarm The SSPS General Warning annunciators are not operated through the multiplexing scheme but are signaled directly from slave relays in the trains.

If trouble in both trains should develop simultaneously, the reactor will be tripped automatically by the General Warning system. An alarm relay is energized when none of conditions exist. When a problem occurs in a train, the alarm relay in that train deenergizes, opening a contact in a parallel path between the UV coil on the breaker and the undervoltage coil driver. However, the alarm relay of the other train must deenergize for the circuit to open and trip the reactor. Another contact on the alarm relay activates a Control Room annunciator.

Page 40 of 40 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From OP5.1.SYS.ES1.LN, Page 42 Revision # 1 REACTOR TRIP SYSTEM The Solid State Protection System automatically trips the plant whenever the plant conditions monitored by nuclear and process instrumentation reach specified limits. The system also provides alarms that alert the plant operator when manual action is required to prevent a plant trip. When the protection system logics receive the trip signals from the field inputs, they send a signal to deenergize the undervoltage coils and, via a separate relay at the reactor trip breakers, energize the shunt trip coils of the reactor trip breakers. (Auto trips only trip the UV coils on the bypass breakers, their shunt trips are caused only by manual trips.) This causes the breakers to open, interrupting power to the Control Rod Drive Mechanisms, which allows the Rod Control Cluster Assemblies to drop into the core.

A bypass breaker in parallel with each reactor trip breaker allows on line testing of the trip breakers.

The train A protection system trips the train A reactor trip breaker and the train B bypass breaker. The train B reactor trip breaker and train A bypass breakers are tripped by the train B protection system.

When a reactor trip breaker is bypassed the protection train associated with that breaker is considered to be inoperative. The bypass breakers are interlocked such that, if an attempt is made to close a second bypass breaker while one is closed and both are connected (racked in), the second bypass breaker will shut but both bypass breakers will be immediately reopened. This prevents both trains from being bypassed simultaneously. Actually, both reactor trip breakers may open too. The reactor trip breakers would open due to a General Warning in both trains, as mentioned above, because having a bypass breaker closed and connected causes a General Warning in the train that it is bypassing. If the bypass breakers open fast enough, the General Warning may not open the reactor trip breakers.

Page 41 of 41 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 013 A3.02 Importance Rating 4.1 Engineered Safety Feature Actuation System: Ability to monitor automatic operation of the ESFAS including: Operation of actuated equipment Proposed Question: Common 10 Which ONE (1) of the following would cause the:

1. Spent Fuel Pool Cooling Pumps to trip?

2. Why does the Spent Fuel Pool Cooling Pump trip?

3. Where can the Spent Fuel Pool Cooling Pump be restarted?

A. Phase A Containment Isolation Signal or Spent Fuel Pool high level.

Prevent Spent Fuel Pool overflow on Spent Fuel Pool high level.

Locally at the Motor Control Center.

B. Safety Injection Actuation Signal or Spent Fuel Pool high level.

Prevent Spent Fuel Pool overflow on Spent Fuel Pool high level.

Locally at the Spent Fuel Pool Panel.

C. Safety Injection Actuation Signal or Spent Fuel Pool low level.

Prevent Spent Fuel Pool Cooling Pump cavitation on Spent Fuel Pool low level.

Locally at the Motor Control Center.

D. Phase A Containment Isolation Signal or Spent Fuel Pool low level.

Prevent Spent Fuel Pool Cooling Pump cavitation on Spent Fuel Pool low level.

Locally at the Spent Fuel Pool Panel.

Proposed Answer: C Page 42 of 42 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible because there is an ESFAS signal that trips the SFP Cooling Pump, however, it is a Safety Injection Actuation Signal (SIAS). It is a misconception that the SFP Cooling Pumps are tripped to prevent SFP overflow on SFP high level, the pump trips on low SFP level. All AUTO trips are overridden by LOCAL SFP Cooling Pump start at the MCC.

B. Incorrect. Plausible because the SFP Cooling Pump will trip on an SIAS, however, it is a misconception that SFP Cooling Pumps are tripped to prevent SFP overflow on SFP high level. All AUTO trips are overridden by LOCAL SFP Cooling Pump start at the MCC.

C. Correct. SFP low level, Safety Injection Actuation Signal or Blackout will trip SFP Cooling Pumps.

These trips are BYPASSED when the SFP Cooling Pump is started at its feeder breaker Motor Control Center.

D. Incorrect. Plausible because it is a misconception that local operation of SFP Cooling Pumps from SFP Panel overrides trips, however, a SFP low level will trip the SFP Cooling Pumps to prevent cavitation.

Technical Reference(s) M1-2235, Sheet 01 Attached w/ Revision # See SOP-506, Section 3.0 Comments / Reference Proposed references to be provided during examination: None Learning Objective: DISCUSS the effects on plant equipment and unit operation upon actuation of an ESF or Reactor Trip signal. (OP51.SYS.ES1.OB10)

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 43 of 43 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From M1-2235, Sheet 01 Revision # CP-8 Page 44 of 44 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SOP-506, Section 3.0 Revision # 17 Page 45 of 45 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 013 K6.01 Importance Rating 2.7 Engineered Safety Features Actuation System: Knowledge of the effect that a loss or malfunction of the following will have on the ESFAS: Sensors and detectors Proposed Question: Common 11 Given the following condition:

• One (1) Intermediate Range (IR) Containment pressure detector is failed low.

Which ONE (1) of the following conditions would generate an Engineered Safety Features Containment HI-3 pressure signal?

A. One additional IR Containment pressure detectors sensing pressure 6.2 psig.

B. One additional IR Containment pressure detectors sensing pressure 18.2 psig.

C. Two additional IR Containment pressure detectors sensing pressure 6.2 psig.

D. Two additional IR Containment pressure detectors sensing pressure 18.2 psig.

Proposed Answer: D Explanation:

A. Incorrect. Plausible because the failed channel plus the additional channel would meet the 2/4 coincidence required for HI-3 actuation, however, one of them is failed low. Additionally, 6.2 psig is the setpoint for HI-2 not HI-3.

B. Incorrect. Plausible because the failed channel plus the additional channel would meet the 2/4 coincidence required for HI-3 actuation, however, one of them is failed low. 18.2 psig is the setpoint for HI-3.

C. Incorrect. Plausible because two additional channels would meet the 2/4 coincidence required for HI-3 actuation, however, 6.2 psig is setpoint for HI-2 not HI-3.

D. Correct. Two additional channels would meet the 2/4 coincidence required for HI-3 actuation and 18.2 psig is the correct HI-3 setpoint.

Technical Reference(s) ALM-0022, 1-ALB-2B, Window 2.10 Attached w/ Revision # See ALM-0022, 1-ALB-2B, Window 3.10 Comments / Reference Proposed references to be provided during examination: None Page 46 of 46 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Learning Objective: LIST and EXPLAIN the Containment Spray System design features which provide for the trips, permissives, and interlocks associated with the following:

• Containment Spray/Phase B Isolation automatic actuation (OP51.SYS.CT1.OB09)

Question Source: Bank # SYS.CT1.OB09-11 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 Page 47 of 47 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ALM-0022, 1-ALB-2B, Window 3.10 Revision # 9 Page 48 of 48 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ALM-0022, 1-ALB-28, Window 2.10 Revision # 9 Page 49 of 49 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 002 K3.02 Importance Rating 3.0 Containment Cooling System: Knowledge of the effect that a loss or malfunction of the CCS will have on the following:

Containment instrumentation readings Proposed Question: Common 12 Which ONE (1) of the following identifies the impact on Reactor Coolant System pressure indication caused by a loss of Containment Cooling?

A. RCS wide range pressure (PI-405) will indicate higher than actual pressure; RCS narrow range pressure indication (PI-403A) will indicate lower than actual pressure.

B. Both RCS wide and narrow range indications (PI-405 and PI-403A) will read lower than actual.

C. Both RCS wide and narrow range indications (PI-405 and PI-403A) will read higher than actual.

D. RCS wide range pressure (PI-405) will indicate lower than actual pressure; RCS narrow range pressure indication (PI-403A) will indicate higher than actual pressure.

Proposed Answer: B Explanation:

A. Incorrect. Plausible because the narrow range pressure indication will read lower than actual, however, the wide range pressure indication pressure transmitter is of the same construction and will also read lower.

B. Correct. An increase in Containment temperature results in increased Containment pressure which causes Reactor Coolant System pressure to read less than actual. PI-403A is a narrow range indicator used for LTOP conditions (range 0-700 psig) and uses the same type and range of pressure transmitter as PI-405, both of which go down as pressure in Containment pressure rises.

C. Incorrect. Plausible if thought that an increase in Containment temperature results in a decreased Containment pressure which would cause Reactor Coolant System pressure to read less than actual.

D. Incorrect. Plausible because the wide range pressure indication will read lower than actual, however, the narrow range pressure indication pressure transmitter is of the same construction and will also read lower.

Technical Reference(s) LO21.GFC.SEN.LN, Page 12 Attached w/ Revision # See LO21.MCO.MI7.LP, Pages 17 & 18 Comments / Reference Proposed references to be provided during examination: None Page 50 of 50 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Learning Objective: EXPLAIN why instrument accuracy changes under accident conditions.

(LO21.MCO.MI7.OB02)

EXPLAIN how various vital instruments respond to accident conditions.

(LO21.MCO.MI7.OB03)

Question Source: Bank # MCO.MI7.OB03-1 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam CPNPP 2005 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 5 55.43 Comments /

Reference:

From LO21.GFC.SEN.LN, Page 12 Revision # 2 ENVIRONMENTAL EFFECTS ON OPERATION Environmental conditions such as pressure, temperature, and radiation levels can lead to pressure measurement inaccuracies and to potential detector damage.

Pressure detectors are usually made of materials with a low temperature coefficient of expansion.

Therefore, varying temperature conditions do not cause significant variations or errors in pressure measurements.

Pressure variations in the reference connection of a D/P cell, however, do affect measurements. If the reference connection (typically the low-pressure side) is vented to an enclosed volume (containment building atmosphere), any pressure variations in that enclosed volume will introduce errors into the pressure measurements. If containment building pressure increases, the differential pressure sensed by the differential pressure (D/P) cell decreases, and the output signal also decreases. Bourdon tubes are similar, since atmospheric pressure provides the reference. Instruments not using atmosphere as a reference (such as D/P cells across a flow orifice) remain unaffected by atmospheric pressure changes.

Page 51 of 51 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From LO21.MCO.MI7.LP, Pages 17 & 18 Revision # 0 7 1. Pressure and differential pressure (P) detectors.

a. Several manufacturers' detectors are currently completely qualified.

b. Qualification program results are not uniform, but within required accuracy limits.

c. Primary transducer effects:

1) Bourdon tubes/diaphragms do not have much of a response to adverse containment conditions.

a) Material has a low temperature expansion coefficient.

b) Sealed case around tube.

c) When temperature and pressure increase, the transducer output decreases slightly.

d) If gauge pressure is measured, an increase in containment pressure from 15 to 60 psig will lower output by 45 psig.

Page 52 of 52 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 026 K4.01 Importance Rating 4.2 Containment Spray System: Knowledge of CSS design feature(s) and/or interlock(s) that provide for the following: Source of water for CSS, including recirculation phase after LOCA Proposed Question: Common 13 The Containment Spray Pump suction should be shifted to the Containment Sump at a Refueling Water Storage Tank level of __________ and the Containment Spray Pumps should be stopped at a Refueling Water Storage Tank level of __________.

A. 6% 0%

B. 12% 6%

C. 24% 6%

D. 33% 9%

Proposed Answer: A Explanation:

A. Correct. The Containment Spray Pumps suction swaps to the Containment Sump at an RWST level of 6%. Containment Spray Pumps are secured when RWST level reaches 0%, if running, per EOS-1.3A.

B. Incorrect. Plausible because the RWST Empty configuration was previously 12% and Containment Spray Pumps were secured at 6% (prior to GSI-191, Experimental Studies of Loss-of-Coolant-Accident-Generated Debris Accumulation and Head Loss).

C. Incorrect. Plausible because the Containment Spray Pumps suction swap to the Containment Sump was previously at an RWST level of 24% and the Containment Spray Pumps were secured at 6% (prior to GSI-191, Experimental Studies of Loss-of-Coolant-Accident-Generated Debris Accumulation and Head Loss).

D. Incorrect. Plausible because the ECCS Pumps suction swaps to the Containment Sump at 33%

and ECCS Pumps are secured at 9%, if running.

Technical Reference(s) ALM-0042A, 1-ALB-4B, Windows 3.7 & 4.7 Attached w/ Revision # See EOS-1.3A, Entry Conditions Comments / Reference EOS-1.3A, Step 3 Caution and Step 4 Proposed references to be provided during examination: None Page 53 of 53 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Learning Objective: STATE the physical connections and EVALUATE the cause-effect relationship between the Containment Spray System and the following systems, components or events:

• Containment Sumps (OP51.SYS.CT1.OB13)

Question Source: Bank # SYS.CT1.OB13-1 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, 10 55.43 Page 54 of 54 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ALM-0042A, 1-ALB-4B, Window 3.7 Revision # 7 Page 55 of 55 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ALM-0042A, 1-ALB-4B, Window 3.7 Revision # 7 Page 56 of 56 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ALM-0042A, 1-ALB-4B, Window 4.7 Revision # 7 Page 57 of 57 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ALM-0042A, 1-ALB-4B, Window 4.7 Revision # 7 Page 58 of 58 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From EOS-1.3A, Entry Conditions Revision # 8 Page 59 of 59 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From EOS-1.3A, Step 3 Caution Revision # 8 Comments /

Reference:

From EOS-1.3A, Step 4 Revision # 8 Page 60 of 60 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 039 A1.03 Importance Rating 2.6 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: Primary system temperature indications, and required values, during main steam system warm-up Proposed Question: Common 14 Given the following conditions on Unit 1:

• A Reactor and Turbine trip has just occurred from 100% power.

• The Loop #1 TAVE Channel has failed high.

Which ONE (1) of the following describes the effect on the Steam Dump System due to this failure and any operator action that is required?

A. The Steam Dumps will remain open and reduce TAVE to below the No-Load TAVE of 557ºF. Close the Steam Dump Valves by placing one of the two Steam Dump Interlock Select Switches to OFF.

B. The Steam Dumps will remain open and reduce TAVE to below the No-Load TAVE of 557ºF. Take MANUAL control of the Plant Trip Controller and close the Steam Dump Valves.

C. The Steam Dump System will automatically shift to the Pressure Control Mode and control at set pressure. Ensure the set pressure maintains TAVE at the No-Load value of 557ºF.

D. The Steam Dump System will automatically shift from the Load Rejection Controller to the Plant Trip Controller and the Steam Dump Valves will close. Select the faulty TAVE signal on the Tave Channel DEFEAT Switch.

Proposed Answer: A Page 61 of 61 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Explanation:

A. Correct. The Steam Dump Valves will control based on the differential temperature between the Average TAVE and No-Load TAVE. The Average TAVE will be higher based on the channel failure and keep the valves open past where they should be closed. EOP-0.0A, Reactor Trip or Safety Injection requires the operator to stop dumping steam if temperature goes below 557ºF and this is accomplished by selecting a Steam Dump Interlock Switch to OFF.

B. Incorrect. Plausible because the Steam Dump Valves will lower temperature below the program, however, the Plant Trip Controller has no manual controls for operators to use.

C. Incorrect. Plausible because it could be thought that the system would swap to Pressure Control because this is the mode normally used at low power or during plant startup and the TAVE signal would not be controlling, however, the system does not automatically shift to Pressure Control.

D. Incorrect. Plausible because the automatic shift to the Plant Trip Controller does occur and it could be thought that this control would be unaffected by the TAVE failure, however, the TAVE failure affects both the Load Rejection and Plant Trip Controllers and the Steam Dump Valves would still remain open below No-Load TAVE. The action to defeat the failed TAVE signal is also an action directed by ABN-704, TC/N-16 Instrumentation Malfunction.

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 OP51.SYS.SD1.LN, Figure 3 OP51.SYS.SD1.LN, Page 13 Proposed references to be provided during examination: None Learning Objective: STATE the function of the Steam Dump Mode Selector and Interlock Selector Switches and DESCRIBE Steam Dump operation for each position of each switch. (OP51.SYS.SD1.OB10)

EXPLAIN the operation of the Steam Dump System, explanation should include the following Modes, conditions or evolutions:

• Tave Mode (OP51.SYS.SD1.OB12)

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 62 of 62 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ABN-704, Step 2.2.a Revision # 10 Page 63 of 63 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ABN-704, Steps 2.3.2 and 2.3.3 Revision # 10 Comments /

Reference:

From EOP-0.0A, Step 9 RNO a Revision # 8 Page 64 of 64 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From OP51.SYS.SD1.LN, Figure 3 Revision # 2 STEAM DUMP CONTROL SYSTEM System is shown in TAVE Mode at Power LOAD REJECTION PLANT TRIP CONTROLLER STEAM CONTROLLER Tref AVE TAVE TAVE CONTROLLER PT-505 NO LOAD STEAM PRESSURE HEADER SET PT. PRESSURE (PIC-507) (PT-507)

HI-1 HI-1 B/S 9.5°F B/S 20.0°F HI-2 HI-2 CTRL B/S B/S 40.0°F OUTPUT M/A 14.0°F 1%

100% 11.11% 100% 2.5% 1.8 psid

°F °F DUMP 50% DUMP 50% PC-507 DEMAND DEMAND 0% 0%

5°F 9.5°F 14.0°F 0°F 20.0°F 40.0°F MODE SELECTOR SWITCH TAVE STM OPENS CLOSES RESET P-4 TR B PRESS RX TRIP & BYPASS BKRS OPEN Closes on STM Closes on TAVE PRESS Selected Selected Opens on STM PRESS Selected DEMAND I/P CONV. INDICATION GROUP 1 (TRIP OPEN) GROUP 2 (TRIP OPEN)

• COOLDOWN VALVES FULL FULL FULL FULL OPEN OPEN OPEN OPEN DEMAND DEMAND DEMAND DEMAND FULL FULL FULL FULL CLOSED 0% 25% CLOSED25% 50% CLOSED 50% 75% CLOSED 75% 100%

BANK 1 VLV POSITIONERS BANK 2 VLV POSITIONERS BANK 3 VLV POSITIONERS BANK 4 VLV POSITIONERS PV-2369A, B, C TV-2370A, B, C TV-2370D, E, F TV-2370G, H, J TR A VALVE AIR POSITIONER RX TRIP TRN. A SUPPLY (TYPICAL)

P-4 P-12 553°F C-9 TRIP OPEN CONDENSER BISTABLES

1. 3 TRN. B AVAILABLE LOAD OFF P-12 VENT BYPASS REJECTOR ON 553°F

1. 4 RESET INTLK (PT-506) C-7 VENT TRN A SWITCH BOOSTER #1 RELAY VENT

1. 2 INTERLOCK OFF BYPASS ON BYPASS STEAM RESET INTLK SWITCHES TO PRESSURE FROM STEAM DUMP HEADER MAIN TRN B SWITCH MODE (MAIN STEAM CROSS TIE)

CONDENSER OP51.SYS.SD1.FG03 Page 65 of 65 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From OP51.SYS.SD1.LN, Page 13 Revision # 2 CONTROLLERS Load Rejection Controller The Load Rejection Controller is the controlling controller when the Main Turbine is above 15% power (approximately 170MWe). This controller compares Average Tave to the reference temperature generated from pressure transmitter PT-505. When Average Tave is 5ºF above the reference temperature, an output exists from the controller. The output (demand) signal can be seen on the Steam Dump Demand Meter (u-UI-500) on CB-08. The output of this controller increases 11.11% (9.43% for Unit 2) for every degree the temperature mismatch increases above 5ºF.

In order for the Main Steam Dump to Condenser Valves to open when an output exists from this controller, solenoid 4 must be energized from one of its arming signals. The arming signal would be from a load rejection(C-7). This arrangement prevents a failed Reactor Coolant System Tave instrument (or one of the instrument's inputs) from opening the Steam Dump Valves and causing an unnecessary cooldown of the Reactor Coolant System.

Plant Trip Controller The Plant Trip controller is controlling when the unit has experienced a reactor trip from an at power condition. This controller uses the Average Tave signal to overcome the no-load temperature (557ºF) preprogrammed bias to generate an output signal. The output (demand) signal can be seen on the Steam Dump Demand Meter (u-UI-500) on Control Board 08. The output of this controller increases 2.5%

(3.2% for Unit 2) for every degree the temperature mismatch increases above 0ºF. The purpose of this controller is to allow the Steam Dump System to restore Tave to 557ºF and maintain this temperature.

This aids the operators in diagnosing events that may have lead to the reactor trip.

In order for the Main Steam Dump to Condenser Valves to open when an output exists from this controller, solenoid 4 must be energized from one of its arming signals. In this case the arming signal would be the Train A P-4 signal which would result from the Reactor Trip Breaker (RTA) opening.

Steam Pressure Controller The Steam Pressure Controller is controlling when the unit is below 15% power or is shutdown. This controller compares actual Main Steam Header Pressure to the desired pressure set on the M/A station by the operator. The Main Steam Header Pressure signal is generated by PT-507.

When the signal generated by PT-507 increases one psig above the set point on the M/A station, an output signal is generated to open the Steam Dump Valves. The output of this controller increases 1%

for every 1.8 psid.

The Main Steam Dump to Condenser Valves should open when an output exists from this controller.

For this controller to generate an output, the Steam Dump Mode Selector Switch must be selected to the Steam Pressure mode. This action places the controller in the control circuit and arms the Steam Dump Valves. The only time the Steam Dump Valves would not open would be when Main Condenser vacuum and Circulating Water Pumps have not made up the C-9 Condenser Available signal.

Page 66 of 66 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 039 K1.05 Importance Rating 2.5 Main and Reheat Steam System: Knowledge of the physical connections and/or cause-effect relationships between the MRSS and the following systems: Turbine generator Proposed Question: Common 15 Given the following conditions:

• Reactor power is 57% with a power increase in progress on Unit 2.

• A failure of the 2A Moisture Separator Reheater (MSR) Drain Tank Level Control System results in the following digital alarms:

• 2RB01L901 XV05 - MSR A SHELL LEVEL HIGH.

• 2RB01L801 XV06 - MSR A SHELL LEVEL HIGH.

• 2RB01L801 XV09 - MSR A SHELL LEVEL TRIPPED.

Which ONE (1) of the following describes the actions that occur due to this failure?

Steam to the Moisture Separator Reheater is isolated and A. 2-LV-2722, MSR A SHELL DRN TK NORM LVL CTRL VALVE fails closed.

B. 2-LV-2723, MSR A SHELL DRN TK ALT LVL CTRL VALVE receives a CLOSE signal.

C. only the Main Turbine trips.

D. the Main Turbine and Reactor both trip.

Proposed Answer: D Explanation:

A. Incorrect. Plausible because the MSR Shell Drain Tank Normal Level Control Valve fails closed on a loss of power or instrument air, however, in this condition the normal level control valve would be open on high level in the Drain Tank.

B. Incorrect. Plausible because a signal is sent to the valve, however, the MSR Shell Drain Tank Alternate Level Control Valve would be open on high level in the Drain Tank C. Incorrect. Plausible because a Turbine trip is initiated, however, the Reactor will also trip above P-9 (50%).

D. Correct. Given the power level and the high MSR level, both the Reactor and Turbine will trip.

Page 67 of 67 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Technical Reference(s) ALM-4000B, Pages 42, 43, 44 & 51 Attached w/ Revision # See Comments / Reference Proposed references to be provided during examination: None Learning Objective: STATE the physical connections and EVALUATE the cause-effect relationships between the Main Steam System and the following systems, components or events:

• Extraction Steam and Heater Drains System (OP51.SYS.MR1.OB14)

Question Source: Bank # SYS.MR1.OB14-2 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam CPNPP 2007 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 4 55.43 Comments /

Reference:

From ALM-4000B, Page 44 Revision # 1 Page 68 of 68 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ALM-4000B, Page 51 Revision # 1 Page 69 of 69 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ALM-4000B, Page 42 Revision # 1 Page 70 of 70 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ALM-4000B, Page 43 Revision # 1 Page 71 of 71 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 059 G 2.2.38 Importance Rating 3.6 Main Feedwater System: Equipment Control: Knowledge of conditions and limitations in the facility license Proposed Question: Common 16 Given the following conditions:

• Annunciator 1-ALB-8A, Window 2.4 - SG 2 FWIV N2 PRESS LO is in alarm.

• The field operator reports that the Nitrogen Accumulator pressure is 2000 psig.

Which ONE (1) of the following describes the impact on the plant due to this condition?

A. As the pressure drifts lower the Steam Generator #2 Feedwater Isolation Valve could close.

B. Low nitrogen pressure can cause the Feedwater Isolation Valve closure time to exceed the surveillance stroke time.

C. The automatic closure feature is unaffected but the remote closure from the Control Room would be slow.

D. There is no impact on valve OPERABILITY because feedwater flow closes the valve once the CLOSE signal is received.

Proposed Answer: B Explanation:

A. Incorrect. Plausible because it could be thought that the valve was held open by nitrogen pressure and spring shut.

B. Correct. At 2040 psig the nitrogen pressure may not be high enough to close the isolation valve in the time required by design ( 5 seconds).

C. Incorrect. Plausible because it could be thought that the accumulator is used for manual operation and auto closure is unaffected.

D. Incorrect. Plausible because it could be thought that MSIVs are closed by steam flow pushing the valve closed.

Technical Reference(s) ALM-0081A, 1-ALB-8A, Window 2.4 Attached w/ Revision # See Comments / Reference Proposed references to be provided during examination: None Page 72 of 72 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Learning Objective: LIST and EXPLAIN the Main Feedwater System design features which provide for the trips, permissives, and interlocks associated with the following:

• Feedwater Isolation (OP51.SYS.MF1.OB11)

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 ALM-0081A, 1-ALB-8A, Window 2.4 Revision # 8 Page 73 of 73 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 061 A2.04 Importance Rating 3.4 Auxiliary Feedwater System: Ability to (a) predict the impacts of the following malfunctions or operations on the AFW; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Pump failure or improper operation Proposed Question: Common 17 Given the following conditions:

• Turbine Driven Auxiliary Feedwater Pump (TDAFWP) has tripped on overspeed.

• The crew is responding per ABN-305, Auxiliary Feedwater System Malfunction, Section 4.0, TDAFWP Malfunction.

• The cause of the TDAFWP trip was corrected and the overspeed trip linkage has been reset per ABN-305, Auxiliary Feedwater System Malfunction, Attachment 1, HV-2452 TDAFW PUMP TRIP & THROTTLE VALVE.

• The Safeguards NEO is opening 1-HV-2452, AFWPT 1-01 TRIP AND THROT VLV.

Which ONE (1) of the following:

1.) Identifies the response of the TDAFWP as 1-HV-2452 reaches the full open position?

2.) What action, if any, is required?

A. 1.) TDAFWP speed remains at 0 rpm.

2.) Place 1-SK-2452A, AFWPT SPD CTRL in RAISE to raise TDAFWP speed.

B. 1.) TDAFWP speed rises to 2000 rpm.

2.) Adjust 1-SK-2452A, AFWPT SPD CTRL as required for desired output.

C. 1.) TDAFWP speed rises to 3000 rpm.

2.) Adjust 1-SK-2452A, AFWPT SPD CTRL as required for desired output.

D. 1.) TDAFWP speed rises to 4075 rpm.

2.) Place 1-SK-2452A, AFWPT SPD CTRL in LOWER to lower TDAFWP speed.

Proposed Answer: B Page 74 of 74 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible because 1-SK-2452A, AFWPT SPD CTRL is positioned to 0% output, however, speed will increase to a minimum speed of 2000 rpm independent of the controller position.

B. Correct. As the AFWPT Trip and Throttle Valve is opened, turbine speed will increase to 2000 rpm.

At this point the operator will control AFWPT speed as required using 1-SK-2452A, AFWPT SPD CTRL.

C. Incorrect. Plausible because 3000 rpm is the speed at which the actual overspeed trip resets on the AFWPT; however, at 0% demand on 1-SK-2452A, AFWPT SPD CTRL the turbine speed will only increase to 2000 rpm.

D. Incorrect. Plausible if thought that 1-SK-2452A, AFWPT SPD CTRL is still at 100% demand which is the normal at power position which would cause turbine speed to increase to 4075 rpm. At this point, the operator will control AFWPT speed as required using 1-SK-2452A, AFWPT SPD CTRL.

Technical Reference(s) ABN-305, Step 4.3.2 RNO Attached w/ Revision # See SOP-304A, Step 5.5.4.f thru j Comments / Reference Proposed references to be provided during examination: None Learning Objective: ANALYZE the effects of a loss of electrical signal or instrument air to the TDAFW pump governor in terms of speed. (OP51.SYS.AF1.OB14)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7, 8, 10 55.43 Page 75 of 75 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ABN-305, Step 4.3.2 RNO Revision # 6 Page 76 of 76 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SOP-304A, Step 5.5.4.f thru j Revision # 16 Page 77 of 77 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 061 K2.02 Importance Rating 3.7 Auxiliary Feedwater System: Knowledge of bus power supplies to the following: AFW electric drive pumps Proposed Question: Common 18 Given the following conditions:

• Unit 1 is in MODE 3 with the Shutdown Rods withdrawn.

• Steam Generator levels are being maintained between 65% and 70% using the Motor Driven Auxiliary Feed Water Pumps.

• An electrical perturbation results in an 86-2 Lockout Relay actuation on 6.9 KV Safeguards Bus 1EA1.

• Emergency Diesel Generator 1-02 is tagged out for fuel injector leak repair.

Assuming NO operator actions, which ONE (1) of the following describes the status of the Auxiliary Feedwater Pumps?

Motor Driven Auxiliary Feedwater Pump 1-01 is __________.

Motor Driven Auxiliary Feedwater Pump 1-02 is __________.

Turbine Driven Auxiliary Feedwater Pump is __________.

A. running.

running.

running.

B. stopped.

running.

running.

C. running.

stopped.

running.

D. stopped.

running.

stopped.

Proposed Answer: A Page 78 of 78 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Explanation:

A. Correct. An 86-1 Lockout Relay causes Bus 1EA1 to deenergize, however, an 86-2 Lockout Relay causes Bus 1EA1 to reenergize when its associated diesel breaker closes. Therefore, MDAFW Pump 1-01 will have power available, MDAFW Pump 1-02 will continue to operate, and the TDAFW Pump will auto start on the Blackout Sequencer (BOS) Operator Lockout.

B. Incorrect. Plausible because an 86-1 Lockout Relay actuation would keep MDAFW Pump 1-01 from running and MDAFW Pump 1-02 will be running along with the TDAFW Pump.

C. Incorrect. Plausible because the TDAFW Pump receives an auto start signal on the BOS Operator Lockout, however, MDAFW Pump 1-02 continues to operate.

D. Incorrect. Plausible because an 86-1 Lockout Relay causes Bus 1EA1 to deenergize, however, an 86-2 Lockout Relay causes Bus 1EA1 to reenergize when its associated diesel breaker closes.

Therefore, MDAFW Pump 1-01 will have power available, MDAFW Pump 1-02 will continue to operate, and the TDAFW pump will auto start on the BOS Operator Lockout.

Technical Reference(s) ABN-602, Step 2.3.3 Note Attached w/ Revision # See ABN-602, Attachment 1 Comments / Reference Proposed references to be provided during examination: None Learning Objective: IDENTIFY the specific power supply including source of control power voltage for the Motor Driven Auxiliary Feedwater Pumps. (OP51.SYS.AF1.OB05)

Question Source: Bank #

Modified Bank # SYS.AF1.OB05-2 (Note changes or attach parent)

New Question History: Last NRC Exam CPNPP 2009 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7, 8 55.43 Page 79 of 79 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ABN-602, Step 2.3.3 Note Revision # 7 Page 80 of 80 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ABN-602, Attachment 2 Revision # 7 Page 81 of 81 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ABN-602, Attachment 2 Revision # 7 Page 82 of 82 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SYS.AF1.OB05-2 Revision # 03/09 Given the following conditions:

• Unit 1 is in MODE 3 with the Shutdown Rods withdrawn.

• Steam Generator levels are being maintained between 65% and 70% using the Motor Driven Auxiliary Feed Water (AFW) Pumps.

• An electrical perturbation results in an 86-1 Lockout Relay actuation on 6.9 KV Safeguards Bus 1EA1.

Assuming NO operator actions, which ONE (1) of the following describes the status of the Auxiliary Feedwater Pumps?

Motor Driven AFW Pump 1-01 is __________.

Motor Driven AFW Pump 1-02 is __________.

Turbine Driven Auxiliary Feedwater Pump is __________.

A. running.

running.

running.

B. stopped.

running.

running.

C. running.

stopped.

running.

D. stopped.

running.

stopped.

Page 83 of 83 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 062 K1.04 Importance Rating 3.7 AC Electrical Distribution System: Knowledge of the physical connections and/or cause-effect relationships between the AC distribution system and the following systems: Offsite power sources Proposed Question: Common 19 Given the following conditions:

• Unit 1 is operating at 35% power in a normal system alignment.

• Transformer 1UT experiences a Differential Current Relay actuation (87/1UT).

• All others systems function as designed and NO other failures occur.

Which ONE (1) of the following is the resultant electrical bus status for Unit 1?

The Non-Safeguards 6.9 kV Buses __________ the 6.9 kV Safeguards Buses are energized by __________.

A. fast transfer to Transformer 1ST; Transformer XST1 B. slow transfer to Transformer 1ST; Transformer XST1 C. fast transfer to Transformer 1ST; Transformer XST2 D. slow transfer to Transformer 1ST; Transformer XST2 Proposed Answer: C Explanation:

A. Incorrect. Plausible because a fast transfer to Transformer 1ST will occur, however, the 6.9 kV Safeguards Buses are energized by Transformer XST2.

B. Incorrect. Plausible because a transfer to Transformer 1ST will occur, however, it is a fast transfer for the 6.9 kV Non-Safeguards Buses and the 6.9 kV Safeguards Buses are energized by Transformer XST2.

C. Correct. Given the conditions listed, a fast transfer to Transformer 1ST will occur and the 6.9 kV Safeguards Buses are energized by Transformer XST2.

D. Incorrect. Plausible because the 6.9 kV Safeguards Buses are energized by Transformer XST2, however, a fast transfer to Transformer 1ST will occur.

Technical Reference(s) OP51.SYS.AC2, Pages 17, 37 & 38 Attached w/ Revision # See Comments / Reference Page 84 of 84 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Proposed references to be provided during examination: None Learning Objective: COMPARE and CONTRAST the 6.9KV bus 86-1 and 86-2 relay operation to include the effect on the preferred, alternate and emergency power supplies.

(OP51.SYS.AC2.OB11)

ANALYZE the indications and DESCRIBE the mitigation strategy for the following procedures:

• ABN-602, Response to a 6900/480V System Malfunction (OP51.SYS.AC2.OB15)

DESCRIBE any Unit Differences between the Unit 1 and Unit 2 6.9KV / 480v AC Distribution Systems. (OP51.SYS.AC2.OB17)

Page 85 of 85 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Question Source: Bank # SYS.AC2.OB15-1 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Page 86 of 86 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From OP51.SYS.AC2 Page 17 Revision # 1 SAFEGUARDS 6.9KV DISTRIBUTION SYSTEM The Safeguards 6.9KV AC Distribution Systems is equipped with preferred, alternate, and standby power sources (Figure 2).

Figure 2 - 6.9KV Safeguards Buses Preferred Power Sources The preferred power system for Unit 1 consists of a 345Kv to 6.9KV supply via transformer XST2.

Unit 2 is supplied by 138Kv to 6.9KV transformer XST1. These sources supply power to the Class 1E buses of their unit during all modes of plant operation.

Comments /

Reference:

From OP51.SYS.AC2, Page 37 Revision # 1 Fast and Slow Transfer In the event that the normal power source loses power or a fault occurs on the source causing a lockout condition, the 6.9KV buses will automatically transfer to the alternate source of power. For the Non-Safeguards buses, this could mean a Fast or Slow Transfer would occur. For the Safeguards buses, either a Slow Transfer to the alternate source or an EDG start and transfer to EDG power is available.

Fast Transfer - Non-Safety Buses Only (Figure 14)

Page 87 of 87 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From OP51.SYS.AC2, Page 38 Revision #1 Figure 14 - Non-Safeguards 6.9KV Bus 1A2-2 Breaker - Fast Transfer Referring to Figure 14, which uses bus 1A2 as an example, in order for a Fast Transfer to occur on a Non-Safeguards 6.9KV bus, the following things must occur:

Alternate power breaker handswitch must be in AUTO (CS / 1A2-2)

The preferred power breaker must be in connect and open (52be/1A2-1 and 52h/1A2-1)

Source voltage from the alternate source must be >85% (27X3 / 1ST )

Bus frequency and frequency of the alternate source must be matched within 40 degrees (25X / 1A2)

No 86 lockouts for the bus or for the alternate power source (86-1X / 1ST, 86-2X1 / 1ST, & 86 / 1A2)

Alternate power source breaker in connect (52h / 1A2-2)

All this has to occur within 0.25 seconds, or the Fast Transfer will be blocked and a Slow Transfer will occur (62-2 1A2-2 TDPU relay)

Fast Transfer applies ONLY to the Non-Safeguards 6.9KV buses. Safeguards buses DO NOT have an Automatic Fast Transfer feature. If the alternate power supply voltage or phasing is not correct or if the transfer does not take place in 0.25 seconds, then Fast Transfer will be blocked via opening respective contacts (25X or 62-2 TDO) and the Slow Transfer will occur on the bus.

Page 88 of 88 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 062 A1.01 Importance Rating 3.4 AC Electrical Distribution System: Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the AC distribution system controls including: Significance of diesel generator load limits Proposed Question: Common 20 Given the following conditions:

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

• A decision must be made regarding the operation of the Diesel Generators (DGs).

• All systems and components have operated as expected.

Which ONE (1) of the following should be performed and the basis for that action?

The Diesel Generators...

A. are stopped, when not required, ensuring adequate fuel oil availability for future DG operation.

B. are stopped, when not required, due to limits on the length of time the DGs are operating unloaded.

C. remain operating to ensure the Safeguards Buses have adequate backup power available.

D. remain operating as needed to prevent the Safeguards Buses from being overloaded.

Proposed Answer: B Explanation:

A. Incorrect. Plausible because the Diesel Generators are stopped, however, the reason is due to the limits on the length of time they can operate unloaded.

B. Correct. The Diesel Generators are stopped, ensuring their availability should power be lost.

C. Incorrect. Plausible because with single failure criteria, leaving the Diesel Generator running will ensure adequate backup power is immediately available, however, there are limits on the length of time they can operate unloaded.

D. Incorrect. Plausible if thought that the Diesel Generators automatically synchronized to the Safeguards Buses, however, they remain in standby until required due to a loss of power.

Page 89 of 89 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Technical Reference(s) EOP-1.0A, Attachment 4, Step 10 Bases Attached w/ Revision # See Comments / Reference Proposed references to be provided during examination: None Learning Objective: DISCUSS the bases for all Steps, Cautions and Notes included in EOP-1.0.

(LO21.ERG.E1A.OB105)

Question Source: Bank # SJ3.XG2.OB100-1 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 Page 90 of 90 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From EOP-1.0A, Attachment 4, Step 10 Bases Revision # 8 Page 91 of 91 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 063 A3.01 Importance Rating 2.7 DC Electrical Distribution System: Ability to monitor automatic operation of the DC electrical system, including: Meters, annunciators, dials, recorders, and indicating lights.

Proposed Question: Common 21 Given the following conditions of Unit 1 DC Safeguards Bus voltage:

• DC Bus 1ED1 is at 134 volts and 2 amps CHARGE.

• DC Bus 1ED2 is at 135 volts and 1 amp CHARGE.

• DC Bus 1ED3 is at 135 volts and 1 amp CHARGE.

• DC Bus 1ED4 is at 134 volts and 2 amps CHARGE.

After a plant transient the following is observed:

• DC Bus 1ED1 is at 123 volts and 190 amps DISCHARGE.

• DC Bus 1ED2 is at 134 volts and 2 amps CHARGE.

• DC Bus 1ED3 is at 124 volts and 70 amps DISCHARGE.

• DC Bus 1ED4 is at 135 volts and 1 amp CHARGE.

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

A loss of...

A. Motor Control Center XEB1-1 has occurred.

B. Safeguards Bus 1EA1 has occurred.

C. Motor Control Center XEB2-1 has occurred.

D. Safeguards Bus 1EA2 has occurred.

Proposed Answer: B Page 92 of 92 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible because this is a Train A Motor Control Center, however, it powers common loads as opposed to the Battery Chargers.

B. Correct. A loss of Safeguards Bus 1EA1 will deenergize 480 V Motor Control Centers 1EB1-1 and 1EB3-1 which power their respective Battery Chargers. As a result, the Battery will become the only source of power and begin to discharge as shown.

C. Incorrect. Plausible because this is a Train B Motor Control Center, however, it powers common loads as opposed to the Battery Chargers.

D. Incorrect. Plausible because two Motor Control Centers have become deenergized, however, it is the Train A Safeguards Bus that is affected as opposed to Train B.

Technical Reference(s) SOP-604A, Attachments 5 & 6 Attached w/ Revision # See Comments / Reference Proposed references to be provided during examination: None Learning Objective: DESCRIBE how a Safety Injection and Blackout signal affects the DC Electrical System. (OP51.SYS.DC1.OB07)

STATE the general power supply for the DC Electrical System (Safeguards and non-Safeguards) for the following:

• 125 Volt DC Train A Safeguards (Busses uED1 and uED3)

(OP51.SYS.DC1.OB08)

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 93 of 93 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SOP-604A, Attachment 5 Revision # 10 Page 94 of 94 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SOP-604A, Attachment 5 Revision # 10 Page 95 of 95 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SOP-604A, Attachment 5 Revision # 10 Page 96 of 96 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SOP-604A, Attachment 5 Revision # 10 Page 97 of 97 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SOP-604A, Attachment 6 Revision # 10 Page 98 of 98 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SOP-604A, Attachment 6 Revision # 10 Page 99 of 99 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 064 K6.07 Importance Rating 2.7 Emergency Diesel Generator System: Knowledge of the effect that a loss or malfunction of the following will have on the EDG system: Air receivers Proposed Question: Common 22 Given the following conditions:

• The Unit 1 Reactor was tripped due to a loss of Instrument Air.

• During recovery, a Safety Injection occurred.

• Prior to the Unit trip, Starting Air Receiver 1-02 for Diesel Generator 1-01 was isolated for maintenance.

• Diesel Generator 1-01 started and tripped on overspeed.

• After the start, the pressure in the OPERABLE Starting Air Receiver was 140 psig.

Which ONE (1) of the following describes the response of Diesel Generator 1-01 when Starting Air Receiver pressure is restored to greater than 150 psig?

Diesel Generator 1-01 will...

A. automatically start due to the Safety Injection Actuation Signal.

B. start when the Emergency Start/Stop Switch is taken to START.

C. NOT start until the overspeed trip is RESET locally at the engine.

D. NOT start until the overspeed trip is RESET with the MASTER SWITCH (RLMS).

Proposed Answer: C Page 100 of 100 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible because a Safety Injection Actuation Signal will generate an automatic start of the Diesel Generator, however, the overspeed trip is one of the few trips NOT bypassed following an SIAS.

B. Incorrect. Plausible because performing this action would start the Diesel Generator, however, it will also automatically start with an SIAS.

C. Correct. Given the conditions listed, the overspeed trip must be reset prior to starting the Diesel Generator. The overspeed and generator differential fault trips are the only two trips not bypassed upon an SIAS.

D. Incorrect. Plausible because it could be thought that the Remote Local Maintenance Switch (RLMS), when taken to the RESET position, could be used to reset the overspeed trip.

Technical Reference(s) OP51-SYS-ED1-LN, Pages 18 & 19 Attached w/ Revision # See Comments / Reference Proposed references to be provided during examination: None Learning Objective: LIST and EXPLAIN the Emergency Diesel Generator System design features which provide for the trips, permissives, and interlocks associated with the following:

• Air pressures associated with Normal and Emergency Starts.

(OP51.SYS.ED1.OB07)

Question Source: Bank # SYS.ED1.OB07-6 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7, 8 55.43 Page 101 of 101 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From OP51-SYS-ED1-LN, Pages 18 & 19 Revision # 1 The diesel generator sets can be manually started (either locally or remotely) or automatically started with either a normal or an emergency start signal. Bus undervoltage generates an automatic diesel emergency start signal. A safety injection actuation signal also generates an automatic diesel emergency start signal.

When provided with a normal start signal, the diesel generator will rotate with air until either the engine speed is greater than 200 rpm or 5 seconds elapse, whichever comes first, so long as sufficient starting air pressure is available to roll the engine. Following a normal start the diesel will run with all automatic shutdown protections available and interlocks will remain available to prevent the diesel generator from closing onto its associated 6.9 kV bus if the bus is faulted.

When provided with an emergency start signal, the diesel generator will rotate with air until either the engine speed is greater than 200 rpm or the starting air receivers' pressure decreases to less than 150 psig. So long as the emergency start signal is maintained, the diesel will run with all automatic shutdown protections disabled except for the overspeed trip and the generator differential fault trip.

Interlocks which prevent closing a diesel generator onto its associated 6.9 kV bus when the bus or the diesel generator has a phase-to-ground fault are defeated so long as the emergency start signal is maintained. Defeating this interlock allows a diesel generator to power its associated bus with a phase-to-ground fault.

Page 102 of 102 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 073 A2.02 Importance Rating 2.7 Process Radiation Monitoring System: Ability to (a) predict the impacts of the following malfunctions or operations on the PRM system; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Detector failure Proposed Question: Common 23 Given the following condition:

• The PC-11 is alarming due to OPERATE FAILURE - CHANNEL NO PULSES RECEIVED on X-RE-5251A, Auxiliary Building Low Volume Waste Monitor.

Which ONE (1) of the following describes the impact on the Auxiliary Building Drains and the required actions to mitigate the condition?

A. Auxiliary Building Drains will remain in the normal alignment and X-RE-5251A would still initiate automatic actions on high radiation. Notify I&C to investigate the loss of counts alarm on the PC-11 per ALM-3200, Digital Radiation Monitoring System.

B. Auxiliary Building Drains will automatically divert to the Co-Current Waste System.

Verify the automatic actions occurred, declare X-RE-5251A INOPERABLE, and implement the actions of the Offsite Dose Calculation Manual.

C. Auxiliary Building Drains will remain in the normal alignment and X-RE-5251A would NOT initiate automatic actions on high radiation. Dispatch an operator to locally align Auxiliary Building Drains to the Co-Current Waste System.

D. Auxiliary Building Drains will automatically be stopped and the Auxiliary Building Sump Pumps will stop. Verify the automatic actions occurred and stop waste producing activities in the affected areas until manual alignments can occur.

Proposed Answer: B Page 103 of 103 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible because it could be thought that an OPERATE FAILURE would not result in automatic initiation and the high radiation feature is still available. Also the action to have I&C investigate is somewhat correct, however, this monitor in OPERATE FAILURE will initiate the automatic actions and is a required instrument per the Offsite Dose Calculation Manual (ODCM).

B. Correct. The loss of pulses is an OPERATE FAILURE condition that results in automatic actions that must be verified. The detector is out of service and the ODCM actions apply.

C. Incorrect. Plausible because the response would be correct for some monitors, however, this monitor in OPERATE FAILURE will initiate the automatic actions and is a required instrument per the ODCM.

D. Incorrect. Plausible because it could be thought that an OPERATE FAILURE would result in these automatic actions and the specified actions make sense to prevent sumps from backing up, however, the loss of pulses is an OPERATE FAILURE condition that results in automatic actions that must be verified. The detector is out of service and the ODCM actions apply.

Technical Reference(s) ALM-3200, Pages 38, 83, & 102 Attached w/ Revision # See Comments / Reference Proposed references to be provided during examination: None Learning Objective: LIST and EXPLAIN the Digital Radiation Monitoring System design features which provide for the trips, permissives, and interlocks associated with the following monitors:

• Turbine Building Drains (OP51.SYS.RM1.OB07)

DRAW and EXPLAIN a one-line diagram of the Digital Radiation Monitoring System, similar to Figure 1 including; major components, major component interfaces, and component operating conditions for the following operating modes/conditions:

• Operate Failures (OP51.SYS.RM1.OB03)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 11, 12, 13 55.43 Page 104 of 104 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ALM-3200, Page 38 Revision # 4 Page 105 of 105 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ALM-3200, Page 83 Revision # 4 Page 106 of 106 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ALM-3200, Page 102 Revision # 4 Page 107 of 107 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 076 A3.02 Importance Rating 3.7 Service Water System: Ability to monitor automatic operation of the SWS, including: Emergency heat loads Proposed Question: Common 24 Given the following conditions:

• Unit 1 Station Service Water (SSW) was in an abnormal lineup with Station Service Water Pump (SSWP) 1-01 running and SSWP 1-02 in STANDBY.

• Subsequently SSWP 1-01 trips and SSWP 1-02 automatically starts.

• Component Cooling Water Pump (CCWP) 1-02 is out of service for preventative maintenance.

Which ONE (1) of the following describes the SSWP auto start feature and how cooling to Component Cooling Water (CCW) System heat loads will be accomplished per ABN-501, Station Service Water System Malfunction?

SSWP 1-02 auto started when A. SSWP 1-01 supply breaker tripped OPEN (electrical interlock).

Ensure CCWP 1-01 running with both Train Safeguards Loop Isolation Valves open.

B. the return header flow in Train A SSW dropped to 14,000 gpm.

Cross connect Train A SSW Unit 1 with Train A SSW Unit 2.

C. the return header flow in Train A SSW dropped to 14,000 gpm.

Cross connect Train A CCW Unit 1 with Train A CCW Unit 2.

D. the header pressure in Train A SSW dropped to 10 psig.

Ensure CCWP 1-01 running with both Train Safeguards Loop Isolation Valves open.

Proposed Answer: D Page 108 of 108 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible because it could be thought that this was an electrical interlock and the CCW System cooling is accomplished via flow though the Safeguards Isolation Valves and the Non-Safeguards header to the header being cooled by the running SSW Pump.

B. Incorrect. Plausible because it could be thought that the auto start feature was on low flow and that the cooling option would be to cross-tie Unit SSW trains.

C. Incorrect. Plausible because it could be thought that the auto start feature was on low flow and that the only cooling option would be to cross-tie Unit CCW trains.

D. Correct. The standby Service Water Pump starts on low header pressure of 10 psig and the CCW System cooling is accomplished via flow though both Safeguards Loop Isolation Valves to the header being cooled by the running SSW Pump.

Technical Reference(s) OP51.SYS.SW1.LN, Page 26 Attached w/ Revision # See ABN-501, Step 2.3.3 & 2.3.4 Note Comments / Reference Proposed references to be provided during examination: None Learning Objective: DESCRIBE the interrelationships that the following system(s) have with the Station Service Water system and ANALYZE the effect that a loss of these system(s) will have on the Station Service Water System:

• Component Cooling Water System (OP51.SYS.SW1.OB04)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

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

Reference:

From OP51.SYS.SW1.LN, Page 26 Revision # 2 2.47 The SSW pumps and the CCW pumps are train associated pumps. Both pumps receive an automatic start signal on a "BO" or "SI" signal.

2.48 A blackout will sequence the SSW pumps onto their respective buses.

2.49 The low pressure start signal will occur if PIS-4250, on the 10" safety related header, reaches 10 psig with Train B pump in service or PIS-4251, on the 10" safety related header, reaches 10 psig with Train A pump in service.

Page 109 of 109 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ABN-501, Step 2.3.3 & 2.3.4 Note Revision # 8 Page 110 of 110 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 076 K4.06 Importance Rating 2.8 Service Water System: Knowledge of SWS design feature(s) and/or interlock(s) that provide for the following: Service water train separation Proposed Question: Common 25 Given the following conditions:

• A total loss of Station Service Water (SSW) on Unit 1 is in progress.

Which ONE (1) of the following describes Station Service Water Train OPERABILITY on both Units when the Train A SSW on Unit 2 is aligned to supply Train A SSW on Unit 1?

A. One SSW Train remains OPERABLE; Train B SSW on Unit 2.

B. No SSW Trains remain OPERABLE; the cross-tie alignment affects both Trains on Unit 2.

C. Both SSW Trains on Unit 2 remain OPERABLE; manual valves and/or check valves provide Unit/Train separation.

D. Both SSW Trains on Unit 2 and Train A SSW on Unit 1 remain OPERABLE; able to provide minimum flow for post accident heat loads.

Proposed Answer: A Explanation:

A. Correct. With Unit Trains cross-tied or Unit to Unit Trains cross-tied the Trains involved in the cross-tie are considered INOPERABLE and in this case both Unit 1 Trains were already INOPERABLE prior to the cross-tie.

B. Incorrect. Plausible because it could be thought that the cross-tie alignment involved common, non-separated sections of both Trains on Unit 2.

C. Incorrect. Plausible because it could be thought that system design maintained OPERABILITY via check valves and manual valve separation of trains.

D. Incorrect. Plausible because it could be thought that the cross-tied trains remained OPERABLE based on supplying minimum required flows.

Technical Reference(s) ABN-501, Step 5.3.7 Caution Attached w/ Revision # See OP51.SYS.SW1, Figure 1 Comments / Reference Proposed references to be provided during examination: None Page 111 of 111 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Learning Objective: DISCUSS the Technical Specifications, Technical Requirements Manual, or ODCM associated with the components, parameters, and operation of the Station Service Water System, including (Tech Specs provided for Action Requirements > 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />):

• Limiting Condition(s) for Operation (OP51.SYS.SW1.OB13)

IDENTIFY the differences between Unit 1 and Unit 2 Station Service Water System components, parameters, and operation. (OP51.SYS.SW1.OB14)

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 112 of 112 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ABN-501, Step 5.3.7 Caution Revision # 8 Page 113 of 113 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From OP51.SYS.SW1, Figure 1 Revision # 2 Page 114 of 114 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 078 A4.01 Importance Rating 3.1 Instrument Air System: Ability to manually operate and/or monitor in the control room: Pressure gauges Proposed Question: Common 26 Given the following conditions:

• Unit 1 is experiencing an Instrument Air Header leak per ABN-301, Instrument Air System Malfunction.

• The Unit 1 Standby Air Compressor, 1-02 will not load.

• Air header pressure is 44 psig and lowering.

• The in-service Letdown Orifice Isolation Valves are showing dual indication and Letdown flow is lowering.

• 1-FCV-0121, CCP 1-01/1-02 CHRG FLO CTRL VLV demand is lowering and Charging flow as indicated on 1-FI-121A, CHRG FLO is rising with no operator action.

• Steps to align the Common Instrument Air Compressor are being performed.

Which ONE (1) of the following actions would be the highest priority action to take for these conditions?

A. Isolate Letdown and perform actions per ABN-105, Chemical and Volume Control System Malfunction.

B. Attempt to restore Instrument Air and at a header pressure of 30 psig, trip the Reactor.

C. Manually trip the Reactor and go to EOP 0.0A, Reactor Trip or Safety Injection.

D. Cross-connect Unit 1 and Unit 2 Instrument Air Systems per SOP-509A, Instrument Air System Operation.

Proposed Answer: C Page 1 of 1 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible because it could be thought that loss of Letdown control would dictate isolation and performing the actions of ABN-105, Chemical and Volume Control System Malfunction, however, loss of control is occurring and the Reactor should be tripped. Also if letdown isolation was dictated it would be done per ABN-301, Instrument Air System Malfunction.

B. Incorrect. Plausible because it could be thought that the only trip criteria was a low pressure condition, however, the setpoint to trip is 35 psig.

C. Correct. Trip criteria are if pressure drops to 35 psig or if control of systems is lost.

D. Incorrect. Plausible because it is an action in the ABN and this action would be correct at this time, however, the trip criteria of loss of control of systems is a higher priority.

Technical Reference(s) ABN-301, Steps 2.3.3 & 2.3.5 Attached w/ Revision # See Comments / Reference Proposed references to be provided during examination: None Learning Objective: ANALYZE the indications and DESCRIBE the mitigation strategy and major steps taken relative to the Instrument Air system, both initial and subsequent, for:

• ABN-301, Instrument Air System Malfunction (OP51.SYS.IA1.OB14)

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 2 of 2 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ABN-301, Step 2.3.5 Revision # 11 Page 3 of 3 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ABN-301, Step 2.3.3 Revision # 11 Page 4 of 4 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 103 K3.01 Importance Rating 3.3 Containment System: Knowledge of the effect that a loss or malfunction of the containment system will have on the following:

Loss of containment integrity under shutdown conditions Proposed Question: Common 27 Given the following conditions:

• Unit 1 is in MODE 6 with Core Reload in progress.

• Steam Generator 1-01 is being drained completely to Containment using the associated Atmospheric Relief Valve as the vent path.

• Steam Generator Hydrazine concentration is 450 ppm.

Which ONE (1) of the following actions is required prior to draining the Steam Generator into Containment?

A. Ensure the drain hose is routed to the Containment Building Sump.

B. Manually start an associated Containment Building Sump Pump.

C. Suspend fuel movement or initiate a LCOAR for the open penetration.

D. Establish a 10 psig nitrogen overpressure on Steam Generator 1-01.

Proposed Answer: C Explanation:

A. Incorrect. Plausible because with a hydrazine concentration this high it requires a special alignment for draining via this path, however, the drain would be aligned to the Containment Component Cooling Water Drain Tank.

B. Incorrect. Plausible because it could be thought that the Containment Building Sump Pump should be started prior to initiating the drain, however, the procedure only ensures the Containment Building Sump Pump is in AUTO.

C. Correct. With the Atmospheric Relief Valve (ARV) vent path already open, there will be a vent path out of Containment through the drain valve, Steam Generator and ARV which requires stopping fuel movement or initiate an LCOAR for the open penetration.

D. Incorrect. Plausible because nitrogen overpressure or SG ARV vent path are both acceptable draining methods, however, the stem says to use the SG ARV vent path. May select this distractor due to Containment Integrity considerations.

Technical Reference(s) SOP-312A, Step 5.3.P.2 Attached w/ Revision # See Comments / Reference Page 5 of 5 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Proposed references to be provided during examination: None Learning Objective: DESCRIBE how manual valve alignments or removal of pipe caps on Containment penetration drains, vents, or test connections can result in a loss of Containment penetration isolation. (OP51.SYS.CY1.OB12)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 5, 10 55.43 Page 6 of 6 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SOP-312A, Step 5.3.P.2 Revision # 15 Page 7 of 7 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 103 K1.02 Importance Rating Containment System: Knowledge of the physical connections and/or cause-effect relationships between the containment system and the following systems: Containment isolation/containment integrity Proposed Question: Common 28 Given the following conditions:

• IPO-010A, Reactor Coolant System Reduced Inventory Operation is in progress.

• The Reactor Coolant System is being drained down to a Reduced Inventory Condition for Steam Generator nozzle dam installation.

Which ONE (1) of the following is the reason for verifying the status of the Containment Ventilation Isolation Valves prior to initiating a Reactor Coolant System drain down per IPO-010A, Reactor Coolant System Reduced Inventory Operation?

The Containment Ventilation Isolation Valves must be...

A. closed prior to initiating draining of the Reactor Coolant System.

B. capable of being closed if Residual Heat Removal cooling is lost.

C. closed prior to initiating movement of irradiated fuel in Containment.

D. capable of being closed prior to removing the Pressurizer manway.

Proposed Answer: B Explanation:

A. Incorrect. Plausible because the Containment Ventilation Isolation Valves must be closed or capable of being closed via a CVI signal or manually from the Control Room, however, the reason is loss of Residual Heat Removal cooling.

B. Correct. This is the correct condition and reason for verifying valve status.

C. Incorrect. Plausible because Technical Specification LCO 3.9.4 does require the Containment Ventilation Isolation Valves to be OPERABLE, however, they are not required to be closed.

D. Incorrect. Plausible because it could be thought that an open RCS is the basis for verifying valve status, however, the reason is loss of Residual Heat Removal cooling.

Technical Reference(s) IPO-010A, Attachment 1, Step 2.0.B Note Attached w/ Revision # See Technical Specification LCO 3.9.4 Comments / Reference Page 8 of 8 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Proposed references to be provided during examination: None Learning Objective: DESCRIBE the basis for the precautions and limitations, and given major procedure steps relative to the Containment Ventilation system, PLACE them in the proper sequence for:

• IPO-010, Reactor Coolant System Reduced Inventory Operations (OP51.SYS.CL1.OB13)

Question Source: Bank # IPO.010.OB-1 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, 10 55.43 Page 9 of 9 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From IPO-010A, Attachment 1, Step 2.0.B Note Revision # 17 Page 10 of 10 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From Technical Specification LCO 3.9.4 Amendment # 93 Page 11 of 11 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 2 K/A # 072 K5.02 Importance Rating 2.5 Area Radiation Monitoring System: Knowledge of the operational implications of the following concepts as they apply to the ARM system: Radiation intensity changes with source distance Proposed Question: Common 29 Given the following conditions:

• 1-RE-6259, FILT STOR AREA 873 Radiation Monitor is indicating 20 mrem/hr.

• 1-RE-6259 is physically located 20 feet away from a filter that is responsible for the indication.

• An operator must hang a clearance on a valve that is located 10 feet from the radioactive filter.

• The filter is considered a point source.

Which ONE (1) of the following is the dose rate in the area the operator will be hanging the clearance?

A. 5 mrem/hr B. 20 mrem/hr C. 40 mrem/hr D. 80 mrem/hr Proposed Answer: D Explanation:

A. Incorrect. Plausible if thought I1 x (D2)2 = I2 x (D1)2 20 x 102 = I2 x 202 = 5 mrem/hr.

B. Incorrect. Plausible if thought that the dose at 10 feet equaled the dose at the filter.

C. Incorrect. Plausible if thought I1 x (D1) = I2 x (D2) 20 x 20 = I2 x 10 = 40 mrem/hr.

D. Correct. I1 x (D1)2 = I2 x (D2)2 20 x 202 = I2 x 102 = 80 mrem/hr.

Technical Reference(s) GFE.RR4.LN, Page 31 Attached w/ Revision # See Comments / Reference Proposed references to be provided during examination: None Page 12 of 12 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Learning Objective: STATE the inverse square law. (LO21.NEO.RR4.OB111)

CALCULATE the dose rate at varying distances from point sources, line sources, plane sources, and tank sources. (LO21.NEO.RR4.OB113)

Question Source: Bank #

Modified Bank # GFE.RR4.OB103-1 (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 Generic Fundamentals Revision # N/A I1 x (D1)2 = I2 x (D2)2 20 x 202 = I x 102 = 80 mrem/hr Comments /

Reference:

From GFE.RR4.LN, Page 31 Revision # 0 The simplest type of radiation source emanates from a single point in space. The radiation intensity decreases with the square of the distance. The phenomena is called the inverse square law, and is defined mathematically by:

I1 x (D1)2 = I2 x (D2)2 Comments /

Reference:

From GFE.RR4.OB103-1 Revision # 0 Given the following conditions:

• 1-RE-6259, FILT STOR AREA 873 Radiation Monitor is indicating 10 mrem/hr.

• 1-RE-6259 is physically located 20 feet away from a filter that is responsible for the indication.

• An operator must hang a clearance on a valve that is located 5 feet from the radioactive filter.

• The filter is considered a point source.

Which ONE (1) of the following is the dose rate in the area the operator will be hanging the clearance?

A. 20 mrem/hr B. 40 mrem/hr C. 80 mrem/hr D. 160 mrem/hr Page 13 of 13 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 2 K/A # 011 K2.02 Importance Rating 3.1 Pressurizer Level Control System: Knowledge of bus power supplies to the following: PZR heaters Proposed Question: Common 30 Which ONE (1) of the following components is powered from Safeguards 480 VAC Bus 1EB1?

Pressurizer...

A. Heater Control Group C.

B. Power Operated Relief Valve 1-PCV-455A.

C. Power Operated Relief Isolation Valve 1-8000A.

D. Vent Valve 1-HV-3610.

Proposed Answer: A Explanation:

A. Correct. PZR Heater Group C is powered from 1EB1.

B. Incorrect. Plausible because the valve is used to prevent opening of a PRZR Safety and has an AUTO open function on RCS pressure, however, the valve is powered from 1ED1-1 DC Switchgear.

C. Incorrect. Plausible because the valve is used to isolate a leaking or stuck open PORV, however, the valve is powered from 1EB3-2 Motor Control Center.

D. Incorrect. Plausible because the valve is used to reduce RCS pressure when both PORVs are not available in FRH-0.1A for Bleed and Feed operations, however, the valve is powered from 1ED2-1 DC Switchgear.

Technical Reference(s) OP51.SYS.AC1.LN Attached w/ Revision # See Comments / Reference Proposed references to be provided during examination: None Learning Objective: DRAW and EXPLAIN a one-line diagram of the AC Distribution System including breakers, power supplies, transformers and major loads to include:

• Plant Support System (OP51.SYS.AC1.OB02)

Page 14 of 14 Rev e

ES-401 CPNPP Mar 2010 NRC 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 Power Supply/Breaker List for uEB-1 Revision # 0 XEB1-1 (E1-0010/A) 1BR CPX-EPTRNT-44, XEC3-3 and SIG Panel E1-0024/02A; E1-0046: E1-1426/05 NOTE: Basically, this trashes the PC-11 and numerous Rad Monitors. Affects gatronics 1C CPX-ELTRET-05, CONTROL BLDG LTG TRANSFORMER, (EAB9 & ECB1)

E1-0901 sh 01, OUTAGE POWER 1F CPX-RMTRET-03, CONTROL ROOM VENT NORTH INTAKE RAD MONITOR, X-RE-5895A, TECH SPEC 3.3.3.1, OUTAGE POWER 1J CPX-VAFNID-01, CR Exhaust Fan X-01, E1-0035 sh 1M CPX-VAFNID-03, CR Kit/Toil. Exhaust Fan X-03, E1-0035 sh 2BL CPX-VAFUPK-21, CONTROL ROOM EMER PRESSURIZATION HEATER, TECH SPEC 3.7.10 E1-0035 sh 2BR CPX-VAACCR-02, CRAC X-02 Heater, E1-0035 sh 2J CPX-VAFNCB-05, CR Pressurzation Fan X-05, TECH SPEC 3.7.10 E1-0035 sh .

2M MCC & MOTOR SPACE HEATERS 3B CPX-VAFUPK-01, PRIMARY PLANT ESF HEATER X-01, TECH SPEC 3.7.12 E1-0036 sh HANG TS CLEARANCE ON FAN HANDSWITCH 3G CPX-VAFNCB-23, CR Emerg. Pzr. Fan X-23, TECH SPEC 3.7.10 3M CPX-VAFNAV-27, AB Equipment Room Exhaust Fan X-27 Page 15 of 15 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 2 K/A # 056 K1.03 Importance Rating 2.6 Condensate System: Knowledge of the physical connections and/or cause-effect relationships between the Condensate System and the following systems: MFW Proposed Question: Common 31 Given the following conditions:

• Unit 1 is operating at 100% power.

• The Condensate Polisher is in service.

• 1-FV-2239, Condensate Pump Recirculation Valve air supply line breaks.

Which ONE (1) of the following describes the sequence of events that occurs in the Feed and Condensate Systems as a result of this failure?

The Condensate Recirculation Valve will fail A. CLOSED and if Condensate Polisher flow drops to < 5000 gpm the Condensate Polisher Bypass Valve, 1-PV-2242 will start to throttle open to maintain proper differential pressure across the Polisher Beds.

B. OPEN and if Main Feedwater Pump suction pressure lowers to 280 psig 1-PV-2242, Condensate Polisher Bypass Valve will throttle closed.

C. OPEN and if Main Feedwater Pump suction pressure lowers to 250 psig 1-PV-2286, Low Pressure Heater Bypass Valve will open.

D. CLOSED and if total Condensate System flow drops to < 6000 gpm with the Condensate Recirc Valve, 1-FV-2239 closed the Condensate Reject Valve, 1-LV-2211/12 will open for Condensate Pump protection.

Proposed Answer: C Page 16 of 16 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible because it could be thought that the failure position would be closed and the Polisher Bypass Valve does maintain differential pressure across the Condensate Polisher.

B. Incorrect. Plausible because the failure position is correct and it could be thought that the Polisher Bypass Valve would close to supply more Condensate flow, however, the Polisher Bypass Valve opens.

C. Correct. The Condensate Recirc Valve will fail open and if Main Feedwater Pump suction pressure drops to 250 psig then the Low Pressure Heater Bypass Valve opens to supply a greater flow to the Main Feedwater Pump suction.

D. Incorrect. Plausible because it could be thought that the Condensate Recirculation Valve failed closed and the Condensate Reject Valve would open to protect the Condensate Pump in low flow conditions.

Technical Reference(s) OP51.SYS.CO1, Pages 34, 35, & 38 Attached w/ Revision # See Comments / Reference Proposed references to be provided during examination: None Learning Objective: STATE the functions, operation and interlocks of the following Condensate System components:

• Condensate Pump Recirculation Valve (OP51.SYS.CO1.OB05)

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 OP51.SYS.CO1, Page 34 Revision # 1 FAILURE MODES FV-2239 fails open in the event of a loss of air or power to the solenoid. On a loss of output signal from the positioner, the valve will fail closed as long as instrument air is maintained to the positioner.

Page 17 of 17 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From OP51.SYS.CO1, Page 35 Revision # 1 REMOTE The MCB control for this valve consists of 3-position CLOSE/AUTO/OPEN, spring return to auto, handswitch (Figure 8). In the OPEN position the valve will open and in the CLOSE position the valve will close.

Figure 8 - PV-2286 Logic Diagram In the AUTO position the valve will open when turbine load is greater than 15% (sensed by generator MW output) and any 2 of 3 MFP suction header pressure switches (u-PS-2286A, E, F) decrease to 250 psi.

Page 18 of 18 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From OP51.SYS.CO1, Page 38 Revision # 1 Figure 9 - PV-2242 Logic Diagram The CP System controls consist of a Pressure Indicating Controller (PIC) that provides for manual and automatic operation of u-PV-2242. The PIC is accessed via the System Overview Control Screen Page 19 of 19 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 2 K/A # 071 A1.06 Importance Rating 2.5 Waste Gas Disposal System: Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the Waste Gas Disposal System controls including: Ventilation system Proposed Question: Common 32 Given the following conditions:

• Unit 1 is operating at 100% power.

• A Waste Gas Release of Gas Decay Tank #1 is in progress.

• A loss of MCC XEB1-2 has resulted in the loss of four of the Primary Plant Ventilation Exhaust Fans.

• X-RE-5701, Waste Gas Radiation Monitor has just gone into ALARM.

Which ONE (1) of the following describes the proper actions to take in this situation?

A. Ensure X-HS-0014, Waste Gas Discharge Control Valve closed and then terminate the release lineup.

B. Suspend the release by closing X-HS-0014, Waste Gas Discharge Control Valve.

Reinitiate the release when ventilation is restored.

C. Reduce the release flowrate by lowering the setting on X-HC-0014, GWPS Waste Gas Discharge Pressure Controller until the alarm is clear.

D. Start additional Primary Plant Ventilation Exhaust Fans per SOP-816A, Primary Plant Ventilation System until the alarm clears.

Proposed Answer: A Explanation:

A. Correct. The radiation alarm would automatically close X-HS-0014, Waste Gas Discharge Control Valve and then terminating the release would be the correct action.

B. Incorrect. Plausible because it could be thought that suspending the release would be appropriate; however, there are no provisions for suspending and then reinitiating a Waste Gas release.

C. Incorrect. For an ALERT alarm, the required action per RWS-201, Gaseous Waste Processing System is to reduce the release rate until the alarm is clear; however, since this is not defined the release is terminated.

D. Incorrect. Plausible because it could be thought that restoring ventilation dilution flow would be the appropriate action, however, the required action for an alarm is to terminate the release.

Page 20 of 20 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Technical Reference(s) RWS-201, Step.5.4.5.M.5 Attached w/ Revision # See ALM-3200, Attachment 3 Comments / Reference Proposed references to be provided during examination: None Learning Objective: ANALYZE the indications and DESCRIBE the mitigation strategy and major steps taken relative to the Digital Radiation Monitoring system, both initial and subsequent, for:

• ALM-3200, Alarm Procedure DRMS (OP51.SYS.RM1.OB13)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 11, 12 55.43 Page 21 of 21 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From RWS-201, Step.5.4.5.M.5 Revision # 19 Page 22 of 22 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ALM-3200, Attachment 3 Revision # 4 Page 23 of 23 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 2 K/A # 075 A2.02 Importance Rating 2.5 Circulating Water System: Ability to (a) predict the impacts of the following malfunctions or operations on the circulating water system; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Loss of circulating water pumps Proposed Question: Common 33 Given the following conditions:

• Unit 2 is at 20% power.

• The Perimeter NEO reports all four (4) Circulating Water Pumps are cavitating badly due to debris in front of the Trash Racks.

• Annunciator 2-ALB-9A, Window 1.11, CWS LOC PNL TRBL is in alarm.

• The following indications exist on CB10;

• 2-PR-2042, CNDSR A SHELL PRESS indicates 24.9 Hg and lowering.

• 2-PR-2042, CNDSR B SHELL PRESS indicates 24.7 Hg and lowering.

• All four (4) Circulating Water Pumps current indications are fluctuating between 25 and 220 amps on;

• 2-II-2800, CWP 1 MOTOR CURRENT

• 2-II-2801, CWP 2 MOTOR CURRENT

• 2-II-2802, CWP 3 MOTOR CURRENT

• 2-II-2803, CWP 4 MOTOR CURRENT Which ONE (1) of the following actions should be taken?

Trip the...

A. Turbine and use the Steam Dump Valves to control TAVE.

B. Turbine and use the Atmospheric Relief Valves to control TAVE.

C. Reactor and use the Steam Dump Valves to control TAVE.

D. Reactor and use the Atmospheric Relief Valves to control TAVE.

Proposed Answer: D Page 24 of 24 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible because tripping the Turbine is correct if power is less than 10%, however, the Main Steam Isolation Valves would be closed if no Circulating Water Pumps were operating and the Steam Dump Valves would not be available to control TAVE.

B. Incorrect. Plausible because using the Atmospheric Relief Valves is correct because the Steam Dump Valves would not be available, however, power must be less than 10%.

C. Incorrect. Plausible because tripping the Reactor is correct with power level greater than 10%,

however, ABN-304 requires closing the Main Steam Isolation Valves if no Circulating Water Pumps are operating.

D. Correct. Given the conditions listed, tripping the Reactor and using the Atmospheric Relief Valves to control TAVE are the required actions.

Technical Reference(s) ABN-304, Step 2.3.1 Attached w/ Revision # See Comments / Reference Proposed references to be provided during examination: None Learning Objective: DESCRIBE the following procedures which govern the operation of the Circulating Water System, including significant prerequisites, precautions, and notes associated with each operating procedure which are required to be considered by either Licensed Operators or Non-Licensed Operators:

• ABN-304, Main Condenser and Circulating Water System Malfunction (OP51.SYS.CW1.OB11)

Question Source: Bank # SYS.CW1.OB11-13 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 5, 10 55.43 Page 25 of 25 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ABN-304, Step 2.3.1 Revision # 8 Page 26 of 26 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 2 K/A # 001 K4.03 Importance Rating 3.5 Control Rod Drive System: Knowledge of CRDS design feature(s) and or interlock(s) which provide for the following: Rod control logic Proposed Question: Common 34 Given the following conditions with Unit 1 at 100% power:

• Control Rod F2 in Bank B, Group 1 drops to the bottom of the core.

• When rod realignment is begun for Control Rod F2, Annunciator 1-ALB-6D, Window 1.6 - CONTROL ROD CTRL URGENT FAIL goes into alarm.

Which ONE (1) of the following is the cause of the alarm condition?

A. The Power Cabinet has an Urgent Alarm because it can no longer supply the demanded current to any rods in Control Bank B, Group 1.

B. The Logic Cabinet has an Urgent Alarm because the Slave Cycler for Control Bank B has been deenergized.

C. The Power Cabinet has an Urgent Alarm because it can no longer supply the demanded current to any rods in Control Bank B, Group 2.

D. The Logic Cabinet has an Urgent Alarm because the Pulser for Control Bank B has been deenergized.

Proposed Answer: C Explanation:

A. Incorrect. Plausible because it could be thought that the affected group would not receive demanded current, however, the other Control Rod Group in the affected Bank is impacted.

B. Incorrect. Plausible because it could be thought that the Slave Cycler is deenergized, however, the Logic Cabinet inhibits overlapped Control Rod motion in AUTO or MANUAL.

C. Correct. Given the listed alarm, the other Control Bank B Group Step Counter is prohibited from operating and a reduced current on the movable and stationary grippers is applied.

D. Incorrect. Plausible because it could be thought that the Pulser is deenergized, however, the Logic Cabinet inhibits overlapped Control Rod motion in AUTO or MANUAL.

Technical Reference(s) ALM-0064A, 1-ALB-6D, Window 1.6 Attached w/ Revision # See ABN-712, Step 3.3.15 Note Comments / Reference Page 27 of 27 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Proposed references to be provided during examination: None Learning Objective: LIST the inputs and automatic actions for the following alarms:

• URGENT alarm in a Power Cabinet (OP51.SYS.CR1.OB13)

ANALYZE the indications and DESCRIBE the mitigation strategy and major steps taken relative to the Rod Control System, both initial and subsequent, for:

• ABN-712, Rod Control System Malfunction (OP51.SYS.CR1.OB15)

Question Source: Bank # SYS.CR1.OB401-2 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 28 of 28 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ALM-0064A, 1-ALB-6D, Window 1.6 Revision # 6 Page 29 of 29 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ABN-712, Step 3.3.15 Note Revision # 10 Page 30 of 30 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 2 K/A # 055 K3.01 Importance Rating 2.5 Condenser Air Removal System: Knowledge of the effect of a loss or malfunction of the CARS will have on the following:

Main condenser Proposed Question: Common 35 Given the following condition with Unit 1 operating at 100% power:

• The Seal Water Pump associated with the running Condenser Vacuum Pump loses suction due to low Seal Tank level.

Which ONE (1) of the following describes the FIRST indication that would alert the Control Room Operators to this malfunction and the AUTOMATIC action that should occur?

A. Lowering Main Condenser vacuum and automatic start of the standby Condenser Vacuum Pump at 24 inches Hg.

B. Annunciator 1-ALB-9A, Window 1.12 - CNDSR ANY VAC PMP TRIP causing an automatic start of the standby Condenser Vacuum Pump.

C. Annunciator 1-ALB-9B, Window 4.3 - CNDSR VAC PMP ANY SUCT PRESS LO causing an automatic start of the standby Condenser Vacuum Pump.

D. Lowering Seal Water Pump discharge pressure and automatic start of the standby Condenser Vacuum Pump at 24 inches Hg.

Proposed Answer: A Explanation:

A. Correct. There are no alarms for this condition and without Condenser Vacuum Pump sealing, Condenser vacuum will lower until the standby Condenser Vacuum Pump starts at 24 Hg.

B. Incorrect. Plausible because it could be thought that an auto trip existed for this condition and that the standby Condenser Vacuum Pump would start on the trip, however, there is no alarm or control function for the Seal Water portion of the system.

C. Incorrect. Plausible because it could be thought that the loss of sealing would result in a low suction pressure condition with an immediate start of the standby Condenser Vacuum Pump, however, the suction line is tied to the Condenser and will change as Condenser Vacuum changes with a setpoint 22 inches Hg.

D. Incorrect. Plausible because it could be thought that there was an indication for this serious condition and the AUTO start of the standby Condenser Vacuum Pump at 24 Hg lowering would be the correct system response.

Page 31 of 31 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Technical Reference(s) OP51.SYS.CV1.LN, Page 16 Attached w/ Revision # See ALM-0091A, 1-ALB-9A, Window 1.12 Comments / Reference ALM-0092A, 1-ALB-9B, Window 4.3 ABN-304, Section 3.2 Proposed references to be provided during examination: None Learning Objective: LIST and EXPLAIN the Condenser Vacuum and Water Box Priming System design features, which provide for the trips, permissives and interlocks associated with the following:

• CEV Pumps (OP51.SYS.CV1.OB10)

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 OP51.SYS.CV1.LN, Page 16 Revision # 4 INSTRUMENTS AND CONTROLS FOR THE CONDENSER VACUUM SYSTEM These instruments and controls facilitate system operation and provide indications, alarms, and controls locally and in the Control Room. Requirements include: Monitoring vacuum in the CEV pump suction lines, automatically starting the standby CEV pump on increased suction pressure, controlling seal water makeup The following components are located inside the local control box mounted on each CEV Unit skid:

• u-PS-2970A/B/C Lo Vacuum Switch (suction valve & standby pump control)

• u-PS-2971A/B/C Lo Differential Vacuum Switch (suction valve control)

• Air solenoid for u-HV-2956/2957/2958, CEV pump suction valve

• u-PS-2972A/B/C Lo Vacuum Switch (spray water control)

CONDENSER VACUUM SYSTEM CEV Pumps: (FIG. 7&8) The CEV Pumps are controlled from the Main Control Board (CB-10).

Each pump has a three-position switch for ON-OFF- STANDBY. A pump will start if its control switch is turned to ON. A pump in STANDBY will automatically start on low vacuum (< 24" Hg) in the suction header (u-PS-2970A/B/C). The suction valve for each pump is interlocked with the pump control circuit such that the valve opens on a pump start if differential pressure across the valve is less than 1" Hg vacuum (u-PS-2971A/B/C) and closes on a pump stop signal.

Page 32 of 32 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ALM-0091A, 1-ALB-9A, Window 1.12 Revision # 8 Comments /

Reference:

From ALM-0092A, 1-ALB-9B, Window 4.3 Revision # 9 Comments /

Reference:

From ABN-304, Section 3.2 Revision # 8 Page 33 of 33 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 2 K/A # 035 A2.02 Importance Rating 4.2 Steam Generator System: Ability to (a) predict the impacts of the following malfunctions or operations on the SG; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Reactor trip/turbine trip Proposed Question: Common 36 Given the following conditions:

• Unit 1 was at 100% power when a manual Reactor trip was initiated.

• Reactor Trip Breakers did not open, Reactor Power is 80% and neutron flux is slowly lowering.

• Feedwater Isolation has occurred.

Which ONE (1) of the following:

1.) Identifies the primary operational concern?

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

A. 1.) Steam Generator water inventory.

2.) Enter FRS-0.1A, Response to Nuclear Power Generation/ATWT and trip the Turbine within 30 seconds.

B. 1.) Steam Generator water inventory.

2.) Enter FRH-0.1A, Response to Loss of Secondary Heat Sink and stop all Reactor Coolant Pumps.

C. 1.) Reactor Coolant System inventory.

2.) Enter FRS-0.1A, Response to Nuclear Power Generation/ATWT and trip the Turbine within 30 seconds.

D. 1.) Reactor Coolant System inventory.

2.) Enter FRH-0.1A, Response to Loss of Secondary Heat Sink and stop all Reactor Coolant Pumps.

Proposed Answer: A Page 34 of 34 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Explanation:

A. Correct. Given the conditions listed, Steam Generator water inventory is the primary operational concern. Entry into FRS-0.1A is the correct response for this condition.

B. Incorrect. Plausible because loss of Steam Generator water inventory is the primary operational concern, however, this is the incorrect procedure for the conditions listed.

C. Incorrect. Plausible because the procedure entry is correct, however, the Subcriticality Critical Safety Function takes precedence over the Heat Sink Critical Safety Function.

D. Incorrect. Plausible because it could be thought that with a Loss of Feedwater, entry into FRH-0.1A was the correct procedure, however, the Subcriticality Critical Safety Function takes precedence over the Heat Sink Critical Safety Function.

Technical Reference(s) FRS-0.1A, Attachment 2 Attached w/ Revision # See EOP-0.0A, Step 1 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 FRS-0.1A/B, Response to Nuclear Generation/ATWT. (LO21.FRG.FS1.OB04)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 5, 10 55.43 Page 35 of 35 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From FRS-0.1A, Attachment 2 Revision # 8 Page 36 of 36 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From EOP-0.0A, Step 1 Revision # 8 Page 37 of 37 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 2 K/A # 017 K6.01 Importance Rating 2.7 Incore Temperature Monitoring System: Knowledge of the effect of a loss or malfunction of the following ITM system components: Sensors and detectors Proposed Question: Common 37 Given the following condition with Unit 1 operating at 100% power:

• All four (4) Pressurizer pressure channels indicate 2235 psig and stable.

• All four (4) TAVE channels indicate 585ºF and stable.

• Core Exit Thermocouple T7401A (TRAIN A) has an open circuit.

• Core Exit Thermocouple T7435A (TRAIN A) indicates 640ºF.

• Core Exit Thermocouple T7444A (TRAIN A) indicates 639ºF.

• Core Exit Thermocouple T7404A (TRAIN A) indicates 628ºF.

• All remaining TRAIN A Core Exit Thermocouples read between 638ºF and 629ºF.

Which ONE (1) of the following describes the Core Cooling Monitoring System indication resulting from the failure of Core Exit Thermocouple T7401A?

Train A...

A. CORE EXIT TEMP, 1-TI-3611-2 (on CB05) will indicate 2300ºF until Core Exit Thermocouple T7401A is manually removed from scan.

B. RCS SAT MARGIN, 1-TI-3611-1 (on CB05) will indicate 25ºF Subcooled with no manual action required.

C. RCS SAT MARGIN, 1-TI-3611-1 (on CB05) will indicate 300ºF Superheat until Core Exit Thermocouple T7401A is manually removed from scan.

D. CORE EXIT TEMP, 1-TI-3611-2 (on CB05) will indicate 640ºF with no manual action required.

Proposed Answer: D Page 38 of 38 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible if thought that an open circuit caused the CET indication to fail high, however, the indication would fail low and the display would show the next highest Core Exit Thermocouple (T7435A).

B. Incorrect. Plausible if thought that RCS SAT MARGIN is based on lowest reading Train A CET (T7404A), however, it is based on the highest.

C. Incorrect. Plausible if thought that an open circuit caused the CET indication to fail high, however, the indication would fail low and the Core Cooling Monitor microprocessor will automatically discard the invalid signal and the RCS SAT MARGIN indication will be calculated using the next highest reading CET (T7435A).

D. Correct. The Core Cooling Monitor microprocessor monitors for failed detectors and only allows valid outputs to be used on the CORE EXIT TEMP and RCS SAT MARGIN indications.

Technical Reference(s) OP51.SYS.RC3.LN, Page 14 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. (OP51.SYS.RC3.OB03)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Page 39 of 39 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From OP51.SYS.RC3.LN, Page 14 Revision # 0 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 40 of 40 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 2 K/A # 014 G 2.2.38 Importance Rating 3.6 Rod Position Indication System: Equipment Control: Knowledge of conditions and limitations in the facility license Proposed Question: Common 38 Given the following conditions:

• The Digital Rod Position Indication (DRPI) Accuracy Mode Switch (S106) is in the A + B position.

• Indicated Control Bank D Rod positions on the DRPI display bezel on CB07 are 12 steps.

With NO DRPI failure present, which ONE (1) of the following states what indicated Control Rod Position will be on the DRPI display bezel on CB07 when Data A only or Data B only is selected via the Accuracy Mode Switch (S106)?

Switch in Data A only Switch in Data B only A. 6 steps 6 steps B. 6 steps 18 steps C. 12 steps 12 steps D. 18 steps 6 steps Proposed Answer: C Explanation:

A. Incorrect. Plausible because of the misconception surrounding half-accuracy. In this condition it could be thought that with only Data A or Data B only switches the indication would be 1/2 of actual.

B. Incorrect. Plausible if thought that half-accuracy added to one indication and subtracted from the other.

C. Correct. With no failures present, DRPI indication is independent of the Data A or Data B switch position.

D. Incorrect. Plausible if thought that half-accuracy added to one indication and subtracted from the other.

Page 41 of 41 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Technical Reference(s) ABN-712, Step 4.3.1 Note Attached w/ Revision # See ABN-712, Attachment 2 Comments / Reference OP51.SYS.RI1.LN, Page 12 OP51.SYS.RI1.LN, Figure 6 Proposed references to be provided during examination: None Learning Objective: DESCRIBE the DRPI system accuracy determination for the following conditions:

• A and B Data

• A Data only

• B Data only (OP51.SYS.RI1.OB11)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 6 55.43 Page 42 of 42 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ABN-712, Step 4.3.1 Note Revision # 10 Page 43 of 43 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ABN-712, Attachment 2 Revision # 10 Page 44 of 44 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ABN-712, Attachment 2 Revision # 10 Page 45 of 45 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From OP51.SYS.RI1.LN, Page 12 Revision # 0 Full and Half Accuracy Figure 6 depicts the relationship between full accuracy and half accuracy. As an example, if the drive rod enters the area that is surrounded by the coil designated A4 with both Data A and Data B available (full accuracy), DRPI would indicate 30 steps. In this example, coils A4 and B3 would be penetrated by the drive rod. Data A and Data B each will indicate the highest coil penetrated by the drive rod. Actual position of the rod can be determined to be greater than 27 steps (the top of A4 coil) and less than 33 steps (the top of the B4 coil). One step is added to the error for mechanical tolerance, resulting in an accuracy of +/-4 steps. DRPI calculates rod position by adding the number of coils of each train that are penetrated above rod bottom, and multiplying by 6, the number of steps between coils. Therefore, the indication would be 3 coils (Data A) + 2 coils (Data B) = 5 coils x 6 steps per coil = 30 steps.

With a Data B failure, the accuracy would now be +4 steps, -10 steps. Data A still would be indicating 3 coils. The central control cards would replace the missing Data B by combining Data A with itself.

The output from the central control cards would be 2 times the Data A indication (6 coils), times 6 steps per coil, or 36 steps. The actual rod height could be as low as 27 steps or as high as 39 steps due to the B4 coil not indicating. Thus with the indication reading 36 steps and the actual rod height is 27 steps, there is a difference between actual and indicated of -9 steps. With the extra step added to the error for mechanical tolerance, this would create the -10 step half accuracy for a loss of Data B. If the actual rod height is 39 steps, there would a difference between indicated and actual rod height of +3 steps. With the mechanical tolerance added in, it would equal +4 steps. Therefore the half accuracy for a loss of Data B is -10, +4 steps. This same sequence may be used to explain the +10, -4 step error for a loss of Data A.

Page 46 of 46 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From OP51.SYS.RI1.LN, Figure 6 Revision # 0 COIL B21 INDICATED ROD POSITION PLACEMENT 237 (STEPS) DATA A DATA B A21 231 FULL A ONLY B ONLY B6 5 60 60 57 5 60 54 A6 51 4 48 48 B5 45 4 48 42 A5 39 3 36 36 B4 33 3 36 30 A4 27 2 24 24 B3 21 2 24 18 A3 15 1 12 12 B2 9 1 12 6

A2 0 0 3

0 0 0 ROD BOTTOM B1 A1 Page 47 of 47 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 065 AA2.08 Importance Rating 2.9 Loss of Instrument Air: Ability to determine and interpret the following as they apply to the Loss of Instrument Air: Failure modes of air-operated equipment Proposed Question: Common 39 Given the following conditions:

• The Unit 1 Turbine Driven Auxiliary Feedwater (TDAFW) Pump is operating.

• Unit 1 has lost Instrument Air pressure.

Which ONE (1) of the following states how the loss of Instrument Air pressure affects the TDAFW Pump?

A. Steam supply valves fail closed.

B. The TDAFW Pump trips.

C. Speed lowers to minimum.

D. Speed rises to maximum.

Proposed Answer: D Page 48 of 48 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible if thought that the valves would fail closed to prevent an overspeed condition, however, the steam supply valves fail open.

B. Incorrect. Plausible if thought that the TDAFW will overspeed, however, the governor continues to function.

C. Incorrect. Plausible if thought that the speed governor fails to minimum on a Loss of Instrument Air.

D. Correct. The speed governor fails to maximum on a Loss of Instrument Air.

Technical Reference(s) ABN-301, Attachment 2, Page 25 of 42 Attached w/ Revision # See OP51.SYS.AF1, Pages 21 & 25 Comments / Reference Proposed references to be provided during examination: None Learning Objective: ANALYZE the effects of a loss of electrical signal or instrument air to the TDAFW pump governor in terms of speed. (OP51.SYS.AF1.OB14)

Question Source: Bank # SYS.AF1.OB14-3 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam CPNPP 2007 Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 Page 49 of 49 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ABN-301, Attachment 2, Page 25 of 42 Revision # 11 Page 50 of 50 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From OP51.SYS.AF1, Page 21 Revision # 0 TDAFWP STEAM SUPPLY The two steam supply lines contain normally closed, pneumatic diaphragm isolation valves. These air operated valves fail open, ensuring that the turbine accelerates to design speed within 85 seconds, on loss of air supply or electrical power. Each valve is provided with a safety class air accumulator to permit the valves to be closed in the event of an instrument air failure. The accumulators are sized to close the steam supply valves and maintain them closed for 7.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />.

The main steam line to the TDAFW pump is equipped with condensate traps to remove any moisture buildup in the lines. Turbine steam is exhausted to the atmosphere through a safety related roof vent.

The turbine steam exhaust line on Unit 1 is equipped with a condensate trap to eliminate moisture buildup in the exhaust line and turbine. These traps are routed to a flash tank located in the pipe trench outside the TDAFW pump room. In both Units, selected traps are provided with level switches which provide signals to actuate annunciator window 2.6, "ANY TD AFWP D\POT LVL HI" alarm on ALB-8B on the Main Control Board. This annunciator provides indication of excessive condensation and/or moisture buildup in the steam supply line to the TDAFW Pump turbine.

Page 51 of 51 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From OP51.SYS.AF1, Page 25 Revision # 0 The action of the servo is a "push-pull" motion. The movement of the servo rod in the upwards direction causes a decrease in turbine speed, while a downward movement causes an increase in speed.

The speed at which the governor will control is determined by the force exerted by the speeder spring on the "toes" of the flyweights. Speeder spring force is determined by the position of the piston in the speed setting cylinder. The position of the piston is determined by the volume of oil trapped in the area above the piston. The direction and rate of oil flow into or out of this area is controlled by the speed setting pilot valve plunger which is mechanically linked to the bellows. If the plunger is moved downward, uncovering the upper edge of a meter port in the bushing, oil at pump discharge pressure is allowed to flow into the speed setting cylinder. This displaces the piston downward, further increasing speeder spring tension and thus increasing the speed setting. If the plunger is moved upward, uncovering the lower edge of the metering port, oil is permitted to drain from the cylinder. This allows the spring to raise the piston, decreasing speeder spring force and thus lowering the speed setting. A loss of instrument air to the governor or electrical power to the I/P controller would result in full extension of the control bellows and a "full open" signal to the differential servo, i.e., the governor would fail to the maximum speed position.

Page 52 of 52 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 008 AK3.05 Importance Rating 4.0 Pressurizer Vapor Space Accident: Knowledge of the reasons for the following responses as they apply to the Pressurizer Vapor Space Accident: ECCS termination or throttling criteria Page 53 of 53 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Proposed Question: Common 40 Given the following conditions:

• Unit 1 was at 100% power.

• Reactor Trip and Safety Injection were actuated due to rapidly lowering Reactor Coolant System pressure.

• All plant equipment actuated as designed.

• Ten (10) minutes after the trip the following indications are present:

• Reactor Coolant System pressure is 1485 psig and slowly rising.

• Pressurizer level is 100%.

• Saturation Margin is 20ºF subcooled.

• Cold Leg temperatures are 576ºF and slowly lowering.

• Reactor Vessel Level Indication System 79 inches above plate light is DARK.

• Containment pressure is 1.15 psig and slowly rising.

• Containment Recirculation Sump levels indicate 808 feet and stable.

• PC-11 GRID 4 all monitors are GREEN.

• All Steam Generator pressures are 1085-1090 psig and stable.

• All Steam Generator narrow range levels are 50-60% under operator control.

Which ONE (1) of the following:

1.) Identifies the event?

2.) What action should be taken based on the present indications?

A. 1.) Small Break Loss of Coolant Accident in Reactor Vessel Head.

2.) Transition to EOS-1.1A, Safety Injection Termination due to RCS Inventory.

B. 1.) Small Break Loss of Coolant Accident in the Pressurizer steam space.

2.) Remain in EOP-0.0A, Reactor Trip or Safety Injection due to RCS Subcooling.

C. 1.) Small Break Loss of Coolant Accident in the Pressurizer liquid space.

2.) Transition to EOS-1.1A, Safety Injection Termination due to Secondary Heat Sink.

D. 1.) Small Break Loss of Coolant Accident in RCS Loop #4 Hot Leg.

2.) Remain in EOP-0.0A, Reactor Trip or Safety Injection due to RCS Pressure.

Proposed Answer: B Page 54 of 54 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible because a bubble could exist in the Reactor Vessel Head which would cause the Pressurizer to be full, however, transitioning to EOS-1.1A based on RCS Inventory is not correct because subcooling criteria for SI Termination has not been met.

B. Correct. Given the conditions listed, a pressurizer steam space leak exists and remaining in EOP-0.0A is required. Sub cooling criteria is not met to support throttling of Safety Injection flow.

C. Incorrect. Plausible because a leak in the Pressurizer liquid space could lower portions of the Pressurizer to saturation conditions which might cause the Pressurizer to indicate full, however, transitioning to EOS-1.1A based on Secondary Heat Sink is not correct because subcooling criteria for SI Termination has not been met.

D. Incorrect. Plausible because remaining in EOP-0.0A is correct albeit for the wrong reason.

Additionally, it could be thought that a leak in the Loop #4 RCS Hot Leg could create saturation conditions that caused the Pressurizer to indicate full since that this is where the Pressurizer taps into the RCS.

Technical Reference(s) EOS-1.1A, Steps 14, 15 & 21 Attached w/ Revision # See EOS-1.1A, Attachment 1.A Comments / Reference OP51.SYS.RC1.LN, Page14 Proposed references to be provided during examination: None Learning Objective: EVALUATE plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrumentation while responding to a Reactor Trip or Safety Injection. (LO21. ERG.E0A.OB13)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

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

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From EOS-1.1A, Step 14 Revision # 8 Comments /

Reference:

From EOS-1.1A, Step 15 Revision # 8 Page 56 of 56 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From EOS-1.1A, Step 21 Revision # 8 Page 57 of 57 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From EOS-1.1A, Attachment 1.A Revision # 8 Comments /

Reference:

From OP51.SYS.RC1.LN, Page14 Revision # 1 MAJOR RCS LOOP PENETRATIONS

• Pressurizer spray lines connect to loops 1 and 4 cold legs. Spray scoops allow the pressurizer spray flow to be assisted by reactor coolant loop flow velocity.

• Chemical and Volume Control System (CVCS) charging connects to loops 1 and 4 cold legs. Each fuel cycle, charging flow alignment alternates between these penetrations.

• Cold leg injection from the Safety Injection System connects to each cold leg through a 10-inch diameter pipe. During a safety injection the discharge of the safety injection accumulators, safety injection pumps (SIPs) and residual heat removal (RHR) pumps, is directed to the RCS via this piping connection. During cold shutdown operations, RHR pump discharge is directed through this connection for decay heat removal.

• Cold leg injection from the centrifugal charging pumps (CCPs) flows to each cold leg through a 11/2-inch line during safety injection.

• Hot leg injection from RHR pumps and SIPs during the recirculation phase of safety injection.

SIPs inject to all four hot legs, and RHR pumps inject to hot legs 2 and 3. Loops 1 and 4 hot leg injection penetrations also function as the RHR suction source for shutdown plant cooling.

• The pressurizer surge line connects loop 4 hot leg to the bottom of the pressurizer.

• Process sampling lines on loop 1 and 4 hot legs have scoops extending into the coolant stream.

Page 58 of 58 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 057 AK3.01 Importance Rating 4.1 Loss of Vital AC Instrument Bus: Knowledge of the reasons for the following responses as they apply to Loss of Vital AC Instrument Bus: Actions contained in EOP for loss of vital AC instrument bus Proposed Question: Common 41 Given the following conditions:

• Unit 1 has experienced a loss of 118 VAC Protection Bus 1PC1.

• The Reactor Operator ensures the Rod Control System is in MANUAL.

Which ONE (1) of the following is the reason for placing the Rod Control System in MANUAL?

The 118 VAC Protection Bus supplies...

A. power to N-41, Power Range Nuclear Instrument. The failed channel may result in outward Control Rod movement when in AUTO.

B. the Loop 1 TAVE instrument. The failed channel may result in C-16 and outward Control Rod movement when in AUTO.

C. the Logic Cabinet of the Rod Control System. The bus loss may result in a CONTROL ROD CTRL URGENT alarm and loss of AUTO Rod Control.

D. the TREF instrument loop. The failed TREF channel may result in inward Control Rod movement when in AUTO.

Proposed Answer: D Explanation:

A. Incorrect. Plausible because power is lost to this Power Range Nuclear Instrument, however, a loss of power would generate Rod Stops that would prevent outward Rod movement when in AUTO or MANUAL.

B. Incorrect. Plausible because Loop TAVE will lower and C-16 is blocked, however, this inhibits Turbine loading only.

C. Incorrect. Plausible if thought that the Control Rod Control Urgent alarm was brought in by this failure, however, the alarm does not annunciate and does not cause a loss of AUTO Rod control.

D. Correct. The TREF instrument loop fails low which creates a high TAVE -TREF deviation which results in inward Rod motion.

Page 59 of 59 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Technical Reference(s) ABN-603, Steps 2.2.b & 2.3.3.a Attached w/ Revision # See ABN-603, Attachment 1 Comments / Reference ALM-0064A, 1-ALB-6D, Window 1.6 Proposed references to be provided during examination: None Learning Objective: ANALYZE the indications and DESCRIBE the mitigation strategy and major steps taken, both initial and subsequent, for:

• ABN-603, Loss of Protection or Instrument Bus (OP51.SYS.AC3.OB13)

Question Source: Bank # SYS.AC3.OB13 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 60 of 60 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ABN-603, Step 2.2.b Revision # 7 Page 61 of 61 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ABN-603, Step 2.3.3.a Revision # 7 Page 62 of 62 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ABN-603, Attachment 1 Revision # 7 Page 63 of 63 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ALM-0064A, 1-ALB-6D, Window 1.6 Revision # 6 Page 64 of 64 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 058 G 2.4.50 Importance Rating 4.2 Loss of DC Power: Emergency Procedures/Plan: Ability to verify system alarm setpoints and operate controls identified in the alarm response manual Proposed Question: Common 42 Given the following conditions:

• Unit 1 is operating at 75% power.

• Annunciator 1-ALB-10B, Window 1.14 - BATT CHRG BC 1ED1-1/BC 1ED1-2 TRBL is in alarm.

• Switchboard Feeder Breaker, 1ED1/2-9/BKR for Battery Charger BC 1ED1-2 is open and damaged.

Which ONE (1) of the following is an appropriate action assuming no other faults or damage?

A. Restore Battery Charger BC 1ED1-2 to OPERABLE status within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

B. Place BC 1ED1-1 in service and initiate repairs to Breaker 1ED1/2-9/BKR.

C. Ensure DC power supply to Train B Steam Dump Solenoid Valves is OPERABLE.

D. Load shed SWBD 1ED1 to ensure Safeguard Switchgear TRIP & CLOSE function.

Proposed Answer: B Explanation:

A. Incorrect. Plausible because the Battery terminal voltage must be restored within two hours, however, the Battery Charger is not required to be restored to OPERABLE status for seven days B. Correct. Given the conditions listed, this is the required action.

C. Incorrect. Plausible if thought that the Steam Dump Solenoid Valves were 1E powered, however, the valves are powered by a Non-1E source.

D. Incorrect. Plausible because this action would be performed, however, only if the output of the Battery Charger was not within required limits. See ALM-0102A, 1-ALB-10B, Window 1.14, Step 3.j.

Technical Reference(s) ALM-0102A, 1-ALB-10B, Window 1.14 Attached w/ Revision # See Technical Specification LCO 3.8.4.A Comments / Reference LO21.MCO.TA8.LN, Page 16 Proposed references to be provided during examination: None Page 65 of 65 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Learning Objective: ANALYZE the indications and DESCRIBE the mitigation strategy and major steps taken relative to the DC Electrical System, both initial and subsequent, for:

• ALM-0102 A/B, Alarm Procedure u-ALB-10B (OP51.SYS.DC1.OB16)

Question Source: Bank # SYS.DC1.OB16-4 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 66 of 66 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ALM-0102A, 1-ALB-10B-1.14 Revision # 10 Page 67 of 67 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ALM-0102A, 1-ALB-10B-1.14 Revision # 10 Page 68 of 68 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From Technical Specification LCO 3.8.4.A Amendment # 113 Page 69 of 69 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From LO21.MCO.TA8.LN, Page 16 Revision # 0 CONCLUSIONS As noted in Appendices 4A and 4B for Units 1 and 2, respectively, the analysis shows that the criteria stated earlier in this section are satisfied. For an accidental depressurization of the main steam system the DNB design limits are not exceeded. The radiological consequences of this event are not limiting.

This case is bounded by the main steam line rupture case described in Section 15.1.5. An additional case has been examined wherein the steam dump system is assumed to inadvertently actuate. No credit is taken for the proper operation of the non-Class 1E steam dump solenoid valves when the low-low TAVE (P-12) setpoint was exceeded. If a Main Steam Isolation Valve is assumed to fail to close, a non uniform cooldown of the RCS would ensue; however, the event would remain bounded by the analysis of the Steamline Break event presented in Section 15.1.5, and all event acceptance criteria would continue to be met.

Page 70 of 70 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # W/E05 EA2.2 Importance Rating 3.7 Inadequate Heat Transfer - Loss of Secondary Heat Sink: Ability to determine and interpret the following as they apply to the Loss of Secondary Heat Sink: Adherence to appropriate procedures and operation within the limitations in the facility license and amendments Proposed Question: Common 43 Which ONE (1) of the following is the operator action and procedural bases regarding Reactor Coolant Pump operation contained in FRH-0.1A, Response to Loss of Secondary Heat Sink?

A. Continue RCP operation to keep the core covered with two phase fluid.

B. Trip all RCPs to prevent major pump damage and subsequent seal failure.

C. Run at least one RCP to reduce heat input yet maintains some core flow.

D. Trip all RCPs to reduce heat input and conserve Steam Generator inventory.

Proposed Answer: D Explanation:

A. Incorrect. Plausible if thought that pumping two phase fluid during a Loss of Secondary Heat Sink would assist in cooling the core.

B. Incorrect. Plausible because all Reactor Coolant Pumps must be tripped, however, the reason is to reduce heat input and conserve Steam Generator inventory.

C. Incorrect. Plausible because keeping a Reactor Coolant Pump running would ensure flow through the core, however, the potential for RCP damage is significant and the loss of a single pump could result in core recovery due to phase separation.

D. Correct. Failure to trip RCPs exacerbates the response to a Loss of Secondary Heat Sink.

Technical Reference(s) FRH-0.1A, Attachment 4, Step 2 Bases Attached w/ Revision # See Comments / Reference Proposed references to be provided during examination: None Learning Objective: IDENTIFY the four major steps of FRH-0.1, Response of Loss of Secondary Heat Sink. (LO21.MCO.MI4.OB04)

Question Source: Bank # MCO.MI4.OB105-5 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam Page 71 of 71 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Comments /

Reference:

From FRH-0.1A, Attachment 4, Step 2 Bases Revision # 8 Page 72 of 72 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 007 EK3.01 Importance Rating Reactor Trip-Stabilization- Recovery: Knowledge of the reasons for the following responses as they apply to the reactor trip:

Actions contained in EOP for reactor trip Proposed Question: Common 44 Given the following conditions following a unit trip:

• EOP-0.0A/B, Reactor Trip or Safety Injection, Attachment 9, Post Event System Realignment is being performed.

• Component Cooling Water (CCW) Pump flow is greater than 17,500 gpm when the CCW Safeguards Loop cross-connect valves are opened.

• As a result, CCW System non-essential loads were isolated.

Which ONE (1) of the following is the reason for isolating CCW System non-essential loads?

A. Verify the CCW Heat Exchanger tubes are NOT damaged by erosion.

B. Ensure that the CCW Pump does NOT experience run out.

C. Verify CCW System maximum analyzed temperature is NOT exceeded.

D. Ensure the CCW Heat Exchanger tubes are NOT damaged by vibration.

Proposed Answer: B Explanation:

A. Incorrect. Plausible because high flow can cause erosion, however, the limiting condition is CCW Pump run out.

B. Correct. Per the CAUTION in EOP-0.0A, Attachment 9.

C. Incorrect. Plausible if thought that CCW System temperature were the limiting condition, however, the limiting condition is CCW Pump run out.

D. Incorrect. Plausible because high flow can cause vibration, however, the limiting condition is CCW Pump run out.

Technical Reference(s) EOP-0.0A, Attachment 9 Attached w/ Revision # See Comments / Reference Proposed references to be provided during examination: None Learning Objective: DETERMINE the availability of safety related equipment required to respond to a Reactor Trip or Safety Injection. (LO21. ERG.E0A.OB16)

Page 73 of 73 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Question Source: Bank # SJ1.XG1.OB104-3 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 74 of 74 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From EOP-0.0A, Attachment 9 Revision # 8 Page 75 of 75 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 038 EK1.03 Importance Rating 3.9 Steam Generator Tube Rupture: Knowledge of the operational implications of the following concepts as they apply to the SGTR: Natural circulation Proposed Question: Common 45 Given the following conditions:

• A Steam Generator Tube Rupture has occurred on Unit 1.

• EOP-3.0A, Steam Generator Tube Rupture is being performed.

• All Reactor Coolant Pumps have been tripped.

• The step to depressurize the Reactor Coolant System using the Pressurizer Power Operated Relief Valve to minimize break flow and refill the Pressurizer is being performed.

Which ONE (1) of the following is a possible result of performing this step?

A. A rapid rise in PRZR level due to voiding in the Reactor Head during Natural Circulation.

B. A rapid drop in core differential temperature as Natural Circulation is enhanced.

C. A rapid drop in ruptured SG TCOLD due to the loop stagnation during the pressure reduction.

D. A rapid rise in Containment pressure due to overpressurization of the Pressurizer Relief Tank.

Proposed Answer: A Explanation:

A. Correct. The upper head region may void during RCS depressurization if the RCPs are not running. This will result in a rapidly increasing Pressurizer level.

B. Incorrect. Plausible because the just completed cooldown prior to the depressurization would enhance Natural Circulation, however, Core T should not be affected and any voiding that does occur would tend to hinder Natural Circulation.

C. Incorrect. Plausible because a rapid drop in any stagnant loop cold leg may occur, but this is a result of Safety Injection flow into the loop and not due to any depressurization.

D. Incorrect. Plausible because the PRT may rupture, however, this is not likely to occur. Even if the PRT were to rupture, Containment conditions would only change slightly and a rapid rise in pressure would not occur.

Page 76 of 76 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Technical Reference(s) EOP-3.0A, Step 20 Attached w/ Revision # See EOP-3.0A, Attachment 6, Step 20 Bases Comments / Reference Proposed references to be provided during examination: None Learning Objective: Given a procedural step, or sequence of steps from EOP-3.0, STATE the purpose/bases for the step(s). (LO21.ERG.E3A.OB103)

Question Source: Bank # EO3.XG5.OB15-3 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam CPNPP 2005 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 Page 77 of 77 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From EOP-3.0A, Step 20 Revision # 8 Page 78 of 78 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From EOP-3.0A, Attachment 6, Step 20 Caution Bases Revision # 8 Page 79 of 79 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From EOP-3.0A, Step 20 Bases Revision # 8 Page 80 of 80 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 009 EA1.09 Importance Rating 3.6 Small break LOCA: Ability to operate and/or monitor the following as they apply to a small break LOCA: RCP Proposed Question: Common 46 Given the following conditions on Unit 1:

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

• Reactor Coolant System pressure is 1285 psig and slowly lowering.

• Highest Reactor Coolant System Cold Leg temperature is 560ºF.

• Containment pressure is 11 psig and slowly rising.

• All required Safety Systems have actuated properly.

• Reactor Coolant Pumps are running.

• The Emergency Diesels have been stopped.

• Auxiliary Feedwater total flow is 900 gpm supplied by the Motor Driven Auxiliary Feedwater pumps and the highest Steam Generator level is 38% narrow range.

• The crew is transitioning to EOP-1.0, Loss of Reactor or Secondary Coolant.

Which ONE (1) of the following is the first action that needs to be taken?

A. Throttle Auxiliary Feedwater to maintain level 43% to 60%.

B. Initiate a cooldown to restore Reactor Coolant System subcooling.

C. Check an ECCS Pump running and stop all Reactor Coolant Pumps.

D. Stop the Residual Heat Removal Pumps, place them in standby, and reset the Auto Switchover.

Proposed Answer: C Page 81 of 81 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible because it could be thought that secondary conditions warranted this action, however, this action is not directed until at least one Steam Generator is at 50% since an Adverse Containment condition exists.

B. Incorrect. Plausible because it could be thought that this would be the prudent action based on current subcooling, however, the cooldown actions are in EOP-1.0 and do not occur until Step 12.

C. Correct. This is the required action per Step 1 and Attachment 1A for Reactor Coolant System subcooling less than 25ºF (~17ºF).

D. Incorrect. Plausible because it could be thought that pressure well above the shutoff head of the RHR pumps would constitute the criteria for securing the Residual Heat Removal pumps for pump protection, however, this action requires stable or rising pressure.

Technical Reference(s) EOP-1.0A, Attachment 1A, Step 1 Attached w/ Revision # See EOP-1.0A, Attachment 1B, Steps 3, 6 & 8 Comments / Reference Proposed references to be provided during examination: Steam Tables Learning Objective: DISCUSS the bases for all Steps, Cautions and Notes included in EOP-1.0. (LO21.ERG.E1A.OB105)

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 82 of 82 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From EOP-1.0A, Attachment 1A, Step 1 Revision # 8 Page 83 of 83 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From EOP-1.0A, Attachment 1B, Steps 3, 6 & 8 Revision # 8 Page 84 of 84 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 077 AK2.06 Importance Rating 3.9 Generator Voltage and Electric Grid Disturbances: Knowledge of the interrelations between Generator Voltage and Electric Grid Disturbances and the following: Reactor power Proposed Question: Common 47 Given the following conditions:

• A Grid disturbance has resulted in a small reduction in incoming Switchyard voltage and frequency.

Which ONE (1) of the following describes the cause and effect on the plant as a result of this transient?

A. Reactor power will lower slightly due to the reduced speed of Unit motors.

B. Reactive load and Reactor power will drop slightly due to lower voltage.

C. As voltage lowers the amps drawn by equipment lowers causing Reactive load to lower and Reactor Power to rise.

D. As Grid voltage and frequency drop the Main Turbine Control Valves open causing Reactor Power to rise.

Proposed Answer: D Explanation:

A. Incorrect. Plausible because there will be a reduction in motor speeds, however, overall power will be higher as real and reactive loading is picked up by the Main Generator.

B. Incorrect. Plausible because it could be thought that the lower voltage would result in less reactive load and reduced power.

C. Incorrect. Plausible if thought that as voltage drops the amps drawn by equipment also goes down, however, a decrease in voltage should result in a rise in amperage and a resulting change in Reactive load.

D. Correct. The Main Generator will pickup real load from the grid as the grid frequency drops and will also pickup reactive load from the grid as voltage drops. The result is a rise in Reactor power.

Technical Reference(s) ABN-601, Step 9.1.b & 9.2 Attached w/ Revision # See Comments / Reference Proposed references to be provided during examination: None Page 85 of 85 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Learning Objective: DESCRIBE the function and operation of systems, components, and controls required to be operated in response to a Station Electrical malfunction.

(LO21.ABN.601.OB10)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 5 55.43 Page 86 of 86 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ABN-601, Step 9.1.b & 9.2 Revision # 10 Page 87 of 87 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 027 AK3.03 Importance Rating 3.7 Pressurizer Pressure Control Malfunction: Knowledge of the reasons for the following responses as they apply to the Pressurizer Pressure Control Malfunction: Actions contained in EOP for PZR PCS malfunction Proposed Question: Common 48 Given the following conditions on Unit 1:

• A Steam Generator Tube rupture is in progress.

• The crew is implementing EOP-3.0A, Steam Generator Tube Rupture.

• Step 19, Depressurize RCS to Minimize Break Flow and Refill PRZR is being implemented.

• Reactor Coolant Pumps were stopped per Step 1 in EOP-3.0A, Steam Generator Tube Rupture.

Which ONE (1) of the following is the preferred method, per EOP-3.0A, Steam Generator Tube Rupture for reducing Reactor Coolant System pressure and why?

A. 1.) Go to ECA -3.3A, SGTR Without Pressurizer Pressure Control to start RCPs and establish normal Pressurizer Spray.

2.) Establishing normal Pressurizer Spray flow conserves RCS inventory during depressurization.

B. 1.) Turn off all Pressurizer Heaters and open a Power Operated Relief Valve.

2.) Precludes possible failure of the Pressurizer Spray Nozzle if Auxiliary Spray flow was used without Letdown flow.

C. 1.) Initiate minimum Auxiliary Spray flow by isolating normal Spray and Charging flowpaths.

2.) Precludes creating a Large Break Loss of Coolant condition if the Power Operated Relief Valve was used and failed to close.

D. 1.) Turn off all Pressurizer Heaters and open the Pressurizer Vents to Containment.

2.) When normal Pressurizer Spray is unavailable it minimizes Auxiliary Spray thermal stress concerns and Power Operated Relief Valve inventory loss.

Proposed Answer: B Page 88 of 88 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible because it could be thought that the other options were not as acceptable and this option is utilized if there is no other method, however, the procedurally preferred option is cycling the Power Operated Relief Valve and then use of Auxiliary Spray.

B. Correct. The procedurally preferred option is cycling the Power Operated Relief Valve. This is preferred over Auxiliary Spray due to the Auxiliary Spray path creates high thermal stresses on the Spray Nozzles and may not have sufficient capacity.

C. Incorrect. Plausible because it could be thought that this path would be preferred, however, the Power Operated Relief Valve method is preferred due to the Auxiliary Spray path creates high thermal stresses on the Spray Nozzles and may not have sufficient capacity.

D. Incorrect. Plausible because this path is used in the EOPs as a bleed path, however, this path is not part of this step for Reactor Coolant System pressure reduction.

Technical Reference(s) EOP-3.0A, Steps 19 & 20 Attached w/ Revision # See EOP-3.0A, Attachment 6, Step 19 Comments / Reference Proposed references to be provided during examination: None Learning Objective: Given a procedural step, or sequence of steps from EOP-3.0, STATE the purpose/bases for the step(s). (LO21.ERG.E3A.OB103)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 5, 10 55.43 Page 89 of 89 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From EOP-3.0A, Step 19 Revision # 8 Page 90 of 90 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From EOP-3.0A, Step 20 Revision # 8 Page 91 of 91 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From EOP-3.0A, Attachment 6, Step 19 Revision # 8 Page 92 of 92 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 022 AA2.04 Importance Rating 2.9 Loss of Reactor Coolant Makeup: Ability to determine and interpret the following as they apply to the Loss of Reactor Coolant Makeup: How long pressurizer level can be maintained within limits Proposed Question: Common 49 Given the following conditions with Unit 1 at 100% power:

• All four (4) Pressurizer pressure channels indicate 2235 psig and stable.

• All four (4) TAVE channels indicate 585ºF and stable.

• All Charging Pumps have been lost due to gas binding.

• Letdown flow has been isolated.

• Total #1 seal leakoff flow for all Reactor Coolant Pumps is 11 gpm.

• Pressurizer level is 57%.

• Previous Reactor Coolant System Total Calculated Leakage is 0.45 gpm.

• There is no leakage into the Pressurizer Relief Tank.

• Reactor Coolant System temperature has NOT changed.

Which ONE (1) of the following is the time required to restore Reactor Coolant Makeup before Reactor Coolant System Pressure control is lost?

A. 182 minutes B. 245 minutes C. 334 minutes D. 437 minutes Proposed Answer: B Page 93 of 93 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible because if they assumed a value of 65 gal/% level (ABN-103 Attachment 1 Thumb Rule) but also assumed that control was lost when level was less than program (25%) then (57% - 25%) = 32% x 65 gal/% = 2080 gal ÷ 11.45 gal/min = 182 minutes.

B. Correct. With a value of 70 gal/% level (ABN-103 Attachment 3 Thumb Rule) then (57% - 17%) =

40% x 70 gal/% = 2800 gal ÷ 11.45 gal/min = 245 minutes.

C. Incorrect. Plausible because if they added an assumed value of one gpm for each Reactor Coolant Pump for Seal #2 Leak off then this would be the time calculated, however, the Total Calculated Leakage includes the Seal Leak off from the #2 Seals. 7780.3 (PRZR@57%) - 2771.1 (PRZR@17%) = 5009.2 gallons / 15 gpm = 334 minutes.

D. Incorrect. Plausible because system leakage is 11.45 gpm total and the volume to reach the low level heater cutout at 17% PRZR level is 7780.3 (PRZR@57%) - 2771.1 (PRZR@17%) = 5009.2 gallons / 11.45 gpm = 437 minutes, however, this Graph is based on 70ºF.

Technical Reference(s) TDM-804A, Page 24 Attached w/ Revision # See OPT-303, Steps 8.4.1 Comments / Reference ABN-706, Step 2.2.a.1)

ABN-103, Attachments 1 & 3 Proposed references to be provided during examination: TDM-804A, Page 24 ABN-103, Attachments 1 and 3 Learning Objective: STATE the physical connections and EVALUATE the cause-effect relationships between the CVCS and the following systems, components or events:

• RCPs, including seal injection flows. (OP51.SYS.CS1.OB10)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 5, 10 55.43 Page 94 of 94 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From TDM-804A, Page 24 Revision # 2 Page 95 of 95 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From OPT-303, Step 8.4.1 Revision # 13 Page 96 of 96 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ABN-706, Step 2.2.a.1) Revision # 7 Page 97 of 97 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ABN-103, Attachment 1, Page 2 of 2 Revision # 8 Page 98 of 98 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ABN-103, Attachment 3, Page 1 of 1 Revision # 8 Page 99 of 99 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # W/E11 G 2.1.28 Importance Rating 4.1 Loss of Emergency Coolant Recirculation: Conduct of Operations: Knowledge of the purpose and function of major system components and controls Proposed Question: Common 50 Given the following conditions on Unit 1:

• A Loss of Coolant Accident is in progress.

• The crew has just aligned Cold Leg Recirculation per EOS-1.3A, Transfer to Cold Leg Recirculation.

• Residual Heat Removal Pump 1-01 is showing signs of cavitation.

Which ONE (1) of the following describes the required actions to be performed in this situation and the rationale for these actions?

A. Secure Residual Heat Removal Pump 1-01 and allow the Safety Injection Pump to draw suction through the RHR flowpath.

B. Secure Containment Spray Pumps 1-01 or 1-02 to reduce total flow from the Containment Emergency Sump.

C. Reopen Residual Heat Removal Pump 1-01 suction flowpath from the Refueling Water Storage Tank and align makeup to improve Net Positive Suction Head.

D. Secure Safety Injection Pump 1-01 and Centrifugal Charging Pump 1-01 because they will reduce total flow through Residual Heat Removal Pump 1-01.

Proposed Answer: D Page 100 of 100 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible because it could be thought that with this being a large draw on the sump and being the pump that showed cavitation then this would be the correct action, however, the RHR Pump must be running to provide adequate NPSH to the Safety Injection Pump.

B. Incorrect. Plausible because reducing the total flow out of the sump could improve the suction head, however, EOS-1.3A, Transfer to Cold Leg Recirculation specifies securing the associated Safety Injection and Centrifugal Charging Pumps.

C. Incorrect. Plausible because this action could assist suction head if the RWST still had some inventory, however, adding more inventory wouldnt necessarily alleviate the cause of the cavitation which is sump blockage and EOS-1.3A, Transfer to Cold Leg Recirculation specifies securing the associated Safety Injection and Centrifugal Charging Pumps.

D. Correct. EOS-1.3A, Transfer to Cold Leg Recirculation specifies securing the associated Safety Injection and Centrifugal Charging Pumps to reduce total flow through the RHR Pump.

Technical Reference(s) EOS-1.3A, Step 3 Caution Attached w/ Revision # See EOS-1.3A, Attachment 3, Step 3 Bases 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-1.3. (LO21.ERG.E13.OB06)

Given a procedural step, or sequence of steps from EOS-1.3, STATE the purpose/basis for the step(s). (LO21.ERG.E13.OB03)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 5, 10 55.43 Page 101 of 101 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From EOS-1.3A, Step 3 Caution Revision # 8 Comments /

Reference:

From EOS-1.3A, Attachment 3, Step 3 Caution Bases Revision # 8 Page 102 of 102 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # W/E04 EK2.1 Importance Rating 3.5 LOCA outside Containment: Knowledge of the interrelations between the LOCA Outside Containment and the following:

Components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features Proposed Question: Common 51 Given the following conditions on Unit 1:

• The plant is in MODE 4 at 335ºF and 1500 psig for ECCS Check Valve Testing.

• The following Annunciators are in alarm:

• 1-ALB-4B, Window 1.3 - RHRP 1 RCS PRESS HI SUCT VLV 8702A OPEN.

• 1-ALB-4B, Window 2.3 - RHRP 1 RCS PRESS HI SUCT VLV 8701A OPEN.

• All other Main Control Board annunciators are clear or expected.

• Pressurizer level is 50% and lowering.

• Pressurizer Relief Tank level is 80% and stable.

• Pressurizer Relief Tank temperature is 85ºF and stable.

• Unit 1 Safeguards Building radiation levels are rising.

• Unit 1 Safeguards Building Sump levels are rising with higher than normal Sump Pump run times.

Which ONE (1) of the following describes the event in progress and the design feature that should prevent the event?

A Loss of Coolant Accident on Train A Residual Heat Removal suction piping A. inside Containment.

RHR Hot Leg Suction valve open alarm at 415 psig.

B. inside Containment.

RHR Hot Leg Suction Valve auto-closure at 700 psig.

C. outside Containment.

RHR Pump Suction Relief Valve opening at 450 psig.

D. outside Containment.

RHR Hot Leg Suction Valve interlock to close valves at 364 psig.

Proposed Answer: C Page 1 of 107 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible because PRT parameters changing may cause selection of LOCA in Containment while the valve open alarm could prevent or minimize the event with operator action.

B. Incorrect. Plausible because PRT parameters changing may cause selection of LOCA in Containment and auto-closure at 700 psig was a prior design.

C. Correct. LOCA is Outside Containment based on Safeguard Building radiation and Sump parameters and the RHR Pump Suction Relief Valve opens at 450 psig and will relieve 900 gpm.

D. Incorrect. Plausible because LOCA is Outside Containment based on Safeguard Building radiation and Sump parameters but with RCS pressure at 364 psig only prevents opening of the RHR Hot Leg Suction Valve.

Technical Reference(s) ALM-0042A, 1-ALB-4B, Window 2.3 Attached w/ Revision # See ALM-0021A, 1-ALB-2A, Window 4.3 Comments / Reference PO51.SYS.RH1, Pages 22, 23, & 29 Proposed references to be provided during examination: None Learning Objective: IDENTIFY all relief valves associated with the Residual Heat Removal System and STATE their setpoint and function. (OP51.SYS.RH1.OB11)

DESCRIBE system configurations or situations which could result in a loss of Reactor Coolant inventory through the Residual Heat Removal system.

(OP51.SYS.RH1.OB20)

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 2 of 107 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ALM-0042A, 1-ALB-4B, Window 2.3 Revision # 7 Page 3 of 107 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ALM-0042A, 1-ALB-4B, Window 2.3 Revision # 7 Page 4 of 107 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ALM-0021A, 1-ALB-2A, Window 4.3 Revision # 9 Page 5 of 107 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ALM-0021A, 1-ALB-2A, Window 4.3 Revision # 9 Page 6 of 107 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From OP51.SYS.RH1, Page 22 Revision # 4 RESIDUAL HEAT REMOVAL PUMP HOT LEG RECIRCULATION ISOLATION VALVES (U-8701 A&B AND U-8702 A&B)

The RHR Pump Hot Leg Recirculation Isolation Valves align the Reactor Coolant System hot legs to the suction of the RHR Pumps. The suction lines tap off of the RCS at a 45º angle to the RCS piping. The valves and their control circuits are required to provide a barrier between the higher pressure Reactor Coolant System (2235 psig when at power) and the lower pressure RHR system piping which is rated for 600 psig. During normal plant operation, these valves are closed. Once the Reactor Coolant System has been cooled to 350ºF, these valves are opened to allow the RHR System to be placed into its shutdown cooling mode.

Comments /

Reference:

From OP51.SYS.RH1, Page 23 Revision # 4 Since these valves are motor operated valves, they fail in the as is condition on a loss of power from their associated buses. Valves, u-8701B and u-8702A have alternate power supplies so both RHR lines can be isolated for the Reactor Coolant System or one line can be placed in service even with a single power supply failure.

In order for the Reactor Operator to open the Train A RHR Pump Hot Leg Recirculation Isolation Valves (u-8701A and u-8702A), the following interlocks must be met:

• the Containment Sump to RHR Pump Suction Isolation Valve (u-8811A) must be CLOSED, and

• the RWST to RHR Pump Suction Valve (u-8812A) must be CLOSED, and

• the RHR Pump to CCP/SIP Suction Valve (u-8804A) must be CLOSED, and

• detected Reactor Coolant System pressure from pressure transmitter PT-405 (Train A) must be less than 364 psig, and

• the valve handswitch on CB-04 placed in its OPEN position.

The interlocks for the Train B valves utilize their train B counterpart. PT-403 would be used instead of PT-405. The above interlocks are bypassed when control is transferred to the Remote Shutdown Panel for u-8701A and u-8701B.

Page 7 of 107 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From OP51.SYS.RH1, Page 29 Revision # 4 RHR PUMP SUCTION RELIEF VALVES (U-8708A&B)

The RHR System is designed for 600psig, thus when the RCS Hot Leg Recirculation Isolation Valves are open, the RHR System requires overpressure protection. The RHR Pump Suction Relief Valves have a design capacity of 900 gpm and discharge to the Pressurizer Relief Tank.

The relief setpoint is 450 psig and relief capacity requirements are calculated based on 2 analyzed situations:

• A loss of Instrument Air occurs just after one train of RHR is connected to the Reactor Coolant System. One Centrifugal Charging Pump is operating. As a result of the loss of Instrument Air, the Charging Flow Control Valve (u-FCV-121) fails open and the RHR Letdown Flow Control Valve (HCV-0128) fails closed. To prevent the over pressurization of the RHR System, the relief valve is required to relieve 475 gpm at 375ºF.

• A loss of Instrument Air occurs during cold shutdown or refueling operations with one train of RHR connected to the Reactor Coolant System. The analysis conservatively assumes both Centrifugal Charging Pumps are operating since we spend extended time periods in this mode. As a result of the loss of Instrument Air, the Charging Flow Control Valve (u-FCV-121) fails open and the RHR Letdown Flow Control Valve (u-HCV-0128) fails closed. To prevent the over pressurization of the RHR System, the relief valve is required to relieve 770 gpm at 200ºF.

Page 8 of 107 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 026 AA2.03 Importance Rating 2.6 Loss of Component Cooling Water: Ability to determine and interpret the following as they apply to the Loss of Component Cooling Water: The valve lineups necessary to restart the CCWS while bypassing the portion of the system causing the abnormal condition Proposed Question: Common 52 Given the following conditions on Unit 2:

• Component Cooling Water Pumps 2-01 and 2-02 are operating.

• Component Cooling Water Surge Tank Level is lowering with the following Annunciators in alarm:

• 2-ALB-3B, Window 2.4 - CCW SRG TK TRN A LVL HI-HI/LO.

• 2-ALB-3B, Window 1.3 - CCW SRG TK TRN A/B LVL LO-LO.

• Component Cooling Water Surge Tank levels are 39% and slowly lowering on each compartment.

Which ONE (1) of the following describes the required operator actions to identify the leak source?

A. CLOSE the Non-Safeguards Loop Isolation Valves and monitor Tank compartment levels.

If levels stabilize then the leak is on the Non-Safeguards Loop and the Reactor must be tripped due to loss of Reactor Coolant Pump Cooling.

If level continues to drop re-open the Non-Safeguards Loop Isolation Valves.

B. CLOSE the Safeguards Loop Supply and Return Isolation Valves one Train at a time and monitor Tank compartment levels.

If level stabilizes on the isolated side then the leak is on the Non-Safeguards Loop or the opposite Train Safeguards Loop.

Shift the Non-Safeguards Loop to the unaffected Train.

C. CLOSE the Safeguards Loop Supply and Return Isolation Valves one Train at a time and monitor Tank compartment levels.

The leak will be on the side that continues to fall below 37% with the other side stable.

Ensure the Non-Safeguards Loop is aligned to the unaffected Train.

D. CLOSE the Safeguards Loop Supply and Return Isolation Valves one Train at a time and monitor Tank compartment levels.

The leak will be on the side that continues to fall below 58% with the other side stable.

Ensure the Non-Safeguards Loop is aligned to the unaffected Train.

Page 9 of 107 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Proposed Answer: C Explanation:

A. Incorrect. Plausible because it could be thought that with both compartments lowering, the source must be the common piping and the methodology would identify a leak on the Non-Safeguards header, however, the isolation of Reactor Coolant Pump cooling should not be performed if other procedurally specified lineups are available.

B. Incorrect. Plausible because it could be thought that the tank was completely divided and then the methodology would be correct, however, the tanks are common until 37% level which is where the partition plate starts.

C. Correct. The tank is common above 37% on Unit 2 and the leak cannot be identified using this methodology until level reaches 37%. This is the procedurally specified method.

D. Incorrect. Plausible because the tank is common above 58% on Unit 1 and the leak cannot be identified using this methodology until level reaches 58%. This is the procedurally specified method but the wrong Unit.

Technical Reference(s) ABN-502, Steps 3.3.1 Note & 3.3.5 Attached w/ Revision # See OP5.1.SYS.CC1.LN, Figure 1 Comments / Reference ALM-0032B, 2-ALB-3B, Windows 1.3 & 2.4 Proposed references to be provided during examination: None Learning Objective: ANALYZE the indications and DESCRIBE the mitigation strategy for the following procedures as they affect the Component Cooling Water system:

• ABN-502, Component Cooling Water System Malfunctions (OP51.SYS.CC1.OB21)

DESCRIBE any unit differences between the Unit 1 and Unit 2 Component Cooling Water System or components. (OP51.SYS.CC1.OB15)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7, 8, 10 55.43 Page 10 of 107 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From OP5.1.SYS.CC1.LN Figure 1 Revision # 2 Page 11 of 107 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ABN-502, Step 3.3.1 Note Revision # 6 Page 12 of 107 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ABN-502, Step 3.3.5 Revision # 6 Page 13 of 107 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ABN-502, Step 3.3.5 Revision # 6 Page 14 of 107 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ALM-0032B, 2-ALB-3B, Window 1.3 Revision # 3 Page 15 of 107 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ALM-0032B, 2-ALB-3B, Window 2.4 Revision # 3 Page 16 of 107 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 054 AA1.01 Importance Rating 4.5 Loss of Main Feedwater: Ability to operate and/or monitor the following as they apply to the Loss of Main Feedwater: AFW controls, including the use of alternate AFW sources Proposed Question: Common 53 Given the following conditions:

• 1-FCV-510, SG 1 FW FLO CTRL VLV failed open.

• Unit 1 tripped when both Main Feedwater Pumps and the Main Turbine tripped on P-14, Steam Generator Hi-Hi Level on Steam Generator 1-01.

• Auxiliary Feedwater flow was initiated but was subsequently lost.

• All Steam Generator levels are 20-30% narrow range and slowly lowering.

• Reactor Coolant System pressure is 2200 psig and slowly rising.

• All Reactor Coolant Pumps have been stopped.

Which ONE (1) of the following is the FIRST choice for establishing a heat sink when Auxiliary Feedwater flow is lost?

A. Align Condensate Pumps through the FW Isolation Bypass and Feedwater Control Bypass Valves to feed the Steam Generators.

B. Start a Condensate Pump, reset Feedwater Isolation, depressurize a Steam Generator and bypass the Main Feedwater Pumps and establish flow through the Feedwater Isolation Bypass and Feedwater Control Bypass Valves.

C. Actuate Safety Injection to provide water for Core Cooling while continuing to establish Auxiliary Feedwater flow.

D. Start a Condensate Pump, reset Feedwater Isolation, start a Main Feedwater Pump and establish Main Feedwater flow through the Feedwater Isolation Bypass and Feedwater Control Bypass Valves.

Proposed Answer: D Page 17 of 107 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible because this method would be used if a Main Feedwater Pump could not be restarted, however, it is the third option after Auxiliary Feedwater flow is lost.

B. Incorrect. Plausible because this method would be used if a Main Feedwater Pump could not be restarted, however, it is a variation of the third option after Auxiliary Feedwater flow is lost.

C. Incorrect. Plausible because this method (Bleed and Feed) could be used, however, it is last in the list of options to cool the core.

D. Correct. When Auxiliary Feedwater flow is lost, the first option is to attempt to restart a Main Feedwater Pump, reset Feedwater Isolation, and feed the Steam Generators via the Feedwater Bypass Control Valves.

Technical Reference(s) OP51.SYS.MS1.LN, Page 60 Attached w/ Revision # See FRH-0.1A, Flow Chart Comments / Reference Proposed references to be provided during examination: None Learning Objective: Given that a loss of all Auxiliary Feedwater Pumps has occurred, DESCRIBE Operator actions necessary to re-establish flow to the Steam Generators, and DISCUSS the Bases for these actions to ensure compliance with the associated Individual Plant Evaluation. Discussion should include specific flowpath options, methods for aligning each of the flow paths, and the preferred order of establishing the available flow paths. (OP51.SYS.MF1.OB26)

Question Source: Bank #

Modified Bank # SYS.MF1.OB26-1 (Note changes or attach parent)

New Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 Page 18 of 107 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From OP51.SYS.MS1.LN, Page 60 Revision # 1 FRH-0.1 - RESPONSE TO LOSS OF SECONDARY HEAT SINK The operator attempts to restore or establish AFW flow first due to the accessibility of the controls from the main control board. It may be necessary to dispatch personnel because of the inability to obtain 460 gpm AFW flow.

The second method aligns Main Feedwater flow to the SGs. RCPs are tripped to reduce RCS heat input, thereby extending the time available (by conserving SG inventory) to establish feed flow from either the main feedwater or condensate system. Conditions are established to start a FWP, and a FWP is started. Feedwater isolation is reset, the FWIBVs are opened manually and the FCBVs are used to throttle flow to the SG. For Unit 2, if the FWIBVs cannot be opened, then the FPBVs may be used to provide an alternate feedwater flowpath.

If main feedwater flow cannot be established, the third method utilizes the same flowpath (as in the main feedwater case) with a condensate pump providing feedwater to the SG. This method requires depressurizing the RCS (to allow blocking of automatic safety injection signals) and at least one SG to

< 500 psig. This enables the approximately 600 psig Condensate pump discharge pressure to supply feedwater to the SG.

Should all the previously discussed attempts fail to provide feedwater flow to at least one SG then a bleed path is established via a PORV and a feed path is established via SI actuation. Bleed and feed heat removal is maintained until the secondary heat sink is reestablished and verified. After RCS bleed and feed heat removal is established, the operator continues attempts to restore SG narrow range level in at least one SG. After level is established, the effectiveness of the secondary heat sink is verified by decreasing RCS temperatures.

With a verified secondary heat sink, the operator performs a coordinated sequence for ECCS flow reduction and closing of pressurizer PORVs. After the completion of the sequence, the operator is transferred to EOS-1.1, SI TERMINATION, for plant recovery.

Page 19 of 107 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From FRH-0.1A, Flow Chart Revision # 8 Comments /

Reference:

From Exam Bank SYS.MF1.OB26-1 Revision # 12/05/96 Which of the following is the first choice for establishing a heat sink if AFW flow is lost?

(1.0 pt)

A. Aligning Condensate pumps through the Feed Preheater Bypass to feed Steam Generators.

B. Depressurize the Steam Generators to align Main Feedwater through the Feed Preheater Bypass to feed Steam Generators.

C. Actuate SI to provide water for core cooling while continuing to establish AFW flow.

Reset FW Isolation and establish MFW flow to the Steam Generators through FIBV and FCBV.

Page 20 of 107 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 029 EK1.05 Importance Rating 2.8 ATWS: Knowledge of the operational implications of the following as they apply to the ATWS: Definition of negative temperature coefficient as applied to large PWR coolant systems Proposed Question: Common 54 Given the following conditions for Unit 1:

• An Anticipated Transient Without Trip (ATWT) event is in progress.

• FRS-0.1A, Response to Nuclear Power Generation/ATWT is in progress and the Reactor is still NOT tripped.

• Boration CANNOT be initiated because of blockage in the boration flow paths and 1-LCV-112D and 1-LCV-112E will NOT open from the Control Room.

• All Power Range Channels indicate 6%.

• Startup rate is zero (0) on both Intermediate Range Channels.

• Reactor Coolant System average temperature (TAVE) is 570ºF and slowly lowering.

• Average Core Exit Thermocouple temperature is 580ºF and slowly lowering.

• Containment Pressure is 0.3 psig and stable.

Which ONE (1) of the following describes the operator actions under these conditions and the primary reason for taking these actions?

A. Transition to FRS-0.2A, Response to Loss of Core Shutdown, Step 1 as it is now the procedure and step in effect once it has been identified that Containment pressure is less than 5 psig.

B. Remain in FRS-0.1A, Response to Nuclear Power Generation / ATWT, and allow RCS temperature to lower while continuing efforts to establish Emergency Boration.

A lower temperature will maintain an appropriate DNBR margin.

C. Transition to FRS-0.2A, Response to Loss of Core Shutdown. This is required by the Critical Safety Function SUBCRITICALITY Status Tree based because Power Range levels are less than 10%.

D. Remain in FRS-0.1A, Response to Nuclear Power Generation / ATWT, and allow the RCS to heat up while continuing efforts to establish Emergency Boration. The heatup will insert negative reactivity.

Proposed Answer: D Page 21 of 107 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible because review of the SUBCRITICALITY Status Tree would indicate that this is a possible transition, however, exit conditions for FRS-0.1 have not been met.

B. Incorrect. Plausible because procedure entry is correct, however, actions include allowing the RCS to heat up and add negative reactivity.

C. Incorrect. Plausible because review of the SUBCRITICALITY Status Tree would indicate that this is the current CSFST condition, however, exit conditions for FRS-0.1 have not been met.

D. Correct. Remaining in FRS-0.1A and allowing the plant to heat up will insert negative reactivity.

Technical Reference(s) FRS-0.1A, CSFST Attached w/ Revision # See FRS-0.1A, Step 17 RNO 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 FRS-0.1A/B, Response to Nuclear Generation/ATWT. (LO21.FRG.FS1.OB04)

Question Source: Bank #

Modified Bank # FRS.XH1.OB404-5 (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 1, 5, 10 55.43 Page 22 of 107 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From FRS-0.1A, CSFST Revision # 8 Page 23 of 107 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From FRS-0.1A, Step 17 RNO Revision # 8 Comments /

Reference:

From Exam Bank FRS.XH1.OB404-5 Revision # 0 Given the following conditions:

• The reactor fails to trip when required.

• The Operators take actions per the appropriate procedure(s) and obtain the required plant/system/component responses, except that the reactor is still NOT tripped.

• Boration CANNOT be initiated because of blockage in the boration flow paths.

• All Power Range channels indicate 3%.

• Startup rate is zero on both Intermediate Range channels.

Which of the following describes the correct operator actions under these conditions AND the primary reason for taking these actions?

A. Return to the procedure and step in effect. Power is less than 5% and the IR startup rate is zero.

B. Remain in FRS-0.1A, Response to Nuclear Power Generation / ATWT and allow the RCS to heat up while continuing efforts to establish emergency boration.The heatup will insert negative reactivity.

C. Go to FRS-0.2A, Response to Loss of Core Shutdown. This is required by the Subcriticality CSFST based on the current reactor conditions.

D. Remain in FRS-0.1A, Response to Nuclear Power Generation / ATWS, and maintain RCS temperature stable while continuing efforts to establish emergency boration. Stable temperatures preclude positive reactivity insertion by cooldown.

Page 24 of 107 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # W/E12 G 2.4.11 Importance Rating 4.0 Uncontrolled Depressurization of All Steam Generators: Emergency Procedures/Plan: Knowledge of abnormal condition procedures Proposed Question: Common 55 Given the following conditions:

• An Uncontrolled Depressurization of all Steam Generators into Containment is in progress on Unit 1.

• All required Safety Systems have actuated as required.

• Containment pressure is 12 psig and rising.

• The crew has performed the actions in EOP-0.0A, Reactor Trip or Safety Injection and transitioned to EOP-2.0A, Faulted Steam Generator Isolation.

• Step 2 of EOP-2.0A, Faulted Steam Generator Isolation directs the crew to ECA-2.1A, Uncontrolled Depressurization of All Steam Generators.

Which ONE (1) of the following describes the mitigation strategy that will be employed in ECA-2.1A, Uncontrolled Depressurization of All Steam Generators?

A. Feed all Steam Generators with level less than 43% narrow range at 100 gpm and cool the Reactor Coolant System down to Residual Heat Removal entry conditions.

B. Feed all Steam Generators with level less than 50% narrow range at 100 gpm and cool the Reactor Coolant System down to Residual Heat Removal entry conditions.

C. Feed one (1) Steam Generator with level less than 67% narrow range at 100 gpm and cool the Reactor Coolant System down to Residual Heat Removal entry conditions.

D. Isolate all Steam Generators and align for Reactor Coolant System Bleed and Feed prior to the Steam Generators losing all inventory.

Proposed Answer: B Page 25 of 107 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible because the mitigation strategy of ECA-2.1A, Uncontrolled Depressurization of All Steam Generators will feed all Steam Generators at 100 gpm if level is < 43% (without Adverse Containment) and allow the plant to cooldown to Residual Heat Removal entry conditions.

B. Correct. The mitigation strategy of ECA-2.1A, Uncontrolled Depressurization of All Steam Generators will feed all Steam Generators at 100 gpm if level is < 50% (Adverse Containment) and allow the plant to cooldown to Residual Heat Removal entry conditions.

C. Incorrect. Plausible because it could be thought that feeding only one Steam Generator would protect the tube integrity of the other 3 and the goal to cooldown using the secondary to RHR would be correct, however, this is not the strategy employed by ECA-2.1A, Uncontrolled Depressurization of All Steam Generators.

D. Incorrect. Plausible because it could be thought that initiating Reactor Coolant System Bleed and Feed while there was inventory in the secondary would be advantageous to control the Reactor Coolant System cooldown, however, this is not the strategy employed by ECA-2.1A, Uncontrolled Depressurization of All Steam Generators.

Technical Reference(s) ECA-2.1A, Flow Chart Attached w/ Revision # See ECA-2.1A, Step 2 Comments / Reference Proposed references to be provided during examination: None Learning Objective: Given a Note or Caution from ECA-2.1, STATE the reason for the Note or Caution. (LO21.ERG.C21.OB04)

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 26 of 107 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ECA-2.1A, Flow Chart Revision # 8 Page 27 of 107 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ECA-2.1A, Step 2 Revision # 8 Page 28 of 107 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 025 AA1.12 Importance Rating 3.6 Loss of RHR System: Ability to operate and/or monitor the following as they apply to the Loss of Residual Heat Removal System: RCS temperature indicators Proposed Question: Common 56 Given the following conditions on Unit 1:

• Reactor Coolant System temperature is 120ºF.

• Reactor Coolant System level is 80 inches above the core plate.

• Residual Heat Removal System loop flow is 2500 gallons per minute.

• Time after shutdown is 700 hours0.0081 days <br />0.194 hours <br />0.00116 weeks <br />2.6635e-4 months <br />.

A complete Loss of Residual Heat Removal occurs and the Residual Heat Removal System cannot be restored.

Which ONE (1) of the following reflects the amount of time until saturation conditions exist in the Reactor Coolant System?

A. 19.5 minutes B. 22.5 minutes C. 25.0 minutes D. 30.5 minutes Proposed Answer: C Explanation:

A. Incorrect. Plausible because this would be the correct time if Reactor Coolant System initial temperature were 140ºF.

B. Incorrect. Plausible if the interpolation was misread between the 100ºF and 140ºF curves C. Correct. Interpolating between the 100ºF and 140ºF curves at 700 hours0.0081 days <br />0.194 hours <br />0.00116 weeks <br />2.6635e-4 months <br /> after shutdown yields a time to saturation of 25 minutes.

D. Incorrect. Plausible because this would be the correct time if Reactor Coolant System initial temperature were 100ºF.

Technical Reference(s) IPO-010A, Attachment 8 Attached w/ Revision # See IPO-010A, Step 3.2.5 Comments / Reference Page 29 of 107 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Proposed references to be provided during examination: IPO-010A, Attachment 8 Learning Objective: DESCRIBE or STATE how the following concepts or conditions apply to the Residual Heat Removal System:

• Loss of RHR cooling capability (OP51.SYS.RH1.OB14)

DESCRIBE the basis for the precautions and limitations, and given major procedure steps relative to the Residual Heat Removal System, PLACE them in the proper sequence, for:

• IPO-010, Reactor Coolant System Reduced Inventory Operations (OP51.SYS.RH1.OB18)

Question Source: Bank #

Modified Bank # IPO.XO0.OB07-11 (Note changes or attach parent)

New Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 Page 30 of 107 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From IPO-010A, Attachment 8 Revision # 17 Page 31 of 107 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From IPO-010A, Step 3.2.5 Revision # 17 Page 32 of 107 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From Exam Bank IPO.XO0.OB07-11 Revision # 01/06/99 Given the following conditions on Unit 1:

• Reactor Coolant System temperature is 100ºF.

• Reactor Coolant System level is 80 inches above the core plate.

• Residual Heat Removal System loop flow is 2400 gallons per minute.

• Time after shutdown is 850 hours0.00984 days <br />0.236 hours <br />0.00141 weeks <br />3.23425e-4 months <br />.

A complete Loss of Residual Heat Removal occurs and the Residual Heat Removal System cannot be restored.

Which ONE (1) of the following reflects the amount of time until saturation conditions exist in the Reactor Coolant System?

A. 19 minutes B. 21 minutes C. 31 minutes D. 33 minutes Page 33 of 107 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 2 K/A # 061 AK1.01 Importance Rating 2.5 Area Radiation Monitoring System Alarms: Knowledge of the operational implications of the following concepts as they apply to Area Radiation Monitoring System Alarms: Detector limitations Proposed Question: Common 57 Which ONE (1) of the following identifies the use and types of detectors in the Area Radiation Monitoring System?

Area Radiation Monitors can be equipped with either Geiger-Mueller tubes, generally used in low radiation areas where detector __________ is not a concern, or __________ detectors typically used in high radiation areas.

A. sensitivity; scintillation B. saturation; ion chamber C. sensitivity; ion chamber D. saturation; scintillation Proposed Answer: B Explanation:

A. Incorrect. Plausible if thought that Geiger Mueller tubes are not sensitive to radiation. Additionally, a scintillation detector would easily saturate in a high radiation area.

B. Correct. Geiger Mueller tubes are generally used in low radiation areas due to their degree of sensitivity and where detector saturation is not a concern whereas ion chambers are better suited for high radiation area applications.

C. Incorrect. Plausible because ion chamber is used in a high radiation area, however, Geiger Mueller tubes are generally restricted to low radiation areas due to their sensitivity.

D. Incorrect. Plausible because detector saturation is correct, however, a scintillation detector would easily saturate in a high radiation area.

Technical Reference(s) OP51.SYS.RM1.LN, Pages 15 & 23 Attached w/ Revision # See Comments / Reference Proposed references to be provided during examination: None Learning Objective: DESCRIBE operation of the following monitors, description should include flow path, filtration and detection method.

• Area Radiation Monitors (OP51.SYS.RM1.OB09)

Page 34 of 107 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Question Source: Bank # SYS.RM1.OB09-6 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 Comments /

Reference:

From OP51.SYS.RM1.LN, Page 15 Revision # 1 The area monitors use several different types of detectors and hardware configurations to provide detection and alarm over a large dose range. Two types of area monitors are used, the Low Range Monitor and the High Range Monitor. These monitors provide high radiation alarms on increased radiation levels.

The low range area monitors typically use Geiger-Mueller (GM) tubes for detection. The high range area monitors typically use Ion Chamber detectors and are primarily a post accident monitor. The detectors provide input to the RM-80 monitor.

Page 35 of 107 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From OP51.SYS.RM1.LN, Pages 15 & 23 Revision # 1 DETECTORS Four types of radiation detectors are used in the Radiation Monitoring System. These are Geiger-Mueller, Ionization Chambers, Semiconductor, and Scintillation. Figure 3 of OP51.SYS.EC1.LN shows the Gas Amplification Curve for gas filled detectors.

Geiger-Mueller (GM) tubes are used as detectors in the Failed Fuel Monitor, gaseous monitors without a sample pump, and low level area radiation monitors due to their sensitivity to low energy, low occurrence radiation. The GM tubes completely ionize which ensures the signal generated is large enough to be detected by the electronics.

GM tubes can detect even the lowest intensity radiation field due to the total ionization of the gas volume. Therefore, only simple electronic amplification of the detector signal is required. The total ionization of the gas results in two disadvantages in using the GM tube. First, the GM tube cannot distinguish differences in incident radiation energy levels. Second, they have an upper limit on the count rate due to dead time (time for the gas to de-ionize).

Page 36 of 107 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 2 K/A # 001 AA2.03 Importance Rating 4.5 Continuous Rod Withdrawal: Ability to determine and interpret the following as they apply to the Continuous Rod Withdrawal:

Proper actions to be taken if automatic safety functions have not taken place Proposed Question: Common 58 Given the following conditions:

• Power Range Nuclear Instrument Channels read as follows:

• N-41 = 80%.

• N-42 = 79%.

• N-43 = 81%.

• N-44 = 80%.

• Reactor Coolant Loop TAVE Channels read as follows:

• RC LOOP 1 TAVE CHAN I, 1-TI-412 indicates 585ºF.

• RC LOOP 2 TAVE CHAN II, 1-TI-422 indicates 584ºF.

• RC LOOP 3 TAVE CHAN III, 1-TI-432 indicates 597ºF.

• RC LOOP 4 TAVE CHAN IV, 1-TI-442 indicates 586ºF.

• Annunciator 1-ALB-6D, Window 2.10 - AVE TAVE HI is in alarm.

• Control Bank D Rods begin to withdraw at 72 steps per minute.

Which ONE (1) of the following actions should be taken and which procedure should be entered?

A. Ensure 1/1-RBSS, Control Rod Bank Select Switch NOT in AUTO and enter ABN-703, Power Range Instrument Malfunction.

B. Trip the Reactor and enter EOP-0.0A/B, Reactor Trip or Safety Injection while continuing with ABN-712, Rod Control System Malfunction.

C. Ensure 1/1-RBSS, Control Rod Bank Select Switch NOT in AUTO and enter ABN-740A, Control Room Annunciator and Status Light Malfunction.

D. Ensure 1/1-RBSS, Control Rod Bank Select Switch NOT in AUTO and enter ABN-704, Tc/N16 Instrumentation Malfunction.

Proposed Answer: D Page 37 of 107 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible because Rod Control must be placed in MANUAL, however, the wrong procedure is being referenced due to incorrect diagnosis.

B. Incorrect. Plausible because with a high TAVE alarm rod motion should be in the inward direction and the RNO actions of Step 2 require a Reactor trip and EOP-0.0 entry, however, only if Control Rod motion fails to stop.

C. Incorrect. Plausible because Rod Control must be placed in MANUAL and it could be thought that the Tave Hi alarm is due to an annunciator problem; however, the RNO actions of the ABN are not being met.

D. Correct. ABN-704 entry is required and Rod Control must be placed in MANUAL.

Technical Reference(s) ABN-712, Steps 2.3.2 & 2.3.2 RNO Attached w/ Revision # See ABN-712, Step 2.3.4 Comments / Reference ABN-704, Step 2.1 Proposed references to be provided during examination: None Learning Objective: ANALYZE the indications and DESCRIBE the mitigation strategy and major steps taken relative to the Rod Control System, both initial and subsequent, for:

• ABN-712, Rod Control System Malfunction (OP51.SYS.CR1.OB15)

Question Source: Bank #

Modified Bank # ABN-712-OB-54 (Note changes or attach parent)

New Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 Page 38 of 107 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ABN-712, Steps 2.3.2 & 2.3.2 RNO Revision # 10 Page 39 of 107 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ABN-704, Step 2.1 Revision # 10 Page 40 of 107 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ABN-712, Step 2.3.4 Revision # 10 Comments /

Reference:

From Exam Bank ABN-712-OB-54 Revision # 0 Given the following conditions:

• N-41 indicates 80%, N-42 indicates 79%, N-43 indicates 81% and N-44 indicates 80%.

• Annunciator 1-ALB-6D-2.10, AVE TAVE HI goes into alarm.

• Control Bank D Rods begin to step out at 72 steps/minute.

Which ONE (1) of the following actions should be taken and which procedure should be entered?

A. Ensure 1/1-RBSS, Control Rod Bank Select Switch NOT in AUTO and enter ABN-703, Power Range Instrument Malfunction.

B. Trip the Reactor and enter EOP-0.0A/B, Reactor Trip or Safety Injection while continuing with ABN-712, Rod Control System Malfunction.

C. Ensure 1/1-RBSS, Control Rod Bank Select Switch NOT in AUTO and initiate repairs per STA-606, Control of the Maintenance and Work Activities.

D. Ensure 1/1-RBSS, Control Rod Bank Select Switch NOT in AUTO and enter ABN-704, Tc/N16 Instrumentation Malfunction.

Page 41 of 107 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 2 K/A # 037 AA1.13 Importance Rating 3.9 Steam Generator Tube Leak: Ability to operate and/or monitor the following as they apply to the Steam Generator Tube Leak: S/G blowdown radiation monitors Proposed Question: Common 59 Given the following conditions on Unit 2:

• A tube rupture has been diagnosed on Steam Generator 2-02.

• A manual Reactor Trip and Safety Injection have been initiated.

Which ONE (1) of the following is the status of the Steam Generator Blowdown Sample Radiation Monitor?

A. INOPERABLE due to a low flow condition.

B. In service, monitoring for high radiation in the Steam Generator Blowdown stream.

C. INOPERABLE due to a load shed of its power supply.

D. In service, until closure of the Sample Valve on high radiation.

Proposed Answer: A Explanation:

A. Correct. The Safety Injection Signal causes the Steam Generator Blowdown Sampling stream to be isolated. At this point, the Radiation Monitor is INOPERABLE due to a low flow condition.

B. Incorrect. Plausible because there is no indication of high radiation and therefore the Radiation Monitor could still be in service, however, the system was isolated upon Safety Injection.

C. Incorrect. Plausible because the Radiation Monitor is INOPERABLE, however, the reason is due to a low flow condition caused by the safety injection signal D. Incorrect. Plausible if a high radiation condition actually existed, however, there is no reference to the magnitude of the tube leak.

Technical Reference(s) OP51.SYS.SB1.LN, Pages 33 & 34 Attached w/ Revision # See ALM-3200, Attachment 3 Comments / Reference Proposed references to be provided during examination: None Page 42 of 107 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Learning Objective: DESCRIBE the purpose of the Steam Generator Blowdown radiation monitor and its expected response to Steam Generator Tube leakage.

(OP51.SYS.SB1.OB13)

EXPLAIN how Steam Generator Blowdown radiation monitor response to Steam Generator Tube Leakage will be affected by isolating Steam Generator Blowdown flow. (OP51.SYS.SB1.OB14)

Question Source: Bank # SYS.SB1.OB14-1 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 Comments /

Reference:

From OP51.SYS.SB1.LN, Page 33 Revision # 1 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.

Comments /

Reference:

From OP51.SYS.SB1.LN, Page 34 Revision # 1 The SG Blowdown Radiation Monitor will lose sample flow should the blowdown stream isolate.

During a SG tube rupture, blowdown flow can automatically isolate due to the initiation of a Safety Injection Signal or high radiation. When this occurs, the SG Blowdown Radiation Monitor indication on the PC-11 will flash blue until acknowledged. This is a communications failure and will prevent the receipt of "ALERT" and "ALARM" conditions on the PC-11. To determine activity levels previously monitored, the operator must call up historical trends to view the data. To return the monitor to an active status, flow through the SG Blowdown System must be restored and the monitors sample pump restarted.

Page 43 of 107 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ALM-3200, Attachment 3 Revision # 4 Page 44 of 107 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 2 K/A # 076 AA2.03 Importance Rating 2.5 High Reactor Coolant Activity: Ability to determine and interpret the following as they apply to the High Reactor Coolant Activity: RCS radioactivity level meter Proposed Question: Common 60 Which ONE (1) of the following trends on 1-RE-0406 (FFL160), Gross Failed Fuel Monitor would indicate an increase in Reactor Coolant Activity due to an increase in failed fuel?

A. During steady state operations, occasional positive spikes occur in the reading and then returns to normal.

B. Following a power increase, the activity reading increases slowly and then slowly returns to normal.

C. Following a power increase, the activity level increases slowly and then stabilizes at a new, higher level.

D. During steady state operations, only particulate activity reading shows a continuous slow increase.

Proposed Answer: C Explanation:

A. Incorrect. Plausible because it could be thought that the activity levels could change quickly but then be cleaned up quickly, however, this indication is more indicative of a CRUD burst.

B. Incorrect. Plausible because it could be thought that the activity levels would return to normal, however, this is indicative of an increase in CRUD levels that get cleaned up as opposed to the fuel element defect that is continuously contributing to the activity level.

C. Correct. The activity level rising and leveling off is indicative of a fuel defect/failure where the activity reaches a new, higher equilibrium value.

D. Incorrect. Plausible because it could be thought that the activity levels would continue to rise, however, a slow, continuous rise is indicative of a depleting Ion Exchanger or CRUD burst.

Technical Reference(s) ABN-102, Steps 2.1.b, 2.3.1, & 2.3.8 Attached w/ Revision # See Comments / Reference Proposed references to be provided during examination: None Learning Objective: DESCRIBE how the following concepts or conditions apply to the Digital Radiation Monitoring System:

• Radiation data evaluation as related to specific types of events or conditions (OP51.SYS.RM1.OB08)

Page 45 of 107 Rev e

ES-401 CPNPP Mar 2010 NRC 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 11, 12 55.43 Comments /

Reference:

From ABN-102, Step 2.1.b Revision # 7 Page 46 of 107 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ABN-102, Step 2.3.1 Revision # 7 Comments /

Reference:

From ABN-102, Step 2.3.8 Revision # 7 Page 47 of 107 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 2 K/A # W/E13 EK2.1 Importance Rating 3.0 Steam Generator Overpressure: Knowledge of the interrelations between the Steam Generator Overpressure and the following: Components, and functions and control of safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features Proposed Question: Common 61 Given the following condition:

• When isolating a Steam Generator following a Steam Generator Tube Rupture, the Atmospheric Relief Valve on the affected Steam Generator is left in AUTOMATIC with a setpoint of 1160 psig.

Which ONE (1) of the following describes the primary basis for this action?

A. Ensures continued availability of the Atmospheric Relief Valve for heat removal in the event that the intact Steam Generators become unavailable.

B. Provides an isolable overpressure protection path for the Steam Generator and eliminate challenges to the Main Steam Safety Valves.

C. To ensure a continuous amount of heat removal from the isolated Steam Generator for Natural Circulation if Reactor Coolant Pumps are stopped.

D. To maintain Technical Specification requirements for OPERABILITY while in MODE 3.

Proposed Answer: B Page 48 of 107 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible because it could be thought that the requirement could be for heat removal path availability, however, the Atmospheric Relief Valve could be left in MANUAL and CLOSED for this requirement.

B. Correct. The Atmospheric Relief Valve on the ruptured Steam Generator should remain available to limit Steam Generator pressure unless it fails open. This also will eliminate challenges to the Main Steam Safety Valves.

C. Incorrect. Plausible because it could be thought that the requirement could be to ensure continued flow through the isolated Steam Generator, however, the procedural guidance and bases are trying to ensure that the Reactor Coolant System and isolated Steam Generator maintain a higher pressure than the intact Steam Generators for heat removal from the Reactor Coolant System.

D. Incorrect. Plausible because Atmospheric Relief Valve OPERABILITY is required in MODE 3, however, OPERABILITY does not require the controls to be in AUTOMATIC and the alignment is for Steam Generator overpressure protection.

Technical Reference(s) EOP-3.0A, Attachment 6, Step 3 Attached w/ Revision # See EOP-3.0A, Step 3 Comments / Reference Technical Specification LCO 3.7.4 Proposed references to be provided during examination: None Learning Objective: Given a procedural step, or sequence of steps from EOP-3.0, STATE the purpose/bases for the step(s). (LO21.ERG.E3A.OB103)

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 49 of 107 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From EOP-3.0A, Attachment 6, Step 3 Revision # 8 Page 50 of 107 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From EOP-3.0A, Step 3 Revision # 8 Page 51 of 107 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From Technical Specification LCO 3.7.4 Amendment # 109 Page 52 of 107 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 2 K/A # 033 G 2.2.44 Importance Rating 4.2 Loss of Intermediate Range Nuclear Instruments: Equipment Control: Ability to interpret control room indications to verify the status in operation of a system, and understand how operator actions and directives affect plant and system conditions Proposed Question: Common 62 Given the following conditions on Unit 1:

• A Technical Specification required Reactor Shutdown is in progress.

• Power Range Nuclear Instruments are reading 20%.

• Annunciator 1-ALB-6D, Window 3.2 - IR CHAN 2 CMPNSATING VOLT FAIL is in alarm.

• Intermediate Range Channel 35 is reading ~7x10-5 amperes.

• Intermediate Range Channel 36 is reading ~3x10-4 amperes.

Which ONE (1) of the following describes the required operator actions as the shutdown continues?

A. Manually UNBLOCK Intermediate Range High Level Trip prior to the OPERABLE Intermediate Range Channel going below the P-10 setpoint.

B. Manually BYPASS Intermediate Range Channel 35 High Level Trip. When Intermediate Range Channel 36 is at the P-6 setpoint, verify the Source Range Instruments energize.

C. Ensure Intermediate Range High Level Trip automatic UNBLOCK at 5%. When power is less than 1x10-10 amperes on the unaffected Intermediate Range Channel, manually energize the Source Range Instruments.

D. Manually BYPASS Intermediate Range High Level Trip for Channel 36. When power is less than 1x10-10 amperes on the unaffected Intermediate Range Channel, manually energize the Source Range Instruments.

Proposed Answer: D Page 53 of 107 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible because it could be thought that P-10 was from the Intermediate Range Channel and those two Channels would be required to re-instate the High Level Trip, however, P-6 is actuated by the Power Range Channels and the trips from the Intermediate Ranges would be automatically re-instated resulting in a trip from the failed Channel.

B. Incorrect. Plausible because it could be thought that Channel 35 was the failed channel and that P-6 would still energize, however, Channel 36 has the failure and P-6 will not deenergize unless both Channels drop below the P-6 setpoint.

C. Incorrect. Plausible because the Power Range Channels will automatically UNBLOCK the Intermediate Range High Level Trip and it could be thought that the failed Channel was less than the High Level Trip setpoint, however, it UNBLOCKS at 10%. Additionally, the action to manually energize the Source Range Channels is correct.

D. Correct. Channel 36 is the failed channel and is above the High Level Trip setpoint which requires bypassing the High Level Trip prior to going below the P-10 permissive. The P-6 permissive to energize the Source Range Channels requires both of the Intermediate Ranges Channels to go below 1x10-10 amperes and therefore, must be manually energized.

Technical Reference(s) OP51.SYS.EC1.LN, Pages 34, 35, & 39 Attached w/ Revision # See ABN-702, Step 2.3.4.b RNO Comments / Reference Proposed references to be provided during examination: None Learning Objective: LIST and EXPLAIN the Excore Instrumentation System design features which provide for the trips, permissives, and interlocks associated with the following:

• Source Range Detector High Voltage

• Source Range High Flux Trip (OP51.SYS.EC1.OB14)

STATE which Excore Instrumentation Bistables have Block features and DESCRIBE when and how these Block features are utilized.

(OP51.SYS.EC1.OB16)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 6 55.43 Page 54 of 107 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From OP51.SYS.EC1.LN, Page 34 Revision # 2 Figure 10 - Intermediate Range N35/N36 Block Diagram The output from the log current amplifier inputs 2 bistable drivers and an I/A.

A bistable provides input to the Solid State Protection System. It will cause a high level reactor trip at a current equivalent to 25% of full power. This is a one out of two reactor trip. This trip, however, can be manually blocked when permissive circuit P-10 becomes energized. P-10 occurs when reactor power is greater than 10% from two of four power range detectors. The current equivalent of 25% of full power is approximately 6 x 10-5 amperes. The bistable reset occurs at 16% RTP (current equivalent) which ensures that a reactor trip will not be generated when < P-10 setpoint Page 55 of 107 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From OP51.SYS.EC1.LN, Page 35 Revision # 2 A third bistable is used to input the P-6 permissive circuit. The P-6 permissive circuit is energized during reactor startup when 1 out of the 2 intermediate range channels reaches 10-10 amperes. This corresponds to a source range count-rate of approximately 4 x 104 cps. Once the P-6 permissive light is on, the source range trip can be manually blocked. Blocking the source range trip automatically turns off the source range detector high voltage. During reactor shutdown, the source range trip and detector high voltage is automatically activated when both intermediate range channels drop below 10-10 amperes (the actual reset is 5 x 10-11amperes) as P-6 is de-energized. In the event of an under compensation problem with either intermediate range detector (P-6 fails to de-energize), the source range may be manually activated by going to reset on the switches on the Control Board CB -7.

Comments /

Reference:

From OP51.SYS.EC1.LN, Page 39 Revision # 2 Through one B/S, P-10 (nuclear at power-permissive) is supplied. P-10 is an interlock enabled when 2 of 4 channels of the power range N.I. are above 10%. The P-10 interlock is involved in circuits that perform the following functions:

• Allows manual block of the power range, low range, high level trip (25% power).

Comments /

Reference:

From ABN-702, Step 2.3.4.b RNO Revision # 9 Page 56 of 107 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 2 K/A # 024 AK3.02 Importance Rating 4.2 Emergency Boration: Knowledge of the reasons for the following responses as they apply to Emergency Boration: Actions contained in EOP for emergency boration Proposed Question: Common 63 Given the following conditions:

• Unit 1 just tripped from 100% power.

• The Unit was two weeks from a scheduled Refueling Outage.

• Three Control Rods do NOT indicate fully inserted.

• An Emergency Boration is commenced.

Which ONE (1) of the following describes the required amount of boric acid to be injected and the basis for the boration?

A. 3000 gallons of boric acid. Accounts for the maximum reactivity worth of the additional two rods not assumed to stick out to ensure proper SHUTDOWN MARGIN on the most limiting accident which is the Large Break Loss of Coolant Accident.

B. 4500 gallons of Boric Acid. Accounts for the reactivity of each rod stuck out to ensure the Reactor is SHUTDOWN with a negative startup rate.

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

D. 7200 gallons of Boric Acid. Accounts for the reactivity worth associated with all rods failing to insert to ensure the Reactor is SHUTDOWN with a negative startup rate.

Proposed Answer: C Page 57 of 107 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible because SHUTDOWN MARGIN is the concern, however, the required boration is 1800 gallons for each rod stuck out.

B. Incorrect. Plausible because it could be thought that the concern was being subcritical with three rods stuck out and that the required amount to borate was 1500 gallons per rod, however, a boration of 1800 gallons per rod is required to ensure SHUTDOWN MARGIN in the event of a Main Steam Line Break.

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

D. Incorrect. Plausible because 7200 gallons is the amount borated if rod position post-trip is unavailable. It could be thought that this value would account for all rods and be bounding for the three rods stuck out and that the basis would be to ensure the Reactor is subcritical, however, a boration of 1800 gallons per rod is required to ensure SHUTDOWN MARGIN in the event of a Main Steam Line Break.

Technical Reference(s) Technical Specification LCO 3.1.1 Bases Attached w/ Revision # See EOP-0.0A, Step 1 & Step 1 Bases Comments / Reference EOP-0.0A, Attachment 1.A, Step 2 Proposed references to be provided during examination: None Learning Objective: DESCRIBE the function and operation of systems, components, and controls required to be operated in response to a Reactor Trip or Safety Injection.

(LO21. ERG.E0A.OB08)

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 58 of 107 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From Technical Specification LCO 3.1.1 Bases Revision # 57 Page 59 of 107 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From EOP-0.0A, Step 1 Revision # 8 Page 60 of 107 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From EOP-0.0A, Step 1 Bases Revision # 8 Page 61 of 107 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From EOP-0.0A, Attachment 1.A, Step 2 Revision # 8 Page 62 of 107 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 2 K/A # 067 AK1.02 Importance Rating 3.1 Plant Fire on Site: Knowledge of the operational implications of the following concepts as they apply to Plant Fire on Site:

Fire fighting Proposed Question: Common 64 Given the following conditions on Unit 1:

• There is a fire in the Unit 1 Cable Spreading Room.

• Associated Fire Alarms are alarming on Fire Detection Main Control Panel.

Which ONE (1) of the following describes the operation of the fire suppression and ventilation equipment for the Cable Spreading Room?

A. Automatic Dry Pipe Sprinkler System with Hose Stations for backup. Manually secure ventilation at Panel X-CV-01.

B. Manual Halon Suppression with Manual Sprinkler Backup. The Supply Fans automatically trip on the fire alarm.

C. Manual Halon Suppression with Dry Pipe Hose Stations for backup. The Supply and Exhaust Fans remain running in this condition.

D. Automatic Halon Suppression with Manual Sprinkler Backup. Supply and Exhaust Ventilation Fans automatically trip on the fire alarm.

Proposed Answer: D Explanation:

A. Incorrect. Plausible because it could be thought that this was the type employed in this case and that the fans were manual, however, Automatic Halon Suppression with Manual Sprinkler Backup is provided and Supply and Exhaust Ventilation Fans automatically trip on the fire alarm.

B. Incorrect. Plausible because manual sprinklers are the backup and it could be thought that Halon was a manual system and that the ventilation system alignment was manual, however, Automatic Halon Suppression with Manual Sprinkler Backup is provided and Supply and Exhaust Ventilation Fans automatically trip on the fire alarm.

C. Incorrect. Plausible because it could be thought that Halon was a manual system (other areas are manual) and that the backup suppression was just fire hoses.

D. Correct. Automatic Halon Suppression with Manual Sprinkler Backup is provided and Supply and Exhaust Ventilation Fans automatically trip on the fire alarm.

Page 63 of 107 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Technical Reference(s) ABN-803A, Step 2.2 Attached w/ Revision # See FPI-505, Page 3 Suppression Strategy Comments / Reference OP51.SYS.FP1.LN, Page 23 Proposed references to be provided during examination: None Learning Objective: IDENTIFY the areas protected by each of the following types of Fire Suppression systems and EXPLAIN how each is initiated:

• Halon Systems (OP51.SYS.FP1.OB07)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

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

Reference:

From ABN-803A, Step 2.2 Revision # 8 Page 64 of 107 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From FPI-505, Page 3 Suppression Strategy Revision # 3 Page 65 of 107 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From OP51.SYS.FP1.LN, Page 23 Revision # 3 Cable Spread Room Halon System Operation A main reserve selector switch determines bank discharge response to a fire. Halon systems are automatically actuated by ion smoke detectors (at least 2) within the serviced area. Actuation will initiate a fire alarm light and horn at the local panel. 60 seconds later the Halon will be discharged. A blue panel light indicates the completion of bank discharge.

The local fire alarm is also indicated at the control room fire panel. The alarm will provide for damper closure or, in the case of the cable spreading room, stop the supply and exhaust fans.

If it is necessary or desirable to shift over to the remaining bank, it can be discharged by selecting the selector switch to the charged reserve bank. To clear the alarm from the panel, the system reset pushbutton (inside cabinet) should be pressed. Provided the alarm condition is clear, this action will clear the alarm, if the conditions resulting in the alarm still exist, the alarm will reinstate itself. The silence button will silence the horn at the Halon panel.

The panel abort buttons allow a delay of Halon discharge, so long as the button is depressed.

The system discharge switch allows release of the on service bank after the 60 second timer and will activate the fire alarm.

Page 66 of 107 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 3 Group # 1 K/A # 005 AK2.02 Importance Rating 2.5 Inoperable/Stuck Control Rod: Knowledge of the interrelations between the Inoperable/Stuck Control Rod and the following:

Breakers, relays, disconnects, and control room switches Proposed Question: Common 65 Given the following conditions on Unit 1:

• A Control Bank D, Group 1 Rod has been identified as not responding to an insertion or withdrawal signal.

• Initial actions required by ABN-712, Rod Control Malfunction have been performed.

• In trying to determine the ability of the affected rod to trip, I&C Technicians want to place the affected Control Rod on the Hold Bus.

Which ONE (1) of the following describes the limitations of placing Rods on the Hold Bus and the response of those Rods to a Reactor Trip?

A. Place a single Rod on the Hold Bus to ensure no more than one INOPERABLE Rod. In this condition, the Control Rod will NOT trip.

B. Limited to a single Rod Group due to power supply limitations and being continuously attended. In this condition, the Control Rods will NOT trip.

C. Limited to a single Rod Group due to power supply limitations and selection capabilities. In this condition, the Rod Group will trip upon a Reactor Trip.

D. Place either a single Rod or up to two (2) Rod Groups on the Hold Bus. In this condition, the Control Rods will trip upon a Reactor Trip.

Proposed Answer: C Page 67 of 107 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible because it could be thought that only a single rod could be placed on the Hold Bus since the Hold Bus stayed energized on a Reactor Trip, however, the Hold Bus can only have a Group selected and will deenergize on a Reactor Trip.

B. Incorrect. Plausible because only a Group can be selected and it could be thought that the Hold Bus stayed energized on a Reactor Trip, however, the Hold Bus will deenergize on a Reactor Trip.

C. Correct. Only Group selection capability exists and the Hold Bus is powered from the output of the Reactor Trip Breakers and will deenergize resulting in Rod insertion.

D. Incorrect. Plausible because the Hold Bus response on a Reactor Trip is correct and it could be thought that selection of single Rod or Group was available; however, only Group selection capability exists.

Technical Reference(s) SOP-702A, Step 5.3.1.F Attached w/ Revision # See OP51.SYS.CR1.LN, Figure 5, Page 24 Comments / Reference Proposed references to be provided during examination: None Learning Objective: DESCRIBE the basis for the precautions and limitations, and given major procedure steps relative to the Rod Control System, PLACE them in the proper sequence, for:

• SOP-702, Rod Control System (OP51.SYS.CR1.OB14)

STATE the physical connections and EVALUATE the cause-effect relationships between the Rod Control System and the following systems, components or events:

• Effect of stuck or misaligned rod (OP51.SYS.CR1.OB08)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 6 55.43 Page 68 of 107 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SOP-702A, Step 5.3.1.F Revision # 8 Page 69 of 107 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From OP51.SYS.CR1.LN, Figure 5, Page 24 Revision # 0 Figure 5 - Rod Control System Block Diagram Page 70 of 107 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 3 Group # 1 K/A # G 2.1.39 Importance Rating 3.6 Conduct of Operations: Knowledge of conservative decision making practices Proposed Question: Common 66 Which ONE (1) of the following actions is an application of the appropriate level of conservatism in decision-making in order to avoid unacceptable risk?

A. Recognize and reward conservative decisions made that do NOT impact Unit availability or generation capacity.

B. Follow procedures, they represent conservative decisions that have been made in advance.

C. Establish controls to ensure a decision to shutdown a Unit is NOT implemented until approved by the Plant Manager.

D. Place personnel in crisis atmospheres whenever possible to ensure ability to make a decision under pressure.

Proposed Answer: B Explanation:

A. Incorrect. Plausible because recognizing and rewarding conservative decisions is part of the Nuclear Policy Statement, however, not when it avoids securing the Unit.

B. Correct. As described in Operations Guideline 3.

C. Incorrect. Plausible if thought that prior approval by the Plant Manager was required in order to shutdown the Unit.

D. Incorrect. Plausible if thought that placing an individual in a crisis atmosphere would assist in performing conservative decisions.

Technical Reference(s) Nuclear Policy Statement 114 Attached w/ Revision # See Nuclear Policy Statement 208 Comments / Reference Operations Guideline 3, Section 1.2 Proposed references to be provided during examination: None Learning Objective: When faced with uncertain or degrading conditions, EXPLAIN management expectations regarding conservative operational decisions, including removing the main turbine from service and initiating a manual reactor trip.

(OPD8.CDM.IR1.OB01)

Page 71 of 107 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Question Source: Bank #

Modified Bank # (Note changes or attach parent)

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

Reference:

From Nuclear Policy Statement 114 Revision # 3 Ensuring that licensed plant operators have the authority and responsibility to shut a unit down when they believe that shutdown is required or warranted to protect the safety of operations.

Comments /

Reference:

From Nuclear Policy Statement 208 Revision # 1 It is important that capability, such as training and experience, are matched with authority when making a decision.

Managers and supervisors should encourage plant ownership, professionalism, leadership, and a strong safety culture; but, should not place personnel in decision-making situations they do not have the capability or authority to make. Managers and supervisors should:

• Be accountable for the decisions from their organizations.

• Monitor and mentor new employees on decision-making and be cautious of their can do attitude.

• Recognize and reward conservative decision-making made by individuals regardless of outcome; consider this: It was the best decision to make at the time.

• Avoid placing personnel in situations involving a crisis atmosphere.

• Ensure the decision-making authority that is delegated is clear and specific to the individuals and not just globally inferred by signature authority.

• Develop written procedures, guidelines, decision trees, etc., for routine decisions made by an organization. Be sure these processes and the potential implications from the decisions made within them are well understood.

• Ensure personnel understand the difference between routine decisions and decisions where management / supervision (or other expertise) should be involved.

• Encourage a healthy questioning attitude in the people in a department.

Nuclear personnel should:

• Follow procedures. They represent conservative decisions that have been made in advance for us to follow.

• STOP if you are ever unsure that the procedure is adequate for the circumstances you encounter or if you are unsure that you have the capability and the authority to make a conservative decision. Then, notify your Supervisor.

• Involve the proper level of management in decisions. If you are not sure, ask.

• Involve other stakeholders in decisions affecting them with enough time for them to influence the decision.

Page 72 of 107 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From Operations Guideline 3, Section 1.2 Revision # 09/17/09 1.2 Conservative Decision-Making Each Operator has the responsibility for exercising conservative judgment to ensure that safety and quality take precedence over unit availability and generating capacity and that long term availability is considered more carefully than short term generating capacity.

Page 73 of 107 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 3 Group # 1 K/A # G 2.1.29 Importance Rating 4.1 Conduct of Operations: Knowledge of how to conduct system lineups, such as valves, breakers, switches, etc.

Proposed Question: Common 67 Which ONE (1) of the following methods is used to verify that a valve is LOCKED CLOSED per STA-694, Station Verification Activities?

A. Remove the lock, turn the valve in the CLOSE direction, and reinstall the locking device.

B. Remove the lock, turn the valve at least 1/4 turn in the OPEN direction, close the valve, and reinstall the locking device.

C. Verify the locking device is installed, attempt to turn the valve in the OPEN direction, close the valve, and observe valve position indication.

D. Verify the locking device is installed, turn the valve in the CLOSE direction, and conduct a visual verification of the stem position.

Proposed Answer: D Explanation:

A. Incorrect. Plausible because the valve is turned in the CLOSE direction, however, when the valve is already locked the locking mechanism must remain in position.

B. Incorrect. Plausible if thought that unlocking the valve and opening it 1/4 turn was verification that the valve was closed.

C. Incorrect. Plausible because the locking device must be verified installed and valve position indication could be verified, however, the valve is checked in the CLOSE direction.

D. Correct. Per STA-694, Attachment 8.B, Guideline on Component Verification of Operational Activities.

Technical Reference(s) STA-694, Attachment 8.B Attached w/ Revision # See Comments / Reference Proposed references to be provided during examination: None Learning Objective: Given that valve position verification is required, DESCRIBE the actions required to check a manual valve LOCKED CLOSED in accordance with STA-694. (OPD1.ADM.XAD.OB04)

Page 74 of 107 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Question Source: Bank # ADM.XAD.OB04-1 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 75 of 107 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From STA-694, Attachment 8.B Revision # 5 3.1 Valve to be checked "OPEN" For manual valves, the valve operator should be moved a small amount in the closed direction until the valve stem moves, indicating the valve is open. It should then be moved in the open direction until the valve is considered to be in its fully open position. If the valve is not an instrument isolation, the valve should then be turned at least 1/4 turn, but not more than one turn, in the closed direction. Because of instrumentation valve construction, these valves should not be turned more than 1/4 turn in the closed direction. A visual Verification should be made of the valve stem, local position indicator or any other valve component suitable for Verification, if possible.

Valves should not be routinely backseated. Valves that are designed with a backseat may be backseated if approved by the Unit Supervisor.

3.2 Valve to be checked "CLOSED" For manual valves, move the valve operator in the closed direction only. If the valve is in the correct closed position, no motion will occur (avoid over torquing). A visual Verification should then be made of the valve stem, local position indicator or any other valve component suitable for Verification, if possible.

The operator should attempt to ensure that the valve is not just binding or difficult to operate.

3.3 Valve to be checked "LOCKED OPEN" For locked or sealed valves, the lock does not need to be removed. The actual position of the valves and locking devices should be determined by physical contact between the valve and locking device. Try to move the valve in the closed direction to determine that the locking device does keep the valve open. A visual Verification should then be made of the valve stem, local position indicator or other valve component suitable for Verification, if possible.

3.4 Valves to be checked "LOCKED CLOSED" The actual position of the valves and locking devices should be determined by physical contact between the valve and the locking device. Check the valve closed by moving the valve in the closed direction only.

A visual Verification should then be made of the valve stem, local position indicator or any other valve component suitable for Verification, if possible.

3.5 Valves to be checked "THROTTLED" or "LOCKED THROTTLED" with redundant Visual Indication These valves should not be moved unless specifically authorized by the Unit Supervisor. Visual Verification using valve stem position, local indicators, system flow indication, scribe marks or any other valve component suitable for Verification should be used as necessary to verify position, if possible.

Verification by observing Control Room instruments, annunciators, valve position indicators, etc., may be acceptable as long as the Control Room indication is a positive one and is directly observed and documented.

Page 76 of 107 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 3 Group # 1 K/A # G 2.1.19 Importance Rating 3.9 Conduct of Operations: Ability to use plant computers to evaluate system or component status Proposed Question: Common 68 Given the following conditions:

• OPT-114, Digital Radiation Monitoring System trending is in progress.

Which ONE (1) of the following identifies the ACCEPTANCE CRITERIA for each observed Radiation Monitor?

NO unexplained changes greater than a factor of _____ OR any abnormal readings.

A. 2 B. 10 C. 20 D. 100 Proposed Answer: B Explanation:

A. Incorrect. Plausible if thought that doubling was an acceptable factor.

B. Correct. The Digital Radiation Monitoring System is a Plant Computer commonly referred to as PC-11. This is the ACCEPTANCE CRITERIA when performing Radiation Monitoring trending as stated in OPT-114.

C. Incorrect. Plausible given Radiation Monitor output during a trend.

D. Incorrect. Plausible if factor is confused with count rate.

Technical Reference(s) OPT-114, Step 2.1.1 Attached w/ Revision # See Comments / Reference Proposed references to be provided during examination: None Learning Objective: DESCRIBE the basis for the precautions and limitations, and given major procedure steps relative to the Digital Radiation Monitoring system, PLACE them in the proper sequence, for:

• OPT-114, DRMS Trending (OP51.SYS.RM1.OB12)

Page 77 of 107 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Question Source: Bank # SYS.RM1.OB12-1 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, 11 55.43 Comments /

Reference:

From OPT-114, Step 2.1.1 Revision # 7 Page 78 of 107 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 3 Group # 2 K/A # G 2.2.12 Importance Rating 3.7 Equipment Control: Knowledge of surveillance procedures Proposed Question: Common 69 Which ONE (1) of the following identifies the method for reestablishing blocks following completion of the OPT-447A, Solid State Protection System Surveillance Test?

Upon completion of OPT-447A, Solid State Protection System Surveillance Test, reestablish all the appropriate blocks...

A. prior to placing the Permissive and Memories Switches in OFF and after placing the Input Error Inhibit Switch in NORMAL.

B. after placing the Input Error Inhibit Switch in NORMAL and after placing the Permissive and Memories Switches in OFF.

C. after placing the Permissive and Memories Switches in OFF and prior to placing the Input Error Inhibit Switch in NORMAL.

D. prior to placing the Input Error Inhibit Switch in NORMAL and prior to placing the Permissive and Memories Switches in OFF.

Proposed Answer: C Explanation:

A. Incorrect. Plausible if thought that this order would restore all the appropriate blocks.

B. Incorrect. Plausible if thought that this order would restore all the appropriate blocks.

C. Correct. Per OPT-447A, this is the order of switch placement to restore all the appropriate blocks.

D. Incorrect. Plausible because it must be done prior to placing the Input Error Inhibit Switch in NORMAL, however, the Permissive and Memories Switches must be placed in OFF first.

Page 79 of 107 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Technical Reference(s) OPT-447A, Attachment 10.1.5 Attached w/ Revision # See Comments / Reference Proposed references to be provided during examination: None Learning Objective: DETERMINE why all the blocks must be reinstated prior to placing the Input Error Inhibit switch in NORMAL. (OP51.SYS.ES2.OB19)

Question Source: Bank # SYS.ES2.OB19-1 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, 10 55.43 Page 80 of 107 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From OPT-447A, Attachment 10.1.5 Revision # 7 Page 81 of 107 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From OPT-447A, Attachment 10.1.5 Revision # 7 Page 82 of 107 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 3 Group # 2 K/A # G 2.2.25 Importance Rating 3.2 Equipment Control: Knowledge of the bases in Technical Specifications for limiting conditions for operations and safety limits Proposed Question: Common 70 Given that the Unit is in MODE 2:

Which ONE (1) of the following lists of components and protection signals are designed to ensure the Reactor Coolant System pressure Technical Specification Safety Limit of less than or equal to 2735 psig is maintained?

A. Pressurizer Safety Valves, Steam Generator Atmospheric Relief Valves and Pressurizer High Pressure Trip.

B. Pressurizer Safety Valves, Main Steam Safety Valves and Pressurizer High Pressure Trip.

C. Pressurizer Safety Valves, Steam Generator Atmospheric Relief Valves and Pressurizer High Water Level Trip.

D. Pressurizer Safety Valves, Main Steam Safety Valves and Pressurizer High Water Level Trip.

Proposed Answer: B Explanation:

A. Incorrect. Plausible because the PRZR Safeties and High Pressure Trip maintain pressure below 2735 psig; however, no credit is taken for the ARVs.

B. Correct. Pressurizer Safety Valves, Main Steam Safety Valves and Pressurizer High Pressure Trip settings established to ensure RCS pressure Safety Limit will not be exceeded.

C. Incorrect. Plausible because the Pressurizer Safety Valves are part of the RCS pressure protection bases but no credit is taken for the ARVs and the PRZR High Level Trip is a backup to the PRZR High Pressure Trip but only after power is above 10% and Stem states the Unit is in MODE 2.

D. Incorrect. Plausible because the PRZR High Water Level Trip is a backup to the PRZR High Pressure Trip but only after power is above 10% and Stem states the Unit is in MODE 2.

Technical Reference(s) Technical Specification 2.0 Safety Limits Attached w/ Revision # See Technical Specification Safety Limit Bases Comments / Reference Proposed references to be provided during examination: None Learning Objective: EXPLAIN the bases for the Safety Limits. (LO21.RLS.SL1.OB06)

Page 83 of 107 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Page 84 of 107 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 Comments /

Reference:

From Technical Specification 2.0 Safety Limits Amendment # 67 Page 85 of 107 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From Technical Specification Safety Limit Bases Revision # 51 Page 86 of 107 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From Technical Specification Safety Limit Bases Revision # 51 Page 87 of 107 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 3 Group # 3 K/A # G 2.3.4 Importance Rating 3.2 Radiation Control: Knowledge of radiation exposure limits under normal or emergency conditions Proposed Question: Common 71 Given the following condition:

• A 20 year old CPNPP employee has 1500 mRem of TEDE exposure for 2009.

Which ONE (1) of the following describes the MAXIMUM additional amount of 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 additional amount of TEDE exposure that may be received prior to exceeding 10CFR20 (NRC) exposure limits?

A. 500 mRem; 3500 mRem B. 2500 mRem; 3500 mRem C. 500 mRem; 8500 mRem D. 2500 mRem; 8500 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; 10CFR20 limit is 5000 mrem per year.

C. Incorrect. Plausible because the Admin limit is 2000 mrem, however, the 2nd part is only correct if thought that the 10CFR20 5(N-18) rule was still applicable.

D. Incorrect. Plausible because the Admin limit previously was 4000 mrem, however, the 2nd part is only correct if thought that the 10CFR20 5(N-18) rule was still applicable.

Technical Reference(s) STA-655, Attachment 8.A Attached w/ Revision # See STA-655, Step 5.1.1 Comments / Reference STA-655, Step 6.4.3 Proposed references to be provided during examination: None Page 88 of 107 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Learning Objective: STATE the annual exposure limits for Total Effective Dose Equivalent (TEDE),

Total Organ Dose Equivalent (TODE), Eye Dose Equivalent (EDE) and Shallow Dose Equivalent (SDE). (PO11.FND.RE1.OB01)

Question Source: Bank #

Modified Bank # STA.655.OB-1 (Note changes or attach parent)

New Question History: Last NRC Exam CPNPP 2007 Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10, 12 55.43 Page 89 of 107 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From STA-655, Step 5.1.1 Revision # 18 Page 90 of 107 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From STA-655, Step 6.4.3 Revision # 18 Page 91 of 107 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From STA-655, Attachment 8.A Revision # 18 Comments /

Reference:

From STA-655, Attachment 8.A Revision # 18 Page 92 of 107 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From Exam Bank STA.655.OB-1 Revision # 0 Given the following condition:

• A 20 year old CPSES employee has 1750 mRem of TEDE exposure for 2009.

Which ONE (1) of the following describes the MAXIMUM additional amount of TEDE exposure they may receive without additional CPSES authorization (i.e., Plant Manager approval), and what is the MAXIMUM additional amount of TEDE exposure he may receive prior to exceeding 10CFR20 (NRC) exposure limits?

A. 2250 mRem; 3250 mRem B. 2250 mRem; 8250 mRem C. 1250 mRem; 3250 mRem D. 1250 mRem; 8250 mRem Page 93 of 107 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 3 Group # 3 K/A # G 2.3.5 Importance Rating 2.9 Radiation Control: Ability to use radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc.

Proposed Question: Common 72 Given the following conditions:

• A Portable Frisker is being used to perform a whole body frisk.

• Background radiation is at 150 counts per minute.

Which ONE (1) of the following is the CONTINUOUS count rate (background + actual) at which an individual is considered to be contaminated?

A. 175 counts per minute B. 200 counts per minute C. 225 counts per minute D. 250 counts per minute Proposed Answer: D Explanation:

A. Incorrect. Plausible if actual level of 100 cpm above background can not be recalled.

B. Incorrect. Plausible if actual level of 100 cpm above background can not be recalled.

C. Incorrect. Plausible if actual level of 100 cpm above background can not be recalled.

D. Correct. With a background radiation of 150 cpm + a detected radiation level of 100 cpm above background = 250cpm.

Technical Reference(s) STA-653, Step 6.6.2 & Attachment 3 Attached w/ Revision # See Comments / Reference Proposed references to be provided during examination: None Learning Objective: NANTeL Generic RWT Objective: EXPLAIN how to monitor personnel and personal items for contamination, including the use of friskers and personnel contamination monitors.

Question Source: Bank #

Modified Bank # ADM.RP.OB00-2 (Note changes or attach parent)

Page 94 of 107 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 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, 12 55.43 Page 95 of 107 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From STA-653, Step 6.6.2 Revision # 12 Page 96 of 107 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From STA-653, Attachment 3 Revision # 12 Page 97 of 107 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From Exam Bank Audit.Exam.OB00-2 Revision # 0 Given the following conditions:

• A Portable Frisker is being used to perform a whole body frisk.

• Background radiation is at the MAXIMUM allowed level for using a frisker.

Which ONE (1) of the following is the CONTINUOUS count rate (background + actual) at which an individual is considered to be contaminated?

A. 100 counts per minute B. 200 counts per minute C. 300 counts per minute D. 400 counts per minute Page 98 of 107 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 3 Group # 4 K/A # G 2.4.20 Importance Rating 4.8 Emergency Procedures/Plan: Knowledge of the operational implications of EOP warnings, cautions, and notes Proposed Question: Common 73 Given the following conditions:

• Unit 1 has tripped from 100% power with subcooling approaching 210ºF.

• FRP-0.1A, Response to Imminent PTS Condition has been entered.

• Step 2, CAUTION reads as follows:

• If the TDAFW pump is the only available source of feed flow, steam supply to the turbine-driven AFW pump must be maintained from one SG.

Which ONE (1) of the following is the reason for maintaining the steam supply to the Turbine Driven Auxiliary Feedwater Pump?

A. The cooldown rate contribution of steam supplied to the Auxiliary Feedwater Pump would not be noticeable.

B. Large subcooling in the primary could worsen if the Steam Generator tubes become uncovered and/or break due to lack of feedwater.

C. If the Motor Driven Auxiliary Feedwater Pumps are not available then this would potentially require a transfer to a higher priority ERG.

D. Components impacting cooldown are removed from service in order of their effect with the Turbine Driven Auxiliary Feedwater Pump being last.

Proposed Answer: C Explanation:

A. Incorrect. Plausible because the cooldown rate impact is minimal, however, this is not the basis for the CAUTION.

B. Incorrect. Plausible if thought that a large subcooled condition would be made worse due to the depressurization associated with a tube break, however, this procedure takes numerous steps to depressurize and minimize subcooling. Therefore, subcooling would improve.

C. Correct. With no Motor Driven Auxiliary Feed Water Pumps, a loss of the TDAFW Pump could result in a transition to FRH-0.1A, Response to Loss of Secondary Heat Sink.

D. Incorrect. Plausible because these actions are carried out in Step 2 RNO actions of FRP-0.1A, however, this is not the basis for the CAUTION.

Page 99 of 107 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Technical Reference(s) FRP-0.1A, Step 2 Caution Attached w/ Revision # See FRP-0.1A, Attachment 4, Bases Comments / Reference Proposed references to be provided during examination: None Learning Objective: Given a NOTE or CAUTION from FRP-0.1A/B, STATE the reason for the NOTE or CAUTION. (LO21.ERG.FP1.OB04)

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 100 of 107 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From FRP-0.1A, Step 2 Caution Revision # 8 Page 101 of 107 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From FRP-0.1A, Attachment 4, Bases Revision # 8 Page 102 of 107 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 3 Group # 4 K/A # G 2.4.13 Importance Rating 4.0 Emergency Procedures/Plan: Knowledge of crew roles and responsibilities during EOP usage Proposed Question: Common 74 All deviations from the Emergency Response Guidelines shall be made with approval by

__________ and should receive concurrence from __________.

A. a Senior Reactor Operator; the Shift Manager B. a Senior Reactor Operator; the Director, Operations C. the Shift Manager; the Plant Manager D. the Emergency Coordinator; Site Vice President Proposed Answer: A Explanation:

A. Correct. As outlined in ODA-407, Attachment 8.A for ERG Rules of Usage, a licensed Senior Reactor Operator approves the change and it is reviewed by another SRO individual preferably the Shift Manager.

B. Incorrect. Plausible because a Senior Reactor Operator approves the change and concurrence should be received from the Shift Manager, and the Director, Operations must hold an SRO license and this may not be the case.

C. Incorrect. Plausible because the Shift Manager could fulfill the function of the Senior Reactor Operator, however, the Plant Manager must hold an SRO license and this may not be the case.

D. Incorrect. Plausible because unless the Emergency Coordinator duties are turned over, this individual would be the Shift Manager who holds an SRO license, however, the Site Vice President would not necessarily meet the SRO license requirement.

Technical Reference(s) ODA-407, Attachment 8.A Attached w/ Revision # See Comments / Reference Proposed references to be provided during examination: None Learning Objective: DESCRIBE the requirements associated with deviating from an ERG.

(LO21.ERG.XD2.OB28)

Question Source: Bank # ERG.XD2.OB128 Modified Bank # (Note changes or attach parent)

New Page 103 of 107 Rev e

ES-401 CPNPP Mar 2010 NRC 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 ODA-407, Attachment 8.A Revision # 12 Page 104 of 107 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 3 Group # 4 K/A # G 2.4.5 Importance Rating 3.7 Emergency Procedures/Plan: Knowledge of the organization of the operating procedures network for normal, abnormal and emergency evolutions Proposed Question: Common 75 Which ONE (1) of the following procedure groups comprises the Functional Restoration Guidelines at Comanche Peak?

A. EOS, FRG, and CSF Status Trees.

B. ECA, EOS, and FRG.

C. FRG and CSF Status Trees.

D. EOP, ECA, and FRG.

Proposed Answer: C Explanation:

A. Incorrect. Plausible because FRG and CSF Status Tree procedures are correct, however, the EOS procedures are part of the Optimal Recovery Guidelines.

B. Incorrect. Plausible because FRG procedures are correct, however, the EOS and ECA procedures are part of the Optimal Recovery Guidelines.

C. Correct. These are the two sets of procedures that make up the Functional Restoration Guidelines.

D. Incorrect. Plausible because the FRG procedures are correct, however, the EOP and ECA procedures are part of the Optimal Recovery Guidelines.

Technical Reference(s) LO21.ERG.XG1.LN, Pages 13 & 15 Attached w/ Revision # See Comments / Reference Proposed references to be provided during examination: None Learning Objective: LIST and DIFFERENTIATE between the three types of Optimal Recovery Guidelines. (LO21.ERG.XG1.OB04)

STATE the primary function of the Functional Restoration Guides.

(LO21.ERG.XG1.OB07)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Page 105 of 107 Rev e

ES-401 CPNPP Mar 2010 NRC 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 LO21.ERG.XG1.LN, Page 15 Revision # 0 FUNCTION RESTORATION GUIDELINES Guidance for restoration of the plant safety state independent of event sequence is contained in the Function Restoration Guidelines (FRGs). Like the ORGs, they are symptom-based strategies for predefined events. Thus, they are developed using risk assessment and available industry experience.

As with the ORGs, FRGs have been developed for functional failures with a frequency of occurrence of over 10-8 per reactor year. They are similar in format to the ORGs.

The FRG strategies use available equipment to restore safety functions to a satisfied condition from which optimal recovery can be initiated, or continued. They are organized into six basic categories, consistent with the Critical Safety Functions. Each category contains two to five guidelines (FRGs),

depending on the number of status tree termini. The basic categories are summarized below:

SUBCRITICALITY - addresses symptoms associated with an anticipated transient without trip, a return to power condition and a loss of core shutdown condition.

CORE COOLING - addresses symptoms associated with inadequate, degraded and saturated core cooling conditions due to depletion of water in the RCS.

HEAT SINK- addresses symptoms associated with problems concerning secondary heat sink including the following steam generator conditions: 1) high level, 2) low level, 3) overpressure and 4) loss of normal steam release capability.

INTEGRITY - addresses symptoms associated with imminent and anticipated pressurized thermal shock conditions and cold overpressure conditions.

CONTAINMENT- addresses symptoms associated with potential containment overpressure (including hydrogen), flooding, and high radiation level.

INVENTORY- addresses symptoms associated with high pressurizer level, low pressurizer level, and void in the reactor vessel.

A list of FRGs for CPSES is contained in Table 2.

Entry into a FRG is typically from the associated status tree, but may also be directed by a step in the ORG. Following completion of function restoration, the FRG directs the operator to the appropriate ORG for resumption of optimal recovery.

Page 106 of 107 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From LO21.ERG.XG1.LN, Page 13 Revision # 0 The ORGs within each series of guidelines are subdivided into three different types:

EOP guidelines (entry guidelines)

EOS guidelines (sub-guidelines)

ECA guidelines (emergency contingency actions)

This organization results in an entry guideline for each series with associated sub-guidelines and emergency contingency action guidelines. EOSs provide alternate recovery strategies within the event category. ECAs supplement both the entry and sub-guidelines by providing guidance for low probability or unique events. Use of ECAs allows these less probable events to be addressed without unduly complicating the EOP and EOS guidelines.

CPSES Optimal Recovery Guidelines are identified in Table 1.

Page 107 of 107 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 008 G 2.4.21 Importance Rating 4.6 Pressurizer Vapor Space Accident: Emergency Procedures / Plan: Knowledge of the parameters and logic used to assess the status of safety functions, such as reactivity control, core cooling and heat removal, reactor coolant system integrity, containment conditions, radioactivity release control, etc.

Proposed Question: SRO 76 Given the following conditions on Unit 1:

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

• All required Safety System actuations have occurred as designed.

• EOS-1.2A, Post LOCA Cooldown and Depressurization it is being implemented.

• A cooldown is in progress.

• Subcooling is 15ºF and slowly dropping.

• Pressurizer level is 95% and rising.

• The flange light above 49 inches on the Reactor Vessel Level Indication System is NOT lit.

• The Shift Technical Advisor notifies you that the Inventory Safety Function is Yellow based on Pressurizer Level and Reactor Vessel Level.

Which ONE (1) of the following describes the cause of this condition and the actions that should be directed to mitigate this condition?

A. A Pressurizer Steam Space leak is the cause. As long as Emergency Core Cooling Pumps are operating stay in EOS-1.2A, Post LOCA Cooldown and Depressurization and continue efforts to restore subcooling.

B. The Cooldown and Depressurization has resulted in voiding in the Reactor Vessel.

Direct entry into FRI-0.3A, Response to Voids In Reactor Vessel for the Yellow Condition in Inventory and perform the actions for Reactor Vessel Head venting.

C. A Pressurizer Steam Space leak is the cause. Direct entry into FRI-0.1A, Response to High Pressurizer Level to establish Charging and Letdown and energize Pressurizer Heaters to restore subcooling.

D. The Reactor Coolant System cooldown is being limited because RCPs have been secured. Direct entry into EOS-0.3A, Natural Circulation Cooldown with Steam Void in Vessel (With RVLIS) to monitor void growth during the cooldown.

Proposed Answer: A Page 1 of 119 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Explanation:

A. Correct. A break on the steam space of the Pressurizer will result in inventory loss and a loss of the steam bubble in the Pressurizer resulting in a voiding in the Reactor Vessel. With Emergency Core Cooling System operation still required, the Yellow Path to FRI-0.3A, Response to Voids In Reactor Vessel would immediately direct the crew back to the procedure and step in affect in the very first step because a more serious condition is indicated.

B. Incorrect. Plausible because the conditions causing these indications is correct, however, the action to go to FRI-0.3A, Response to Voids In Reactor Vessel, is inappropriate because it would direct the crew immediately back to the procedure and step in affect in the very first step based on Emergency Core Cooling Systems still in operation.

C. Incorrect. Plausible because a steam space leak in the Pressurizer is indicated and it could be thought that the correct action would be to respond to the high Pressurizer level condition, however, the real issue is still subcooling and the proper action would be to stay in the current procedure and restore subcooling. FRI-0.1A would direct returning to the procedure and step in affect in Step #1 if Emergency Core Cooling Systems were still operating.

D. Incorrect. Plausible because it could be thought that due to no forced cooling a Natural Circulation cooldown that allows for void growth would be appropriate but EOS-0.3A should only be used when there is no Safety Injection.

Technical Reference(s) FRI Critical Safety Function Status Tree Attached w/ Revision # See FRI-0.3A, Attachment 7, Step 1 Comments / Reference EOS-0.3A, Purpose Proposed references to be provided during examination: None Learning Objective: Given a set of plant conditions, IDENTIFY the proper transitions through/out of EOS-1.2. (LO21.ERG.E12.OB06)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 5 Page 2 of 119 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From FRI Critical Safety Function Status Tree Revision # 8 Page 3 of 119 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From FRI-0.3A, Attachment 7, Step 1 Revision # 8 Comments /

Reference:

From EOS-0.3A, Purpose Revision # 8 Page 4 of 119 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 011 EA2.01 Importance Rating 4.7 Large Break LOCA: Ability to determine and interpret the following as they apply to a Large Break LOCA: Actions to be taken, based on RCS temperature and pressure - saturated or superheated Proposed Question: SRO 77 The Critical Safety Function Status Trees are being reviewed in order to verify the CORE COOLING Critical Safety Function during a Large Break Loss of Coolant Accident. Plant conditions are as follows:

• Containment pressure is 16 psig and stable.

• Reactor Coolant System Saturation Margin indicates 300ºF SUPERHEAT.

• RVLIS indication 11 inches above the Core Plate is NOT lit.

• Core Exit Thermocouples are 770ºF and stable.

Which ONE (1) of the following procedures would be used to mitigate the CORE COOLING Critical Safety Function and what is the basis for that action?

Enter procedure...

A. FRC-0.2, Response to Degraded Core Cooling due to RCS subcooling less than 55ºF and Core Exit Thermocouples greater than 750ºF.

B. FRC-0.3, Response to Saturated Core Cooling due to RCS subcooling less than 55ºF and Core Exit Thermocouples less than 1200ºF.

C. FRC-0.2, Response to Degraded Core Cooling due to RCS subcooling less than 55ºF and RVLIS indication at 11 inches above the Core Plate NOT lit.

D. FRC-0.3, Response to Saturated Core Cooling due to RCS subcooling greater than 55ºF and RVLIS indication at 11 inches above the Core Plate NOT lit.

Proposed Answer: A Page 5 of 119 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Explanation:

A. Correct. Given that Adverse Containment conditions exist then RCS subcooling must be >55ºF.

With CET temperature greater than 750ºF but less than 1200ºF and RVLIS indication at 11 inches above the Core Plate NOT lit, entry into FRC-0.2A, Response to Degraded Core Cooling is correct.

B. Incorrect. Plausible if thought that both RCS subcooling and Core Exit Thermocouples temperatures were being met. In that case, entry into FRC-0.3A, Response to Saturated Core Cooling would be correct if RVLIS indication was ignored.

C. Incorrect. Plausible because procedure entry is correct and RCS subcooling does not meet Adverse Containment conditions, however, RVLIS indication must be taken into account otherwise the CSFST would direct entry into FRC-0.3A, Response to Saturated Core Cooling.

D. Incorrect. Plausible because Core Exit Temperature is greater than 750ºF and Adverse Containment conditions exist, however, subcooling is not met and FRC-0.2A entry is required not FRC-0.3A, Response to Saturated Core Cooling.

Technical Reference(s) FRC-0.2A, CSFST Attached w/ Revision # See Comments / Reference Proposed references to be provided during examination: None Learning Objective: ANALYZE the indications and DESCRIBE the mitigation strategy and major steps taken relative to the Core Cooling Monitor System/RVLIS, both initial and subsequent, for:

• FRC-0.1, Response to Inadequate Core Cooling

• FRC-0.2, Response to Degraded Core Cooling (OP51.SYS.RC3.OB13)

Question Source: Bank # ERG.XD2.OB16-3 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 6 of 119 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From FRC-0.2A, CSFST Revision # 8 Page 7 of 119 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 015/017 G 2.2.40 Importance Rating 4.7 RCP Malfunctions: Equipment Control: Ability to apply Technical Specifications for a system Proposed Question: SRO 78 Given the following conditions:

• Unit 1 is in MODE 3 at 385ºF and 385 psig following a Refueling outage.

• Reactor Coolant Pump (RCP) 1-02 is tagged out for an oil leak repair.

• A common mode failure caused Non-Safeguards 6.9 kV Buses 1A1 and 1A3 to lockout.

• Non-Safeguards 6.9kV Bus 1A4 is energized with Reactor Coolant Pump 1-04 RUNNING.

• Both Trains of Residual Heat Removal are in the ECCS Standby lineup.

• Both Control Rod Drive Motor Generators are tagged out.

Which ONE (1) of the following identifies the Technical Specification REQUIRED ACTION given the conditions listed?

A. Immediately place the Unit in MODE 4.

B. Place both Residual Heat Removal Loops in service.

C. Initiate actions to restore a second RCS Loop to OPERABLE status.

D. Immediately isolate all dilution flow paths.

Proposed Answer: C Page 8 of 119 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible because this is the REQUIRED ACTION of Technical Specification LCO 3.4.5, CONDITION B, however, the COMPLETION TIME is 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

B. Incorrect. Plausible if thought that the conditions were being met to place RHR in service.

C. Correct. Technical Specification LCO 3.4.5 Action A requires restoring an additional RCS loop to OPERABLE status in 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

D. Incorrect. Plausible because this would be required if the Unit was in MODE 4 and both Loops were INOPERABLE.

Technical Reference(s) Technical Specification LCO 3.4.5 Attached w/ Revision # See Technical Specification LCO 3.4.6 Comments / Reference Proposed references to be provided during examination: None Learning Objective: Given a Technical Specification or a Technical Specification situation, DIAGNOSE the situation and APPLY the LCO and SR Applicability of Section 3.0 to DETERMINE all corrective actions. (LO21.RLS.SL1.OB12)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 2 Page 9 of 119 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From Technical Specification LCO 3.4.5 Amendment # 105 Page 10 of 119 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From Technical Specification LCO 3.4.5 Amendment # 105 Page 11 of 119 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From Technical Specification LCO 3.4.6 Amendment # 105 Page 12 of 119 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 025 G 2.1.7 Importance Rating 4.7 Loss of RHR System: Conduct of Operations: Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation Proposed Question: SRO 79 Given the following conditions with Unit 2 in MODE 5:

• 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> has elapsed since the Unit was shutdown.

• The Reactor Coolant System was being drained to install Steam Generator nozzle dams and Pressurizer level was at 5% Cold Calibrated level.

• Both Trains of Residual Heat Removal were in service for cooling and draining but currently are unavailable.

• Both Pressurizer Power Operated Relief and Block Valves are open.

• The Reactor Coolant System was filled per ABN-104, Residual Heat Removal System Malfunction using a Hot Leg Injection Path and secured at approximately 25%

Pressurizer Cold Calibrated level.

• Maximum Blowdown and Auxiliary Feedwater have been aligned on two OPERABLE Steam Generators and the associated Atmospheric Relief Valves are full open.

• The highest Core Exit Thermocouple currently reads 210ºF.

Which ONE (1) of the following actions should be performed?

A. Alternate raising the level in the OPERABLE Steam Generators to just below full scale. Repeat this action when level lowers to 10% wide range.

B. Align Letdown and re-establish Hot Leg Injection Flow at the maximum flow to maintain Pressurizer level.

C. Stop Blowdown and Auxiliary Feedwater. Allow the Reactor Coolant and Steam Generator to heat up until steaming is achieved via the Atmospheric Relief Valve.

D. Reinitiate Hot Leg Injection flow per ABN-104, Residual Heat Removal System Malfunction to remove decay heat.

Proposed Answer: D Page 13 of 119 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible because it could be thought that this would be a viable solution, however, this path is utilizing a small capacity line to remove inventory and would not provide the required flow for cooling and the procedure directs a once-thru cooling approach using Hot Leg Injection.

B. Incorrect. Plausible because it could be thought that this would be a viable solution, however, this is another ineffectual path based on insufficient pressure to drive any significant Letdown flow.

C. Incorrect. Plausible because it could be thought that this would be a viable solution, however, stopping the only in service path for heat removal, no matter how effective, is not normally an option to consider. The procedure requires a once-thru cooling approach using Hot Leg Injection.

D. Correct. The procedure requires a once-thru cooling approach using Hot Leg Injection which is a proven viable heat removal mechanism.

Technical Reference(s) ABN-104, Step 8.3.9 Attached w/ Revision # See ABN-104, Steps 8.3.12 thru 8.3.17 Comments / Reference Proposed references to be provided during examination: None Learning Objective: IDENTIFY indications for system operating parameters, met as a result of a RHR malfunction and /or a shutdown loss of coolant event, which are entry level conditions for Technical Specifications, ASSESS the limiting conditions for operations and safety limits which may be entered, and APPLY TS action statements which require response within one hour or less.

(LO21.ABN.104.OB07)

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 119 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ABN-104, Step 8.3.8 & 8.3.9 Revision # 8 Page 15 of 119 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Page 16 of 119 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ABN-104, Step 8.3.12 - 8.3.17 Revision # 8 Page 17 of 119 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Page 18 of 119 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 055 G 2.1.20 Importance Rating 4.6 Station Blackout: Conduct of Operations: Ability to interpret and execute procedure steps Proposed Question: SRO 80 Given the following conditions:

• A Station blackout has been in progress for several hours.

• Unit 1 Core Exit Thermocouple temperatures are 1220ºF and rising.

Which ONE (1) of the following identifies the actions to be taken given the conditions listed?

A. Remain in ECA-0.0A, Loss of All AC Power and attempt to restore power to any AC Safeguards Bus.

B. Transition to SACRG-1, Severe Accident Control Room Guideline Initial Response and verify a General Emergency has been declared.

C. Transition to ABN-601, Response to a 138/45 KV System Malfunction and attempt to restore power to any AC Bus.

D. Transition to ECA-0.2A, Loss of All AC Power Recovery With SI Required and manually align Safety Injection Valves in preparation for power restoration.

Proposed Answer: B Explanation:

A. Incorrect. Plausible because this action would continuously be performed, however, once Core Exit Thermocouple temperatures exceed 1200ºF SACRG-1, Severe Accident Control Room Guideline must be implemented.

B. Correct. With Core Exit Thermocouple temperatures greater than 1200ºF, SACRG-1, Severe Accident Control Room Guideline must be implemented.

C. Incorrect. Plausible because this action would be performed, however, it is performed at Step 9 of the ECA-0.0A. Given Core Exit Thermocouple temperature, the Severe Accident Control Room Guideline must be implemented.

D. Incorrect. Plausible because entry into this procedure would be required, however, not until power to at least one AC Safeguards Bus was restored.

Technical Reference(s) ECA-0.0A, Step 24 Attached w/ Revision # See ECA-0.0A, Flow Chart Comments / Reference Proposed references to be provided during examination: None Page 19 of 119 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Learning Objective: Given a major action step, NOTE or CAUTION of ECA 0.0, STATE the basis for the step, NOTE or CAUTION. (LO21.ERG.C00.OB11)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

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

Reference:

From ECA-0.0A, Step 24 Revision # 8 Page 20 of 119 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ECA-0.0A, Flow Chart Revision # 8 Page 21 of 119 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 026 G 2.2.22 Importance Rating 4.7 Loss of Component Cooling Water: Equipment Control: Knowledge of limiting conditions for operations or safety limits Proposed Question: SRO 81 Given the following conditions:

• The Reactor Vessel head has just been installed and is being tensioned following a Unit 2 Refueling outage.

• RCS temperature is currently 110ºF.

• Maintenance reports that Component Cooling Water (CCW) Pump 2-02 has a failed motor bearing.

Under these conditions, which ONE (1) of the following is the HIGHEST Reactor Coolant System temperature that Unit 2 can be increased to WITHOUT violating Technical Specifications?

A. 135ºF B. 195ºF C. 315ºF D. 345ºF Page 22 of 119 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 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 345ºF 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 3.3.7 Attached w/ Revision # See Technical Specification LCO 3.0.4 Comments / Reference Proposed references to be provided during examination: None Learning Objective: Given a Technical Specification or a Technical Specification situation, DIAGNOSE the situation and APPLY the LCO and SR Applicability of Section 3.0 to DETERMINE all corrective actions. (LO21.RLS.SL1.OB12)

Question Source: Bank # SYS.CC1.OB16-2 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 23 of 119 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From Technical Specification LCO 3.3.7 Amendment # 64 Page 24 of 119 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From Technical Specification LCO 3.0.4 Amendment # 109 Page 25 of 119 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 2 K/A # W/E01 & E02 G 2.4.18 Importance Rating 4.0 Rediagnosis and SI Termination: Emergency Procedures / Plan: Knowledge of the specific bases for EOPs Proposed Question: SRO 82 Given the following condition:

• Unit 1 has just entered MODE 4 when Pressurizer level and pressure start to lower uncontrollably.

Which of the following describes the applicability of EOP-0.0A, Reactor Trip or Safety Injection and EOS-0.0A, Rediagnosis while in MODE 4?

EOP-0.0A, Reactor Trip or Safety Injection...

A. is only applicable in MODES 1, 2, and 3.

EOS-0.0A, Rediagnosis is used in MODE 4 to determine the required procedure to enter.

B. is used in conjunction with EOS-0.0A, Rediagnosis in MODE 4 to ensure that the correct evaluation of conditions is being performed.

C. is performed in its entirety prior to transitioning to another procedure.

EOS-0.0A, Rediagnosis can be referenced throughout use of EOP-0.0A to ensure actions are correct.

D. must be used on a step-by-step basis to determine if the required action is still applicable for the plant conditions.

EOS-0.0A, Rediagnosis can be entered once the actions of EOP-0.0A are complete.

Proposed Answer: D Page 26 of 119 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible because EOP-0.0A is used in MODES 1, 2, and 3, however, it can also be used in MODE 4 on a step-by-step basis. EOS-0.0A is only used once the actions of EOP-0.0A are complete.

B. Incorrect. Plausible because both the EOP-0.0A and EOS-0.0A can be used while in MODE 4, however, EOP-0.0A is always completed before EOS-0.0A is referenced.

C. Incorrect. Plausible because EOP-0.0A must be performed in its entirety before EOS-0.0A is referenced, however, each step is evaluated on a case-by-case basis to ensure its applicability.

D. Correct. Per the rules of usage contained in ODA-407, Attachment 8.A, EOP-0.0A is evaluated on a step-by-step basis when in MODE 4 to determine if the required action is still applicable for the plant condition. At that point EOS-0.0A can be referenced if necessary with Safety Injection actuated or required.

Page 27 of 119 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Technical Reference(s) ODA-407, Attachment 8.A, Steps 6 and 7 Attached w/ Revision # See Comments / Reference Proposed references to be provided during examination: None Learning Objective: DESCRIBE how the use of the ERGs is affected by conditions that do not exactly match the Applicability conditions designated on the cover sheet.

(LO21.ERG.XD2.OB11)

STATE conditions under which EOS-0.0, Rediagnosis, is applicable, and DESCRIBE the conditions under which the operator would enter this procedure. (LO21.ERG.XD2.OB27)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

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

Reference:

From ODA-407, Attachment 8.A, Steps 6 and 7 Revision # 12 Page 28 of 119 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Page 29 of 119 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 2 K/A # W/E15 EA2.1 Importance Rating 3.2 Containment Flooding: Ability to determine and interpret the following as they apply to the Containment Flooding: Facility conditions and selection of appropriate procedures during abnormal and emergency operations Proposed Question: SRO 83 Given the following conditions:

• A Large Break Loss of Coolant Accident has occurred on Unit 2.

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

• The initial scan of the Critical Safety Function Status Trees indicates the following:

• Pressurizer level is 0%.

• Containment water level indicates 817 feet and slowly rising.

• Containment Spray has automatically actuated and was verified in EOP-0.0B, Reactor Trip or Safety Injection.

• Containment pressure is 14 psig and slowly lowering.

• Containment Radiation Monitors are in ALARM at 25 REM/hr.

• Reactor Vessel Level Indication System has no lights LIT.

• Core Exit Thermocouples are 292ºF.

Which ONE (1) of the following procedures must be entered to address the above conditions?

A. FRZ-0.2B, Response to Containment Flooding.

B. FRZ-0.1B, Response to High Containment Pressure.

C. FRI-0.2B, Response to Low Pressurizer Level.

D. FRZ-0.3B, Response to High Containment Radiation Level.

Proposed Answer: A Page 30 of 119 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Explanation:

A. Correct. Given the conditions listed, entry into FRZ-0.2B is correct due to a Containment Flooding condition existing.

B. Incorrect. Plausible because Containment pressure is 16 psig and slowly rising, however, it does not meet the Critical Safety Function Status Tree (CSFST) requirement for entry into FRZ-0.1B.

C. Incorrect. Plausible because Pressurizer level is low and the requirements to enter FRI-0.2B are met, however, it is a Yellow Path condition in the CSFST hierarchy.

D. Incorrect. Plausible because Containment radiation level is high and entry conditions for FRZ-0.3B are met, however, it is a Yellow Path condition in the CSFST hierarchy.

Technical Reference(s) FRZ-0.2B, Flow Chart Attached w/ Revision # See FRI-0.2B, Flow Chart Comments / Reference Proposed references to be provided during examination: None Learning Objective: Given the Containment CSF Status Tree and a description of applicable plant conditions, DETERMINE if entry into FRZ-0.2A/B is indicated, and STATE the severity of the challenge, if any. (LO21.ERG.FZ2.OB03)

Question Source: Bank # FRZ.XH5.OB408-2 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 31 of 119 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From FRZ-0.2B, Flow Chart Revision # 8 Page 32 of 119 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From FRI-0.2B, Flow Chart Revision # 8 Page 33 of 119 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 2 K/A # W/E14 G 2.2.38 Importance Rating 4.5 High Containment Pressure: Equipment Control: Knowledge of conditions and limitations in the facility license Proposed Question: SRO 84 Given the following conditions:

• Unit 1 just achieved 100% power the previous shift.

• Annunciator 1-ALB-3A, Window 4.6 - CNTMT NR PRESS HI/LO is in alarm.

• 1-PI-5470A and 1-PI-5470B, Containment Narrow Range Pressure Channels read 1.35 psig.

• Containment Intermediate Range pressure reads 1.4 psig.

• Containment average temperature is 98ºF and stable.

• Three Containment Fan Coolers are running.

• Dew points show no significant trend.

Which ONE (1) of the following describes the probable cause of the Containment pressure and temperature indications and the action required?

A. Containment Pressure is above the limit for normal operation due to normal Containment heatup and air usage.

Direct entry into SOP-801A, Containment Ventilation System and start all available Containment Fan Coolers to reduce pressure to less than 1.3 psig within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

B. Containment Pressure is below the Safety Analysis limit for normal operation due to excessive cooling.

Direct entry into SOP-801A, Containment Ventilation System and secure one Containment Fan Cooler within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

C. Containment Pressure is above the Safety Analysis limit for normal operation due to normal Containment heatup and air usage.

Direct entry into SOP-801A, Containment Ventilation System and reduce Containment pressure to less than 1.3 psig using the Containment Pressure Relief System within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

D. Containment Pressure is below the limit for normal operation due to condensation of fluids inside of Containment.

Direct entry into SOP-801A, Containment Ventilation System and align Containment Fan Coolers to reduce condensation within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

Page 34 of 119 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Proposed Answer: C Explanation:

A. Incorrect. Plausible because pressure is higher than the normal limits and it could be thought that the fan coolers would reduce the pressure, however, the Containment Fan Coolers would slow the rise perhaps but temperature is not abnormally high (>110ºF) and the air usage in Containment must also be accounted for. Additionally, the time frame is 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> not 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

B. Incorrect. Plausible because the time frame to restore the condition is correct and it could be thought that the alarm was on low pressure and that with three Containment Fan Coolers operating the cooling was excessive, however, the pressure is above the high end limit and restoration is by usage of the Containment Pressure Relief System.

C. Correct. The pressure is higher than Safety Analysis assumptions and High Energy leakage into Containment is not indicated based on Containment temperature and dew points. The Alarm Response directs reducing pressure per SOP-801A, Containment Ventilation System if temperature is less than 110ºF and dew points do not show an increasing trend, which specifies the Containment Pressure Relief System for reducing pressure.

D. Incorrect. Plausible because it could be thought that the alarm was on low pressure and that Containment Fan Cooler operation could reduce condensation and reduce pressure, however, the pressure is above the high end limit and the Containment Pressure Relief System must be used to reduce pressure. Additionally, the time frame is 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> not 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

Technical Reference(s) ALM-0031A, 1-ALB-3A, Window 4.6 Attached w/ Revision # See SOP-801A, Step 5.6.5 Comments / Reference SOP-801A, Step 4.1, Limitations Technical Specification LCO 3.6.4 Proposed references to be provided during examination: None Learning Objective: DESCRIBE the basis for the precautions and limitations, and given major procedure steps relative to the Containment Ventilation system, PLACE them in the proper sequence for:

• SOP-801, Containment Ventilation System (OP51.SYS.CL1.OB13)

Given a Technical Specification or a Technical Specification situation, DIAGNOSE the situation and APPLY the LCO and SR Applicability of Section 3.0 to DETERMINE all corrective actions. (LO21.RLS.SL1.OB12)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 2, 5 Page 35 of 119 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ALM-0031A, 1-ALB-3A, Window 4.6 Revision # 7 Page 36 of 119 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SOP-801A, Step 5.6.5 Revision # 13 Page 37 of 119 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SOP-801A, Step 4.1, Limitations Revision # 13 Comments /

Reference:

From Technical Specification LCO 3.6.4 Amendment # 64 Page 38 of 119 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 2 K/A # 033 G 2.2.22 Importance Rating 4.7 Loss of Intermediate Range Nuclear Instruments: Equipment Control: Knowledge of limiting conditions for operations and safety limits Proposed Question: SRO 85 Given the following conditions:

• A Reactor Startup is in progress.

• Both Source Range Nuclear Instruments are deenergized.

• Intermediate Range Nuclear Instrument Channel NI-35 reads 1.7E-9 amps.

• Intermediate Range Nuclear Instrument Channel NI-36 reads 2.1E-9 amps.

Which ONE (1) of the following describes the Technical Specification REQUIRED ACTION if Intermediate Range Nuclear Instrument (IRNI) Channel NI-36 suddenly fails to zero (0) power?

A. Suspend all positive reactivity additions until the Intermediate Range Nuclear Instrument channel is repaired.

B. Place IRNI NI-36 Level Trip in BYPASS and stabilize power slightly less than P-10, Nuclear at Power Permissive.

C. Raise THERMAL POWER to greater than P-10, Nuclear At Power Permissive within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

D. Verify Source Range Nuclear Instrument trips are blocked.

Proposed Answer: C Page 39 of 119 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible because this action would be required if both Intermediate Range Nuclear Instrument Channels were INOPERABLE per Technical Specification LCO 3.3.1, CONDITION G, however, with only one Channel INOPERABLE power is either lowered below P-6 or raised above P-10 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 the power increase could continue with one Intermediate Range Nuclear Instrument Channel INOPERABLE, however, power must be raised above P-10 within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> per Technical Specification LCO 3.3.1, CONDITION F.

C. Correct. As required per Technical Specification LCO 3.3.1, CONDITION F.

D. Incorrect. Plausible because verification that the Source Range Nuclear Instrument trips were blocked would be appropriate and the Reactor Startup could continue, however, a Technical Specification REQUIRED ACTION must be performed.

Technical Reference(s) Technical Specification LCO 3.3.1.F Attached w/ Revision # See From ABN-702, Step 2.3.2 Comments / Reference Proposed references to be provided during examination: None Learning Objective: ANALYZE the indications and DESCRIBE the mitigation strategy and major steps taken relative to the Excore Instrumentation system, both initial and subsequent, for:

• ABN-702, Intermediate Range Instrumentation Malfunction (OP51.SYS.EC1.OB25)

Given a Technical Specification or a Technical Specification situation, DIAGNOSE the situation and APPLY the LCO and SR Applicability of Section 3.0 to DETERMINE all corrective actions. (LO21.RLS.SL1.OB12)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 2 Page 40 of 119 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From Technical Specification LCO 3.3.1.F Amendment # 121 Page 41 of 119 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ABN-702, Step 2.3.2 Revision # 9 Page 42 of 119 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 012 A2.03 Importance Rating 3.7 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 or erratic operation of detectors and function generators Proposed Question: SRO 86 Given the following conditions on Unit 1 at 100% Power:

• INC-7375A, Analog Channel Operational Test Neutron Flux Power Range Channels is being performed.

• The Unit Supervisor is notified that the N-41 Overpower Trip High Range as-found setting was above the allowable ACCEPTANCE CRITERIA.

Which ONE (1) of the following:

1.) Identifies the most limiting impact on Technical Specification compliance due to placing Channel 1 in TRIP for High Power Range Flux being INOPERABLE?

2.) What action is required to restore the N-41 Overpower Trip High Range to OPERABLE status if repairs cannot be completed until after expiration of the Technical Specification COMPLETION TIME limit?

A. 1.) Place the Channel in TRIP within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> 2.) Enter Technical Specification LCO 3.0.7 and perform required testing within allowable exceptions for LCO suspension during testing.

B. 1.) Place the Channel in TRIP within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

2.) Perform a Risk Assessment and restore the Channel OPERABILITY during the next MODE 3 entry.

C. 1.) Place the Channel in TRIP within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

2.) Enter Technical Specification LCO 3.03 and perform required testing within one hour or commence a power reduction and enter MODE 3.

D. 1.) Place the Channel in TRIP within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

2.) Enter Technical Specification LCO 3.0.5 and perform required testing under Administrative Controls.

Proposed Answer: D Page 43 of 119 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible because 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is an allowable COMPLETION TIME in this TS and it could be thought that suspending LCO compliance per TS LCO 3.0.7 would be the appropriate action to invoke for Return to Service Testing, however, the allowable COMPLETION TIME is 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> and Return to Service Testing is allowed by TS LCO 3.0.5.

B. Incorrect. Plausible because 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is an allowable COMPLETION TIME in this TS and there is no time requirement to return this Channel to service, however, the required time to place the channel in TRIP is 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

C. Incorrect. Plausible because 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> is an allowable COMPLETION TIME in this TS and it could be thought that TS LCO 3.0.3 entry was required to perform Return to Service Testing, however, the COMPLETION TIME is 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> and Return to Service Testing is allowed by TS LCO 3.0.5.

D. Correct. The COMPLETION TIME is 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to place the Channel in TRIP. The Return to Service Testing is allowed past the COMPLETION TIME limitation per TS LCO 3.0.5.

Technical Reference(s) Technical Specification Table 3.3.1-1 Attached w/ Revision # See Technical Specification LCO 3.3.1.D Comments / Reference Technical Specification LCO 3.0.3 Technical Specification LCO 3.0.5 Technical Specification LCO 3.0.7 Proposed references to be provided during examination: None Learning Objective: Given a Technical Specification or a Technical Specification situation, DIAGNOSE the situation and APPLY the LCO and SR Applicability of Section 3.0 to DETERMINE all corrective actions. (LO21.RLS.SL1.OB12)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 2 Page 44 of 119 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From Technical Specification Table 3.3.1-1, Item #2a Amendment # 145 Page 45 of 119 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From Technical Specification LCO 3.3.1.D Amendment # 121 Page 46 of 119 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From Technical Specification LCO 3.0.3 Amendment # 64 Page 47 of 119 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From Technical Specification LCO 3.0.5 Amendment # 109 Comments /

Reference:

From Technical Specification LCO 3.0.7 Amendment # 64 Page 48 of 119 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 026 A2.08 Importance Rating 3.7 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: Safe securing of containment spray when it can be done Proposed Question: SRO 87 A Loss of Coolant Accident has occurred on Unit 1:

• Emergency Core Cooling is aligned for Cold Leg Injection.

• Reactor Coolant System subcooling is 20ºF.

• Reactor Coolant System Cold Leg temperatures are 320ºF.

• Refueling Water Storage Tank level is 35%.

• Containment Pressure is 15 psig.

• All Steam Generators are > 43% narrow range with 500 gpm Auxiliary Feedwater flow.

• All Safety Systems actuated as required.

• While evaluating Cold Leg Recirculation capability per EOP-1.0A, Loss of Reactor or Secondary Coolant it is discovered that equipment available will not support Cold Leg Recirculation.

Which ONE (1) of the following:

1.) Identifies the Critical Safety Function that will be the highest priority if Cold Leg Recirculation capability is not restored prior to RWST EMPTY actions?

2.) Describes actions that should be taken to mitigate the situation?

A. 1.) Core Cooling Safety Function will be an ORANGE Path.

2.) Enter FRC-0.2A, Response to Degraded Core Cooling and initiate maximum cooldown to Residual Heat Removal entry conditions.

B. 1.) Core Cooling Safety Function will be a YELLOW Path.

2.) Enter ECA-1.1A, Loss of Emergency Coolant Recirculation, and shutdown all running Containment Spray Pumps to conserve RWST inventory.

C. 1.) Heat Sink Safety Function will be a RED Path.

2.) Enter ECA-1.2A, LOCA Outside Containment and verify proper Residual Heat Removal and Safety Injection system alignments.

D. 1.) Containment Safety Function will be an ORANGE Path.

2.) Enter FRZ-0.1A, Response To High Containment Pressure and align Containment Spray Trains for recirculation from the Containment Sump.

Page 49 of 119 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Proposed Answer: B Explanation:

A. Incorrect. Plausible because it could be thought that this path would become ORANGE when the ECCS Pumps were secured at RWST EMPTY, however, the Core Exit Thermocouples have to reach 750ºF for this condition and the action to cooldown and align Residual Heat Removal cannot work unless inventory is controlled.

B. Correct. The Core Cooling Safety Function is already YELLOW based on subcooling less than 25ºF. The mitigation strategy is contained in ECA-1.1A, Loss of Emergency Coolant Recirculation and the actions are to conserve RWST inventory by reducing flows of all Safeguards Pumps. All four (4) Containment Spray Pumps would be secured based on current value of Containment pressure.

C. Incorrect. Plausible because current Steam Generator level is not high enough given Containment pressure and the RED Path could be pursued, however, there is adequate Auxiliary Feedwater flow in service. With a loss of Cold Leg Recirculation capability the cause could be inadequate Containment Sump level.

D. Incorrect. Plausible because it could be thought that the Containment Safety Function would be in greater jeopardy based on Containment being pressurized and the normal pressure to secure Containment Spray is 3 psig, however, the Containment Path is GREEN below 18 psig.

Technical Reference(s) ECA-1.1A, Purpose & Step 10 Attached w/ Revision # See FRH-0.1A, Status Tree Comments / Reference FRC-0.3A, Status Tree FRZ-0.1A, Status Tree Proposed references to be provided during examination: None Learning Objective: Given a procedural Step, Note or Caution from ECA-1.1, DISCUSS the reason or basis for the Step, Note or Caution. (LO21.ERG.C11.104)

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 50 of 119 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ECA-1.1A, Purpose Revision # 8 Page 51 of 119 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ECA-1.1A, Step 10 Revision # 8 Page 52 of 119 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From FRH-0.1A, Status Tree Revision # 8 Page 53 of 119 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From FRC-0.3A, Status Tree Revision # 8 Page 54 of 119 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From FRZ-0.1A, Status Tree Revision # 8 Page 55 of 119 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 059 G 2.4.9 Importance Rating 4.2 Main Feedwater System: Emergency Procedures/Plan: Knowledge of low power/shutdown implications in accident (e.g., loss of coolant accident or loss of residual heat removal) mitigation strategies Proposed Question: SRO 88 Given the following conditions on Unit 1 post-trip from 100% power:

• Steps 1 through 4 in EOP-0.0A, Reactor Trip or Safety Injection have been performed and the crew is at Step 1 of EOS-0.1A, Reactor Trip Response.

• Reactor Coolant System subcooling is 24ºF.

• Pressurizer pressure is 2150 psig.

• Reactor Coolant System TAVE is 557ºF.

• Steam Generator narrow range levels are as follows:

• 1-01 at 35%, 1-02 at 39%, 1-03 at 40%, and 1-04 at 35%.

• Auxiliary Feedwater is NOT available and Feedwater Flow is at zero (0) gpm.

• Containment Pressure is 0.4 psig.

• Safeguards Bus 1EA1 is deenergized.

Which ONE (1) of the following describes the required actions to be taken?

A. A RED Path exists for the Heat Sink Critical Safety Function. Immediately go to FRH-0.1A, Response to Loss of Secondary Heat Sink and align Main Feedwater to feed the Steam Generators.

B. A RED Path exists for the Heat Sink Critical Safety Function. Immediately go to FRH-0.1A, Response to Loss of Secondary Heat Sink and align for Reactor Coolant System Feed and Bleed.

C. Remain in EOS-0.1A, Reactor Trip Response and align Main Feedwater per the Response Not Obtained Actions for Check Feedwater Status.

D. A YELLOW path exists for the Heat Sink Critical Safety Function. Go to FRH-0.5A, Response to Steam Generator Low Level and restore Feedwater Flow.

Proposed Answer: A Page 56 of 119 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Explanation:

A. Correct. The operator enters EOP-0.0A/B, Reactor Trip or Safety Injection, and transitions to EOS-0.1A, Reactor Trip Response since SI is not actuated or required. Safety Function monitoring is initiated with the transition to another Emergency Response Guideline (ERG) procedure (EOS-0.1A, Reactor Trip Response). Functional Recovery Guideline implementation is initiated since EOP-0.0A, Reactor Trip or Safety Injection automatic action verification is complete (EOP-0.0A, Reactor Trip or Safety Injection, Steps 1 through 4). During subsequent ERG performance, a RED status exists for the Heat Sink Critical Safety Function and FRH-0.1A, Response to Loss of Secondary Heat Sink is implemented.

B. Incorrect. Plausible because a RED path does exist, however, the Steam Generator or Pressurizer pressure criteria for Reactor Coolant System Feed and Bleed are not met.

C. Incorrect. Plausible because actions to restore Feedwater Flow using Main Feedwater are in this procedure and Step, however, the Attachment 8A, ERG Rules of Usage, of ODA-407, Guideline on Use of Procedures, requires that the operator transition to the FRG.

D. Incorrect. Plausible because it could be thought that only a YELLOW path existed on Heat Sink, however, the combination of no Feedwater Flow and Steam Generator levels less than 43%

constitute a RED path for Heat Sink and there are no actions to align Main Feedwater.

Technical Reference(s) ODA-407, Attachment 8A, Step 10 Attached w/ Revision # See ODA-407, Attachment 8A, Example 3 Comments / Reference EOS-0.1A, Attachment 1A FRH-0.1A, CSFST FRH-0.1A, Steps 1, 3, & 7 Proposed references to be provided during examination: None Learning Objective: DESCRIBE the Symptoms or Entry Conditions for FRH-0.1.

(LO21.ERG.FH1.OB03)

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 57 of 119 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ODA-407, Attachment 8A, Step 10 Revision # 12 Page 58 of 119 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ODA-407, Attachment 8A, Example 3 Revision # 12 Page 59 of 119 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From EOS-0.1A, Attachment 1A, Step 1 Revision # 8 Page 60 of 119 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From FRH-0.1A, CSFST Revision # 8 Page 61 of 119 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From FRH-0.1A, Step 1 Revision # 8 Page 62 of 119 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From FRH-0.1A, Step 3 Revision # 8 Page 63 of 119 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From FRH-0.1A, Step 7 Revision # 8 Page 64 of 119 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 062 A2.12 Importance Rating AC Electrical Distribution System: Ability to (a) predict the impacts of the following malfunctions or operations on the AC distribution system; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Restoration of power to a system with a fault on it Proposed Question: SRO 89 Given the following conditions:

• Unit 1 is operating at 100% power.

• A loss of Instrument Bus 1EC1 has occurred due to a fault on the supply breaker to Inverter IV1EC1.

• The Instrument Bus has been powered from the Alternate Source, 1EC3.

Which ONE (1) of the following:

1.) Identifies the most restrictive Technical Specification impact associated with the Electrical Distribution Systems?

2.) Describes the action to be taken to mitigate the situation?

A. 1.) The Diesel is INOPERABLE because the Diesel will be unable to power the Safeguards Buses on a Loss of Offsite Power.

2.) Restore Instrument Bus power to normal within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

B. 1.) The Diesel and associated Blackout Sequencer are INOPERABLE because of inability to restore the Safeguards Bus power after a Loss of Offsite Power.

2.) Restore Instrument Bus power to normal within 30 days.

C. 1.) The AC Electrical Bus subsystem is DEGRADED.

2.) Restore Instrument Bus power to normal within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

D. 1.) The Instrument Bus Inverter is INOPERABLE.

2.) Restore Instrument Bus power to normal within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Proposed Answer: D Page 65 of 119 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible because it could be thought that the Diesel was the most significant component affected and the action length is correct, however, the Instrument Inverter loss is a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> REQUIRED ACTION.

B. Incorrect. Plausible because it could be thought that the combination of the sequencer and the Diesel being INOPERABLE would invoke a more limiting REQUIRED ACTION, however, the REQUIRED ACTION for the sequencer alone is 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> which is already more limiting.

C. Incorrect. Plausible because a degraded Bus Electrical Subsystem is a more limiting condition, however, the Instrument Bus has its own Technical Specification that would not apply unless there was a loss of power to the bus.

D. Correct. The REQUIRED ACTION time for an Inverter INOPERABLE is 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Technical Reference(s) ABN-603, Step 3.3.2 Attached w/ Revision # See Technical Specification LCO 3.8.1.B & F Comments / Reference Technical Specification LCO 3.8.7.A Technical Specification LCO 3.8.9.A & B Proposed references to be provided during examination: None Learning Objective: ANALYZE the indications and DESCRIBE the mitigation strategy and major steps taken, both initial and subsequent, for:

• ABN-603, Loss of Protection or Instrument Bus (OP51.SYS.AC3.OB13)

Given a Technical Specification or a Technical Specification situation, DIAGNOSE the situation and APPLY the LCO and SR Applicability of Section 3.0 to DETERMINE all corrective actions. (LO21.RLS.SL1.OB12)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 2, 5 Page 66 of 119 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ABN-603, Step 3.3.2 Revision # 7 Page 67 of 119 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From Technical Specification LCO 3.8.7.A Amendment # 66 Page 68 of 119 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From Technical Specification LCO 3.8.1.F Amendment # 138 Page 69 of 119 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From Technical Specification LCO 3.8.9.A & B Amendment # 142 Page 70 of 119 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 007 G 2.4.47 Importance Rating 4.2 Pressurizer Relief/Quench Tank System: Emergency Procedures/Plan: Ability to diagnose and recognize trends in an accurate and timely manner utilizing the appropriate control room reference material Proposed Question: SRO 90 Given the following conditions:

• Unit 1 has just completed a Heatup to normal operating temperature and pressure.

• Annunciator 1-ALB-5B, Window 4.3 - PRT LVL HI/LO is LIT.

• Pressurizer Relief Tank level is 88% which is a 10% rise in the last 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

• Pressurizer Relief Tank pressure is 7 psig and slowly rising.

• Pressurizer Relief Tank temperature is 90ºF and stable.

• Pressurizer Power Operated Relief Valve (PORV) discharge temperature is 95ºF and slowly rising.

• Pressurizer Safety Valve Discharge temperatures are all 95ºF and slowly rising.

• Pressurizer level is 25% and stable.

• Volume Control Tank level is 50% and stable.

• Containment temperature is 93ºF and slowly rising.

Which ONE (1) of the following is the procedural guidance that should be implemented given the conditions listed?

A. Perform actions of 1-ALB-5B, Window 3.1 - PRZR PORV OUT TEMP HI and direct the closure of the Power Operated Relief Valve Block Valve associated with the leaking PORV.

B. Enter SOP-801A, Containment Ventilation System and place additional Containment Fan Coolers in service to reduce the level in the Pressurizer Relief Tank to 64% due to the Containment heatup.

C. Enter SOP-109A, Pressurizer Relief Tank and reduce level in the Pressurizer Relief Tank due to the Reactor Makeup Water System Isolation Valve leaking by its seat.

D. Perform OPT-303, Reactor Coolant System Water Inventory to quantify the leakage from the relief valve on the Reactor Coolant Pump #1 Seal Water Return Line.

Proposed Answer: C Page 71 of 119 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible because the Pressurizer Power Operated Relief Valve (PORV) outlet and temperature is rising, however, the rise in PRT temperature for the amount of level increase, the outlet temperature of the PORV and the Reactor Coolant System Inventory change doesnt support this as the source.

B. Incorrect. Plausible because Containment and the PRT are heating up, however, the volume change in the PRT is 10% and the volume change in the PRT for a temperature rise from 32ºF to 90ºF would only be 0.4%.

C. Correct. This is the only viable source based on temperature and inventory parameters. The alarm response directs lowering the level if PRT parameters do not show a step increase.

D. Incorrect. Plausible because this is a relatively cooler source of leakage, however, this source would be indicated by a corresponding drop in Reactor Coolant System Inventory and Volume Control Tank level also would have dropped.

Technical Reference(s) TDM-804A, Pages 15 & 23 Attached w/ Revision # See ALM-0052A, 1-ALB-5B, Window 4.3 Comments / Reference Steam Tables Proposed references to be provided during examination: Steam Tables, TDM-804A, Pages 15 & 23 Learning Objective: Given specific plant and/or monitoring equipment conditions, DESCRIBE the Senior Reactor Operators responsibilities in accordance with CPSES Administrative Guidelines. Discussion should include:

• Selection of procedures and mitigation strategies based on system conditions, system parameters, and/or alarms.

(OPD1.EO0.XG2.OB14)

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 72 of 119 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From TDM-804A, Page 15 Revision # 2 Page 73 of 119 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From TDM-804A, Page 23 Revision # 2 Page 74 of 119 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ALM-0052A, 1-ALB-5B, Window 4.3 Revision # 5 Page 75 of 119 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 2 K/A # 068 A2.04 Importance Rating 3.3 Liquid Radwaste System: Ability to (a) predict the impacts of the following malfunctions or operations on the Liquid Radwaste System; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Failure of automatic isolation Proposed Question: SRO 91 Given the following conditions:

• A Liquid Release is in progress from Plant Effluent Holdup and Monitor Tank (PET)

X-01 to Outfall 101 with X-02, Plant Effluent Holdup and Monitor Tank Pump.

• A High Radiation alarm is received on the PC-11 for X-RE-5253, Liquid Waste Processing Discharge Radiation Detector.

• X-RV-5253, Liquid Waste Discharge Isolation Valve fails to close automatically due to the high radiation signal or close manually from the Liquid Waste Panel.

Which ONE (1) of the following:

1.) Identifies the Regulatory impact of this condition?

2.) What action should be taken to mitigate the situation per ABN-903, Accidental Release Of Radioactive Liquid?

A. 1.) The continued release may have exceeded the Liquid Effluent Limits of the Offsite Dose Calculation Manual.

2.) Secure the release by stopping X-02, Plant Effluent Holdup and Monitor Tank Pump.

B. 1.) The contents of the Plant Effluent Holdup and Monitor Tank may have exceeded the allowable curie content per Technical Requirement 13.10.33, Liquid Holdup Tanks.

2.) Secure the release by stopping X-02, Plant Effluent Holdup and Monitor Tank Pump.

C. 1.) The continued release may have exceeded the Liquid Effluent Limits of the Offsite Dose Calculation Manual.

2.) Secure the release by locally closing XWP-0117, LWPS DISCH HDR VLV 5253 UPSTRM ISOL VLV.

D. 1.) The contents of the Plant Effluent Holdup and Monitor Tank may have exceeded the allowable curie content per Technical Requirement 13.10.33, Liquid Holdup Tanks.

2.) Secure the release by locally closing XWP-0117, LWPS DISCH HDR VLV 5253 UPSTRM ISOL VLV.

Page 76 of 119 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Proposed Answer: C Explanation:

A. Incorrect. Plausible because the ODCM limits are the Regulatory concern, however, stopping the pump may not stop the release and is not the procedurally specified action.

B. Incorrect. Plausible because it could be thought that the Regulatory issue was that the curie content of the tank may be exceeding the limits of the Technical Requirements Manual, however, the tank had been sampled and the curie content was already known to be within limits and the issue is the continued release of a radioactive effluent that was exceeding alarm setpoints that are based on the ODCM.

C. Correct. The ODCM limits are the Regulatory concern and the closure of the upstream manual isolation valve is called for in ABN-903, Accidental Release of Radioactive Liquid and provides a positive stoppage of flow.

D. Incorrect. Plausible because it could be thought that the Regulatory issue was that the curie content of the tank may be exceeding the limits of the Technical Requirements Manual and that stopping the pump may stop the release, however, the ODCM limits are the Regulatory concern and the closure of the upstream manual isolation valve is called for in ABN-903, Accidental Release Of Radioactive Liquid and provides a positive stoppage of flow.

Technical Reference(s) ABN-903, Step 2.3.2 Attached w/ Revision # See ODCM, Step 3.3.3.4 Comments / Reference STA-501, Attachment 8.B, Item #4 Proposed references to be provided during examination: None Learning Objective: ANALYZE the status of the Waste Gas and Liquid Waste Systems and DETERMINE the appropriate system response to various conditions.

(LO21.SYS.RWS.OB03)

APPLY administrative requirements contained in CPNPP procedures and Licensing Basis Documents to the operation of the Waste Gas and Liquid Waste Systems. (LO21.SYS.RWS.OB04)

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 3, 5 Page 77 of 119 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ABN-903, Step 2.3.2 Revision # 6 Page 78 of 119 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ODCM, Step 3.3.3.4 Revision # 30 Page 79 of 119 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From STA-501, Attachment 8.B, Item #4 Revision # 14 Page 80 of 119 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 2 K/A # 034 A4.02 Importance Rating 3.9 Fuel Handling Equipment: Ability to manually operate and/or monitor in the control room: Neutron levels Proposed Question: SRO 92 Given the following conditions during Refueling:

• A fuel assembly is being lowered into the Reactor core.

• Annunciator 1-ALB-6D, Window 2.1 - SR SHTDN FLUX HI has just alarmed.

• Source Range indications are as follows:

• 1-NI-31B, SR COUNT RATE CHAN I is indicating 320 cps.

• 1-NI-32B, SR COUNT RATE CHAN II is indicating 330 cps.

Which ONE (1) of the following actions should be performed?

A. CORE ALTERATIONS may continue if Containment Ventilation Isolation has not actuated. Containment evacuation is NOT required.

B. Movement of the fuel assembly shall continue to place it in a safe location.

Containment evacuation IS required.

C. CORE ALTERATIONS may continue as long as Containment Integrity is met.

Containment evacuation is NOT required.

D. Movement of the fuel assembly must cease immediately.

Containment evacuation IS required.

Proposed Answer: B Explanation:

A. Incorrect. Plausible if thought that Containment Ventilation Isolation is used to confirm high flux at shutdown.

B. Correct. As outlined in RFO-102, the fuel assembly is placed in a safe location and the Containment is evacuated.

C. Incorrect. Plausible because if Containment Integrity is met CORE ALTERATIONS could continue, however, not for the conditions listed.

D. Incorrect. Plausible because a Containment evacuation is required, however, the fuel assembly must first be placed in a safe location.

Page 81 of 119 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Technical Reference(s) RFO-102, Precaution 3.9 Attached w/ Revision # See ALM-0064A, 1-ALB-6D, Window 2.1 Comments / Reference Proposed references to be provided during examination: None Learning Objective: DISCUSS the Precautions, Limitations and major procedural actions associated with RFO-102, Refueling Operation. (OPD1.OP9.X01.OB01)

Question Source: Bank # RFO.FH5.OB17-1 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 6, 7 Page 82 of 119 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From RFO-102, Precaution 3.9 Revision # 12 Page 83 of 119 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From ALM-0064A, 1-ALB-6D, Window 2.1 Revision # 6 Page 84 of 119 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 2 K/A # 086 A2.01 Importance Rating 3.1 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: Manual shutdown of the FPS Proposed Question: SRO 93 Given the following conditions:

• X-04, Electric Motor Driven Fire Pump automatically started.

• When X-HS-4091B, ELEC FIRE PMP STOP pushbutton is depressed, the Electric Motor Driven Fire Pump stops and then immediately restarts.

• FIRE WTR PMPHOUS FIRE PMP X-04 PRESS SW 4091, the AUTO start pressure switch has been determined to be failed.

• 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.

Which ONE (1) of the following:

1.) Identifies the OPERABILITY impact on the Fire Suppression Water System?

2.) Identifies the Compensatory Measures that should be implemented?

A. 1.) One of the 50% capacity Fire Pumps is INOPERABLE.

2.) Restore the Electric Fire Pump to OPERABLE status within 7 days or provide a backup source within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> per SOP-904, Fire Protection Main Water Supply and Fire Pumps System.

B. 1.) X-04, Electric Motor Driven Fire Pump remains OPERABLE because AUTO start is not review criteria for surveillance operability.

2.) Station a continuous fire watch to start X-04, Electric Motor Driven Fire Pump within 4 to 5 minutes when required per SOP-904, Fire Protection Main Water Supply and Fire Pumps System.

C. 1.) One of the 50% capacity Fire Pumps is INOPERABLE.

2.) Provide a backup water source within 30 days per SOP-904, Fire Protection Main Water Supply and Fire Pumps System.

D. 1.) The Fire Suppression Water System remains OPERABLE due to availability of the backup source from the Emergency Fill Fire Pump.

2.) Ensure the Emergency Fill Fire Pump is aligned for AUTO start per SOP-904, Fire Protection Main Water Supply and Fire Pumps System.

Page 85 of 119 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Proposed Answer: A Explanation:

A. Correct. The Fire Pump is out of service for the condition listed and the actions are per the Fire Impairment Plan.

B. Incorrect. Plausible because it could be thought that the Fire Pump remained OPERABLE because the AUTO start is not part of the surveillance review criteria, however, SOP-904 specifically states that the pump is out of service when shutdown and removed from auto service per the procedure.

C. Incorrect. Plausible because one 50% Fire Pump is out of service, however, 7 days are allowed prior to requiring an alternate pump.

D. Incorrect. Plausible because the Emergency Fill Fire Pump is an available pump, however, it does not satisfy the criteria to be one of the OPERABLE Fire Pumps and aligning for AUTO start would not satisfy the requirement as a backup source. The Emergency Refill Pump is not aligned for AUTO start when used as a backup.

Technical Reference(s) SOP-904, Steps 5.4.2, 4.2.G, & 5.6.4.1.B Attached w/ Revision # See STA-738, Attachment 8A, Step 1 Comments / Reference Proposed references to be provided during examination: None Learning Objective: Given that a specific Fire Protection System impairment has been identified, DESCRIBE the implementation of compensatory measures in accordance with STA-738 and the Fire Protection Report. (OPD1.ADM.FP1.OB13)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 1, 5 Page 86 of 119 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SOP-904, Step 5.4.2 Revision # 12 Comments /

Reference:

From SOP-904, Step 4.2.G Revision # 12 Page 87 of 119 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From SOP-904, Step 5.6.4.1.B Revision # 12 Page 88 of 119 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From STA-738, Attachment 8A, Step 1 Revision # 6 Page 89 of 119 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 3 Group # 1 K/A # G 2.1.32 Importance Rating 4.0 Conduct of Operations: Ability to explain and apply system limits and precautions Proposed Question: SRO 94 Which ONE (1) of the following addresses the Core Exit Thermocouple Bases for Technical Specification LCO 3.3.3, Post Accident Monitoring Instrumentation?

The Core Exit Thermocouples (CETs) have three (3) Limiting Condition for Operation requirements. Of these three requirements, one is that the minimum number of CETs that are OPERABLE cannot include the __________ rows of fuel assemblies because they are cooled by __________.

A. outer two; Steam Generator drainage due to refluxing B. outer two; radiation heat transfer to the vessel wall C. outer three; Steam Generator drainage due to refluxing D. outer three; radiation heat transfer to the vessel wall Proposed Answer: A Explanation:

A. Correct. As outlined in Technical Specification LCO 3.3.3, PAMI Bases.

B. Incorrect. Plausible because the outer two rows is correct, however, the reason is due to drainage during refluxing.

C. Incorrect. Plausible because Steam Generator drainage due to refluxing is correct, however, only the outer two rows are included.

D. Incorrect. Plausible if thought that radiation heat transfer to the vessel wall is the basis for Technical Specification LCO 3.3.3, PAMI. Additionally, only the outer two walls are affected during refluxing.

Technical Reference(s) Technical Specification LCO 3.3.3 Bases Attached w/ Revision # See Comments / Reference Proposed references to be provided during examination: None Learning Objective: LIST and DESCRIBE the following Technical Specifications (i.e. LCOs, action statements and conditional surveillance requirements of one hour and less, if applicable) for the Post Accident Instrumentation System:

• Accident Monitoring Instrumentation 3.3.3 (OP51.SYS.PA1.OB14)

Page 90 of 119 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Question Source: Bank # SYS.PA1.OB14-2 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 1, 2 Page 91 of 119 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From Technical Specification LCO 3.3.3 Bases Revision # 57 Page 92 of 119 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From Technical Specification LCO 3.3.3 Bases Revision # 57 Page 93 of 119 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 3 Group # 1 K/A # G 2.1.23 Importance Rating 4.4 Conduct of Operations: Ability to perform specific system and integrated plant procedures during all modes of plant operation Proposed Question: SRO 95 Given the following conditions:

• Day Shift has completed data collection for OPT-102A-1, Mode 1 and 2 Shiftly Surveillances.

• A change is issued and made effective at 1200 to the Offsite Dose Calculation Manual Surveillance Requirement for the Wind Direction channels.

• The change requires that all three (3) Wind Direction Channels be verified to be within 5º of one another.

• Due to the change to the ODCM, a PCN for OPT-102A-1 is issued and made effective at 1200.

• The current OPT-102A-1 states that the Wind Direction Channels be READING APPROXIMATELY THE SAME.

Which ONE (1) of the following describes how the Mid Shift data collection for Wind Direction is checked to be within the required band?

A. The Acceptance Criteria for Wind Direction on the OPT-102A-1 used by Day Shift is one-lined and the new ACCEPTANCE CRITERIA noted.

B. A new copy of OPT-102A-1 containing the updated Wind Direction ACCEPTANCE CRITERIA should be attached to the form used by Day Shift for Mid Shift readings.

C. A new copy of OPT-102A-1 containing the updated Wind Direction ACCEPTANCE CRITERIA should replace the form used by Day Shift after transferring all data from the replaced form.

D. The new ACCEPTANCE CRITERIA for Wind Direction is stated in the COMMENTS Section of the OPT-102A-1 used by Day Shift.

Proposed Answer: B Page 94 of 119 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible because one-lining is an acceptable method of making a procedure change except in this case the in-progress form should be replaced with issuance of the new form to ensure that the correct operational status of equipment is checked.

B. Correct. The initial data collection has been completed on the form, but the second data collection has not been performed, the copy of the form that is in-progress should be replaced with issuance of the new form to ensure that the correct operational status of the equipment is checked.

C. Incorrect. Plausible because a new copy of the form is issued, but the data is not transferred from the old form to the new form.

D. Incorrect. Plausible because it is expected that comments regarding readings recorded are made in the COMMENTS Section of the form, but in this case the in-progress form should be replaced with issuance of the new form to ensure that the correct operational status of equipment is checked.

Technical Reference(s) OPT-102A-1, Page 7 Attached w/ Revision # See STA-202, Page 22 Comments / Reference STA-202-7, Procedure Change Form Proposed references to be provided during examination: None Learning Objective: MAINTAIN Operations Department procedures, INITIATE procedure change requests and APPROVE procedures and procedure revisions.

(OPD1.ADM.XA1.OB10)

Question Source: Bank # ADM.XA1.OB10-2 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 3 Page 95 of 119 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From OPT-102A-1, Page 7 Revision # 36 Page 96 of 119 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From STA-202 Page 22 Revision # 34 PCN 1 Page 97 of 119 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From STA-202-7 Revision # 0 Page 98 of 119 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 3 Group # 2 K/A # G 2.2.38 Importance Rating 4.5 Equipment Control: Knowledge of conditions and limitations in the facility license Proposed Question: SRO 96 Given the following conditions:

• Unit 1 is at 100% power and the Plant Calorimetric is due in accordance with Technical Requirements LCO 13.3.34, Plant Calorimetric Measurement and Technical Specifications Surveillance Requirement 3.3.1.2, Comparing Calorimetric Heat Balance Results.

• I & C has reported that the Leading Edge Flow Meter is INOPERABLE because it failed the self-diagnostics test required prior to the Calorimetric Surveillance and will not be functional for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

• The last Calorimetric was performed 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> ago.

Which ONE (1) of the following identifies the required action for the conditions listed?

A. Perform calorimetric using Feedwater Venturis with RATED THERMAL POWER less than or equal to 3411 MWth then restore power to 100% RTP after calorimetric surveillance is completed.

B. Restore Leading Edge Flow Meter to service within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or reduce power in accordance with Technical Requirements Manual 13.3.34.

C. Perform calorimetric using Feedwater Venturis with RATED THERMAL POWER less than or equal to 3562 MWth then maintain power at 3562 MWth after calorimetric surveillance is completed.

D. Perform calorimetric using Feedwater Venturis with RATED THERMAL POWER at 3612 MWth and adjust Power as required after the calorimetric is completed.

Proposed Answer: C Page 99 of 119 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible because the MWth value listed corresponds to the limit used prior to power uprate of both units. Additionally, with the Leading Edge Flow Meter out of service power must remain at 98.6%.

B. Incorrect. Plausible because power must be reduced per TRM 13.3.34 prior to performing OPT-309, but it is due in 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br /> not 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

C. Correct. This is the correct MWth value for 98.6% power. Unit 1 must remain at 3562 MWth until the Leading Edge Flow Meter is returned to service and a new calorimetric can be performed.

D. Incorrect. Plausible because the Feedwater Venturis must be used, however, the MWth value listed is for 100% power.

Technical Reference(s) OPT-309, Step 5.2.1 Attached w/ Revision # See Technical Specification LCO SR 3.3.1.2 Comments / Reference Technical Requirements Manual 13.3.34 Proposed references to be provided during examination: None Learning Objective: STATE the performance and design attributes of the following Main Feedwater System components, flowpath and features:

• Feedwater flow detectors, including Inputs, Outputs and Indications provided by the Leading Edge Flow Meters (OP51.SYS.MF1.OB02)

Given a Technical Specification or a Technical Specification situation, DIAGNOSE the situation and APPLY the LCO and SR Applicability of Section 3.0 to DETERMINE all corrective actions. (LO21.RLS.SL1.OB12)

Question Source: Bank # SYS.MF1.OB02-14 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 1, 2, 5 Page 100 of 119 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From OPT-309, Step 5.2.1 Revision # 14 Page 101 of 119 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From Technical Specification LCO SR 3.3.1.2 Amendment # 144 Page 102 of 119 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From Technical Requirements Manual 13.3.34 Amendment # 72 Page 103 of 119 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 3 Group # 2 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 97 Given the following conditions:

• Unit 1 is in MODE 1 and Unit 2 is MODE 5.

• Radiation Monitor X-RE-5896B, South Control Room Air Intake is operating normally.

• Radiation Monitor X-RE-5895A, North Control Room Air intake fails LOW.

Which ONE (1) of the following identifies the Technical Specification requirements placed on the Control Room Heating, Ventilation, and Air Conditioning (HVAC) System?

A. Immediately place both Control Room HVAC Trains in the Emergency Recirculation Mode.

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 14 days.

D. Restore the affected Control Room Air Conditioning System Train to OPERABLE status within 30 days.

Proposed Answer: B Explanation:

A. Incorrect. Plausible because this action would be required if both Trains of Control Room Emergency Filtration System (CREFS) Actuation instrumentation were INOPERABLE but only on one Train per Technical Specification LCO 3.3.7.B.

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.

Page 104 of 119 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Technical Reference(s) Technical Specification LCO 3.3.7.A & B Attached w/ Revision # See Technical Specification LCO 3.7.10.A Comments / Reference Technical Specification LCO 3.7.11 Proposed references to be provided during examination: None Learning Objective: LIST and DESCRIBE the following Technical Specifications (i.e., LCOs, action statements and conditional surveillance requirements of one hour and less, if applicable) for the Control Room Ventilation System.

• TS 3.3.7 Control Room Emergency Filtration System (CREFS)

Actuation Instrumentation

• TS 3.7.10, Control Room Emergency Filtration/Pressurization System

• TS 3.7.11, Control Room Air Conditioning System

• TRM 13.7.36, Air Temperature Monitoring (OP51.SYS.HV1.OB20)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 2 Page 105 of 119 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From Technical Specification LCO 3.3.7.A Amendment # 64 Page 106 of 119 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From Technical Specification LCO 3.3.7.B Amendment # 64 Page 107 of 119 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From Technical Specification LCO 3.7.10.A Amendment # 64 Page 108 of 119 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From Technical Specification LCO 3.7.11.A Amendment # 64 Page 109 of 119 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 3 Group # 3 K/A # G 2.3.6 Importance Rating 3.8 Radiation Control: Ability to approve release permit Proposed Question: SRO 98 Given the following conditions:

• Unit 1 is performing power ascension from 98% to 100% at 1% per hour.

• A Containment Vent Permit has been approved with the following data:

• Containment was sampled at 1200 on February 13.

• The release concentration, as read on Radiation Monitor 1-RE-5503, Containment Air PIG Gas detector is 5.7 E-6 µci/cc.

• Prior to commencing the release at 0000 on February 15 the following information is noted:

• Radiation Monitor 1-RE-5503 10 minute trend has risen from 5.9 E-6 µci/cc to 6.5 E-6 µci/cc and stabilized.

Which ONE (1) of the following actions should be performed concerning the Containment Vent Permit?

Release via the Containment Vent Permit...

A. may continue based on the elapsed time since the initial sample.

B. should be voided due to increase in 1-RE-5503 10 minute trend readings.

C. may NOT continue based on the elapsed time since the initial sample.

D. should be voided because a power ascension is in progress.

Proposed Answer: A Page 110 of 119 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Explanation:

A. Correct. Because less than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> has elapsed and radiation level has not increased per the standards set in STA-602, the Containment Vent may be performed.

B. Incorrect. Plausible because radiation readings increasing by a factor of two following a 10 minute trend prior to a release should be terminated, however, this is not the case. STA-603 provides guidance regarding Containment noble gas readings during power ascension.

C. Incorrect. Plausible if thought that too much time had elapsed between samples.

D. Incorrect. Plausible because power ascension will cause Containment noble gas radiation levels to rise, however, since the reading has not changed by greater than a factor of two the release may continue.

Technical Reference(s) STA-603, Step 6.4.6 Attached w/ Revision # See Comments / Reference Proposed references to be provided during examination: None Learning Objective: Given conditions that warrant a radioactive effluent release, EVALUATE and DETERMINE the proper methodology for review and authorization of the release permit. (OPD1.ADM.XA8.OB106)

Question Source: Bank # ADM.XA8.OB06-12 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 4, 5 Page 111 of 119 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From STA-603, Step 6.4.6 Revision # 20 NOTE: IF the Containment PIG is out of service, THEN the Plant Vent Stack Monitors and the Wide Range Gas Monitors will provide Alarm indication, but there will be no trip function available to automatically secure a containment vent or purge.

6.4.6 Release from containment (purges or vents) should be initiated within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> of completion of sampling. Void the permit and initiate a new permit if:

NOTE: Noble gas concentrations in the containment buildings are directly related to reactor power level and RCS leakage. During steady state power level expected noble gas concentrations remain stable increasing slowly over core life. Rad monitor changes by a factor of 2 during steady state power indicate an increase in leakage or other abnormal radiological conditions inside the containment building. Rad monitor changes greater than a factor of 2 are expected when planned and controlled reactor power level changes significantly (e.g., reactor start-up after extended shutdowns or reactor shutdown following extended periods of power operation). Since gaseous activity levels are expected to increase by a factor of greater than 2 during reactor power changes, performing containment vents prior to planned power changes reduces the task complexity.

48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> elapse without initiation of the release, or, During steady state reactor power: The containment PIG noble gas monitor, 1-RE-5503 (CAG-197) or 2-RE-5503 (CAG-297), depending on which unit is being permitted, changes by a factor of 2 or more when 10 minute trends are compared just prior to release initiation. This requirement is not necessary if the containment PIG is out of service or during planned or controlled reactor power level changes.

During planned reactor power changes: When the containment PIG noble gas monitor, 1-RE-5503 (CAG-197) or 2-RE-5503 (CAG-297), depending on which unit is being permitted, changes by a factor of 2 or more when 10 minute trends are compared just prior to release initiation and there are other indications of increased RCS leakage inside containment (e.g., the particulate and iodine channels have increased by a factor of 2 or greater, containment sump pump run time increase, etc.). This requirement is not necessary if the containment PIG is out of service.

NOTE: The containment purge begins as a batch release. The release is considered a batch due to the replacement of air in the containment buildings. Historical information indicates that the majority of radioactivity in the atmosphere of the containment building will be removed in the first 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> of the containment purge. Six hours is used for planning purposes. The actual time period for the batch portion of the purge may vary.

Page 112 of 119 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 3 Group # 4 K/A # G 2.4.41 Importance Rating 4.6 Emergency Procedures / Plan: Knowledge of emergency action level thresholds and classifications Proposed Question: SRO 99 Given the following conditions:

• A Credible Bomb Threat has been received for a device located within the Control Room envelope.

• A search of the Control Room envelope by Security Personnel has not revealed any such device.

Which ONE (1) of the following Emergency Classifications, if any, should be made?

A. NO ACTION THIS CATEGORY B. NOTIFICATION OF UNUSUAL EVENT C. ALERT D. SITE AREA EMERGENCY Proposed Answer: B Explanation:

A. Incorrect. Plausible because Credible Security Threat block 7.A is TRUE and a Control Room Evacuation is not required as no device was found so block 7.E is not entered thus no path to NO ACTION THIS CATEGORY.

B. Correct. Credible Security Threat since block 7.A is TRUE but no device was found, therefore, block 7.B is FALSE, therefore it is a NOTIFICATION OF UNUSUAL EVENT.

C. Incorrect. Plausible because if the Credible Security Threat block 7.A and the Ongoing Security Threat block 7.B were determined to be TRUE with a Significant Security Breach/Compromise block 7.C determined to be FALSE, this would lead to an ALERT.

D. Incorrect. Plausible because if the Credible Security Threat block 7.A, the Ongoing Security Threat block 7.B, and a Significant Security Breach/Compromise block 7.C were determined to be TRUE with a Plant Security Lost block 7.D determined to be FALSE, this would lead to a SITE AREA EMERGENCY.

Technical Reference(s) EPP-201, Attachment 1, Chart 7 Attached w/ Revision # See EPP-201, Attachment 2, Chart 7 Bases Comments / Reference Proposed references to be provided during examination: EPP-201, Attachment 1 Page 113 of 119 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Learning Objective: DESCRIBE the process for Emergency Action Classification and DISCUSS the use of the EPP-201 Emergency Action Classification Charts and Bases.

(EP21.AC1.AG1.OB09)

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 EPP-201, Attachment 1, Chart 7 Revision # 11 Page 114 of 119 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From EPP-201, Attachment 2, Chart 7 Bases Revision # 11 Page 115 of 119 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 3 Group # 4 K/A # G 2.4.38 Importance Rating 4.4 Emergency Procedures / Plan: Ability to take actions called for in the facility emergency plan, including supporting or acting as emergency coordinator if required Proposed Question: SRO 100 Given the following conditions on Unit 2:

• At 1100, the Shift Manager (acting as Emergency Coordinator) declared an ALERT based on RCS leakage of 59 gpm with no indication of failed fuel.

• At 1114, initial notifications to Offsite Agencies were completed.

• At 1200, the Reactor was tripped and Safety Injection actuated due to increased RCS leakage.

• At 1209, Total Safety Injection pump injection was 400 gpm with an indication of failed fuel.

• At 1223, the Shift Manager (acting as the Emergency Coordinator) declared an escalation of the event to a SITE AREA EMERGENCY.

• At 1249, notification of Emergency Classification escalation to Offsite Agencies was completed.

Which ONE (1) of the following statements is correct regarding the reclassification and escalation notification?

A. The reclassification was timely.

The notification was NOT timely.

B. The reclassification was NOT timely.

The notification was timely.

C. The reclassification was timely.

The notification was timely.

D. The reclassification was NOT timely.

The notification was NOT timely.

Proposed Answer: A Page 116 of 119 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Explanation:

A. Correct. The notification was NOT made within 15 minutes but reclassification was made within the 15 minute time period.

B. Incorrect. Plausible as the notification was NOT timely, however, the reclassification was timely.

C. Incorrect. Plausible as the notification was NOT timely, however, the reclassification was timely.

D. Incorrect. Plausible as the notification was NOT timely, however, the reclassification was timely.

Technical Reference(s) EPP-203, Step 4.1.2 Attached w/ Revision # See Comments / Reference Proposed references to be provided during examination: None Learning Objective: DESCRIBE the process for Emergency Action Classification and DISCUSS the use of the EPP-201 Emergency Action Classification Charts and Bases.

(EP21.AC1.AG1.OB09)

IDENTIFY time requirements for emergency notifications.

(EP21.ECN.OW1.OB03)

Question Source: Bank #

Modified Bank # AC1.AG1.OB202-4 (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 117 of 119 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From EPP-203, Step 4.1.2 Revision # 15 1.0 INSTRUCTIONS 1.1 General Information [C-05708]

1.1.1 Following declaration of any emergency classification, identify at least one Plant Page Party System line to transmit information during the emergency.

1.1.2 Notify Somervell County, Hood County, and DPS of any emergency at CPNPP using the dedicated ring-down telephone system.

1.1.2.1 Notification shall be made within 15 minutes if any of the following conditions occur:

• Initial emergency classification;

• Escalation of emergency classification;

• Initial protective action recommendation (PAR);

• Change in protective action recommendation (PAR); or

• Termination of the emergency.

1.1.2.2 As a minimum, notify the offsite agencies every hour unless otherwise directed by the individual agency.

1.1.3 Notify CPNPP personnel of any emergency at CPNPP by the Plant Page Party System, Call-out system, or pager activation.

1.1.3.1 Contact the CPNPP personnel on site by the Plant Page Party System at all times.

1.1.3.2 Contact the ERO members not on site using the Call-out system.

1.1.3.3 Contact key ERO members by pager activation at all times.

1.1.3.4 Squaw Creek Park receives notification when the pager activation occurs. [C-05740]

1.1.4 Notify an NRC resident inspector of any emergency at CPNPP through the site NRC office or by pager.

Page 118 of 119 Rev e

ES-401 CPNPP Mar 2010 NRC Written Exam Worksheet Form ES-401-5 Comments /

Reference:

From AC1.AG1.OB202-4 Revision # 12/12/08 Given the following conditions:

• At 0200, the Shift Manager (acting as Emergency Coordinator) declared an ALERT based on RCS leakage of 59 gpm with no indication of failed fuel.

• 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, RCS leakage indication was 900 gpm with no indication of failed fuel.

• 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 ONE (1) of the following statements is correct regarding the reclassification and escalation notification?

A. The reclassification was timely.

The notification was NOT timely.

B. The reclassification was NOT timely.

The notification was timely.

C. The reclassification was timely.

The notification was timely.

D. The reclassification was NOT timely.

The notification was NOT timely.

Page 119 of 119 Rev e