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2012-03-DRAFT-Written Examination
	ML121080613
Person / Time
Site:	Palo Verde
Issue date:	03/15/2012
From:	Apger G; Operations Branch IV
To:	; Arizona Public Service Co
	laura hurley
References
	ES-401, ES-401-5 50-528/12-003, 50-529/12-003, 50-530/12-003
	Download: ML121080613 (168)
	v • d • e

ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Reactor Operator NRC Exam 1.

This Exam Level RO Appears on: RO EXAM 2012 Tier 1 Group 1 K/A # 4.1 007 EA2.03 Importance Rating: 4.2 Which ONE of the following describes ALL the available locations that ALL (4) RTSG breaker positions can be verified after a Reactor Trip?

(1) PPS Status Panel (2) Supplemental Protection Logic Actuation (SPLA) Cabinets (3) B05 Phase Current Lights (4) Locally at the Breaker A. 1 and 4 Only B. 1, 2 and 4 Only C. 2, 3 and 4 Only D. 1, 2, 3 and 4 Answer: B Reference Id: Q43923 Difficulty: 2.50 Time to complete: 2 10CFR Category: CFR 55.41 (7) 55.41 (7) Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Cognitive Level: Memory Question Source: New Comment:

Proposed reference to be provided to applicant during examination: NONE Technical

Reference:

LOIT Lesson Plan K&A: Ability to determine or interpret the following as they apply to a reactor trip: Reactor trip breaker position.

Learning Objective: L80279 Explain the operation of the RTSG (Reactor Trip Switchgear) Breakers.

Justification:

A. Incorrect: Each SPLA Cabinet has indication of their respective RTSG Breaker, B. Correct: RTSG Breaker position can be verified at these 3 locations.

C. Incorrect: PPS Status Panels do provide indication and the Phase Current Lights on B05 only show the status of C and D legs not individual breakers.

D. Incorrect:Phase Current Lights on B05 only show the status of C and D legs not individual breakers.

REV 0

ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Reactor Operator NRC Exam 2.

This Exam Level: RO Appears on: RO EXAM 2005 RO EXAM 2012 Tier 1 Group 1 K/A #: 4.2 008 AK2.01 Importance Rating: 2.7 Given the following conditions:

x Unit 1 RCS pressure is at 2000 psia.

x A Pressurizer safety/relief valve is leaking to the RDT.

x The RDT is at 10 psig.

Which ONE of the following describes the temperature of the fluid downstream of the relief valve?

A. 170°F B. 190°F C. 240°F D. 280°F Answer: C Reference Id: 4083 Difficulty: 3.00 Time to complete: 4 10CFR Category: CFR 55.41 (14) 55.41 (14) Principles of heat transfer thermodynamics and fluid mechanics.

Cognitive Level: Comprehension / Anal Question Source: PV Bank Not Modified Comment:

Proposed reference to be provided to applicant during examination: Steam Tables Technical

Reference:

Steam Tables, 40EP-9EO03. (LOCA)

K&A: Knowledge of the interrelations between the Pressurizer Vapor Space Accident and the following:

Valves Learning Objective: L10452 Given PZR Safety Valve tailpipe temperatures and the steam tables analyze the data to determine the status of the PZR safety valve.

Justification:

Directions on how to use Mollier Diagram and Steam Tables to determine tailpipe temperature of a leaking PSV.

1. Find the enthalpy of the saturated vapor using Mollier diagram or Table 2.

2. Plot this on the Saturation Line.

3. Draw a horizontal (constant h) line to the pressure that corresponds to where the device is relieving to.

4. If this point lies below the saturation line, follow the pressure line up the saturation line to determine the temperature. If above, compare the point to the Constant Temperature lines.

REV 0

ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Reactor Operator NRC Exam Any choice is plausible if the examinee does not obtain the specific enthalpy for 2000 psia or is off on drawing the lines to the correct values.

A. Incorrect: 170 0F corresponds to a RDT pressure of 10 psig if you go down on the curve.

B. Incorrect: 190 0F corresponds to a RDT pressure of 10 psig if you don't move on the curve.

C. Correct: Steam Tables diagram for a RCS press of 2000 psia and a RDT pressure at 10 psig is 240 0 0F.

D. Incorrect: 280 0F corresponds to a RCS pressure of 1800 psia.

REV 0

ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Reactor Operator NRC Exam 3.

This Exam Level: RO Appears on: RO EXAM 2012 Tier 1 Group 1 K/A #: 4.1 009 EK3.28 Importance Rating: 4.5 Given the following conditions:

x Unit 1 has tripped from 100% power.

x Sub-Cooled Margin is 36°F and lowering slowly.

x Containment Pressure is 2.7 psig and rising slowly.

x Pressurizer level is 20% and lowering slowly.

x RCS Pressure is 1780 psia and lowering slowly.

x SG #1 level is 28% WR and rising slowly.

x SG #2 level is 30% WR and rising slowly.

x SPTAs are in progress.

x NO ESFAS Actuations have occurred.

Which ONE of the following describes the ESFAS Actuations the RO must manually initiate due to the setpoints being exceeded?

A. CIAS ONLY B. SIAS and CIAS ONLY C. SIAS, CIAS and MSIS D. SIAS, CIAS and AFAS-1 Answer: B Reference Id: Q43924 Difficulty: 3.00 Time to complete: 3 10CFR Category: CFR 55.41 (7) 55.41 (7) Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Cognitive Level: Comprehension / Anal Question Source: New Comment:

Proposed reference to be provided to applicant during examination: NONE Technical

Reference:

EOP Setpoint Document and LOIT Lesson Plan K&A: Knowledge of the reasons for the following responses as the apply to the small break LOCA:

Manual ESFAS initiation requirements Learning Objective: L76810 List the parameters and setpoints that will cause PPS actuation.

REV 0

ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Reactor Operator NRC Exam Justification:

A. Incorrect: SIAS and CIAS setpoint is > 3.0 psig in CTMT or < 1837 psia PZR Pressure. CIAS is correct but SIAS is also correct.

B. Correct: SIAS and CIAS setpoint is > 3.0 psig in CTMT or < 1837 psia PZR Pressure.

C. Incorrect: SIAS, CIAS and MSIS setpoint is > 3.0 psig in CTMT. SIAS and CIAS setpoint < 1837 psia PZR Pressure.

D. Incorrect: SIAS and CIAS setpoint is > 3.0 psig in CTMT or < 1837 psia PZR Pressure. AFAS setpoint is < 25.8% WR.

REV 0

ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Reactor Operator NRC Exam 4.

This Exam Level RO Appears on: RO EXAM 2010 RO EXAM 2012 Tier 1 Group 1 K/A # 4.1 011 EK2.02 Importance 2.6 Rating:

Given the following conditions:

x A LOCA event results in a Reactor trip.

x Containment Pressure is 3.5 psig and rising.

x The SPTAs are in progress.

x RCS Subcooling indicates 20 °F.

Which ONE of the following describes the guidance regarding the operation of the RCPs?

A. Trip Two RCPs now (in SPTAs).

B. Trip Four RCPs now (in SPTAs).

C. The CRS shall not direct tripping of RCPs until an EOP is entered.

o D. The running RCPs shall remain operating until saturation conditions exist (0 F subcooling).

Answer: B Reference Id: Q6331 Difficulty: 2.00 Time to complete: 2 10CFR Category: CFR 55.41 55.41 (10) Administrative, normal, abnormal, and (10) emergency operating procedures for the facility.

10CFR Category: CFR 55.41 55.41 (10) Administrative, normal, abnormal, and (10)CFR emergency operating procedures for the facility.55.41 55.41 (7) (7) Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Cognitive Level: Comprehension / Anal Question Source: PV Bank Not Modified REV 0

ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Reactor Operator NRC Exam Comment:

Proposed reference to be provided to applicant during examination: NONE TECHNICAL

REFERENCE:

40EP-9EO01 SPTAs KA STATEMENT: Knowledge of the interrelations between the pumps and the following: Large break LOCA: Pumps.

JUSTIFICATION:

0 A. Incorrect - All RCPs are to be secured with subcooling < 24 F. Candidate may confuse the trip 2 leave 2 strategy with RCS pressure remaining below the SIAS setpoint.

B. Correct - This is the SPTA contingency for loss of subcooling. RCPs should not be operated without adequate subcooling.

C. Incorrect - The expectation is that these pumps will be secured prior to exiting the SPTAs. Candidate may think that this is an early step of the LOCA EOP.

D. Incorrect - This does not meet the standards set by the EOP Technical Guideline. Candidate may 0 0 understand loss of subcooling as < 0 F subcooling, not the procedurally directed < 24 F.

REV 0

ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Reactor Operator NRC Exam 5.

This Exam Level: RO Appears on: RO EXAM 2012 Tier 1 Group 1 K/A #: 4.2 077 AA1.05 Importance Rating: 3.9 Given the following conditions:

x Unit 1 has been manually tripped due to RCS leakage.

x East and West switchyard voltage dropped to 516 kV following a Main Turbine trip.

x East and West Bus switchyard Low-Low voltage alarms are locked in.

x Pressurizer level is 28% and slowly lowering.

x T-cold is stable at 564°F.

x The "B" Essential Cooling Water train has been aligned to supply Nuclear Cooling Water Priority loads.

x Charging flow is 88 gpm.

x The CRS has directed a manual SIAS initiation on trend.

x Pressurizer pressure is 1950 psia and slowly lowering.

Which ONE of the following describes the plant response?

A. Water Reclamation Facility supply breakers will trip open.

B. Reactor Coolant pump cooling water flow will go to 0 gpm.

C. Two RCPs (one in each loop) must be stopped when SIAS is initiated.

D. Charging flow will drop to 0 gpm then recover to 44 gpm 40 seconds after SIAS initiation.

Answer: A Reference Id: Q43997 Difficulty: 3.00 Time to complete: 3 10CFR Category: CFR 55.41 (7) 55.41 (7) Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Cognitive Level: Comprehension / Anal Question Source: New Comment:

Proposed reference to be provided to applicant during examination: None Technical

Reference:

LOIT Lesson Plans K&A: Ability to operate and/or monitor the following as they apply to Generator Voltage and Electric Grid Disturbances: Engineered safety features Learning Objective: L73573 Explain the operation of Switchgear NAN-S05 and NAN-S06 under normal operating conditions.

REV 0

ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Reactor Operator NRC Exam Justification:

A. Correct: WRF supply breakers NAN-S05G and NAN-S06C will trip open on a degraded Switchyard voltage <524 kV and a concurrent SIAS.

B. Incorrect: Candidate may think that the SIAS will isolate the All EW to NC cross tie, the EW 'A' valves will close on SIAS, the cross tie is with the EW 'B supplying, also NC CTMT isolation valves will close on a CSAS.

C. Incorrect: Two RCPs are directed to be tripped when pressure is below 1837 psia and not recovering. Candidate may confuse this with Trip 2 RCPs on SIAS, not the associated pressure.

D. Incorrect: Charging flow will remain the same on the SIAS, a LOP to the busses will cause the CCPs to load shed and sequence on 40 seconds later.

REV 0

ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Reactor Operator NRC Exam 6.

This Exam Level: RO Appears on: RO EXAM 2007 RO EXAM 2012 Tier 1 Group 1 K/A #: 4.2 022 AK3.02 Importance Rating: 3.5 Given the following conditions:

Initial Conditions:

x Unit 1 is operating at 100% power.

x Charging has been secured due to a leak downstream of the Charging Pumps.

x 40AO-9ZZ04, RCP Emergencies, has been entered.

Subsequently:

x The Unit trips due to a LOCA.

x Pressurizer pressure is currently 1500 psia and stable.

x Containment pressure is 2.1 psig and slowly increasing.

x Pressurizer level is 20% and stable.

x RCS T-cold is 560°F.

x RCS T-hot is 563°F.

x RCP 1A seal 2 outlet temperature is 260°F.

x RCP 2A seal 2 outlet temperature is 252°F.

x Safety Injection flow is adequate.

x RCPs 1A/2A have been secured.

Which ONE of the following actions should be taken?

A. Trip the 1B/2B RCPs to prevent pump cavitation.

B. Initiate CIAS, containment pressure is greater than setpoint.

C. Isolate Seal bleedoff to the 1A/2A RCPs to prevent seal damage.

D. Override and energize the class pressurizer heaters to restore pressurizer pressure.

Answer: C Reference Id: Q10375 Difficulty: 3.00 Time to complete: 2 10CFR Category: CFR 55.41 55.41 (10) Administrative, normal, abnormal, and (10) emergency operating procedures for the facility.

Cognitive Level: Comprehension / Anal Question Source: PV Bank Not Modified Comment:

Proposed reference to be provided to applicant during examination: Steam tables and Appendix 2 pump curves Technical

Reference:

40AO-9ZZ04 (RCP emergences)

REV 0

ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Reactor Operator NRC Exam K&A: Knowledge of the reasons for the following responses as they apply to the Loss of Reactor Coolant Pump Makeup: Actions contained in SOPs and EOPs for RCPs, loss of makeup, loss of charging, and abnormal charging Learning Objective: Given RCP motor amps and Upper Thrust Bearing Temperature determine the appropriate action to take based on RCP motor amps and thrust bearing temperature in accordance with 40AO-9ZZ04.

Justification:

A. Incorrect: subcooled margin and NPSH requirements are met B. Incorrect: containment pressure is less than setpoint of 3.0 psig C. Correct: RCP in stby with no seal injection requires that the Bleed Off valve be closed prior to exceeding 250 degrees on Seal 2 outlet temperature D. Incorrect: PZR level is less than 25%, heater cutout REV 0

ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Reactor Operator NRC Exam 7.

This Exam Level: RO Appears on: RO EXAM 2007 RO EXAM 2012 Tier 1 Group 1 K/A #: 42 025 AA2.07 Importance Rating: 3.4 Given the following conditions:

x Unit 1 is in Mode 4 x LPSI pump "B" is providing SDC flow x RCS temperature 325°F x Auxiliary Spray valve "B" fails open NOW x LPSI pump "B" amps are oscillating x SIB-FI-307 (SD Cooling B HDR flow to Loops) is fluctuating x Window 2B06A, SDC TRAIN A/B FLOW LO is alarming Which ONE of the following events/conditions is taking place?

A. LPSI pump B is "cavitating".

B. LPSI pump B is in a "runout" condition.

C. CHB-HV-530 (RWT to Train B SI Pumps) has closed.

D. Inadvertant B train Recirculation Actuation Signal (RAS).

Answer: A Reference Id: Q10357 Difficulty: 2.00 Time to complete: 3 10CFR Category: CFR 55.41 (10) 55.41 (10) Administrative, normal, abnormal, and emergency operating procedures for the facility.

Cognitive Level: Comprehension / Anal Question Source: PV Bank Not Modified Comment:

Proposed reference to be provided to applicant during examination: NONE Technical

Reference:

40EP-9EO11 40AL-9RK2B K&A: Ability to determine and interpret the following as they apply to the Loss of Residual Heat Removal System: Pump cavitation Learning Objective: Given the LMFRP HR-2 is being performed, and SDC is in service describe how adequate SDC flow is determined and what actions may be taken if adequate flow cannot be maintained in accordance with 40EP-9EO11.

JUSTIFICATION:

REV 0

ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Reactor Operator NRC Exam A. Correct: these are classic cavitation indications with lowering PZR pressure and stable temperature B. Incorrect: run out would be high amps and high flow C. Incorrect: SDC suction is thru SI-HV-655 and LPSI suction valve SI-HV-692 is closed isolating SDC flow from RWT D. Incorrect: RAS would trip the LPSI pump REV 0

ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Reactor Operator NRC Exam 8.

This Exam Level: RO Appears on: RO EXAM 2012 Tier 1 Group 1 K/A #: 4.2 026 AK3.03 Importance Rating: 4.0 Given the following conditions:

x Unit 1 has tripped from 100% power.

x A Loss of Turbine Cooling Water has occurred.

Which ONE of the following actions are directed by 40AO-9ZZ03 (Loss of Cooling Water) Appendix B (Minimizing Cooling Load on TC)?

A. Place SBCS system to OFF.

B. Place the Main turbine on the turning gear.

C. Direct SG Blowdown to the Main Condenser.

D. Place the FWPTs Turning Gear Handswitches in PTL.

Answer: D Reference Id: Q44000 Difficulty: 3.00 Time to complete: 3 10CFR Category: CFR 55.41 (10) 55.41 (10) Administrative, normal, abnormal, and emergency operating procedures for the facility.

Cognitive Level: Memory Question Source: New Comment:

Proposed reference to be provided to applicant during examination: NONE Technical

Reference:

40AO-9ZZ03 (Loss of Cooling Water)

K&A: Knowledge of the reasons for the following responses as they apply to the Loss of Component Cooling Water: Guidance actions contained in EOP for Loss of CCW Learning Objective: L10102 Given a sustained loss of the Plant or Turbine Cooling Water system(s) describe the required actions for a sustained loss of the Plant or Turbine Cooling Water System(s) in accordance with 40AO-9ZZ03.

Justification:

A. Incorrect: Step 13 of Appendix B directs transferring heat removal to SBCS Valves 1007 and 1008 or ADVs and then selecting OFF on SBCS Valves 1001 thru 1006. Taking SBCS to OFF will prevent any SBCS valves from opening.

B. Incorrect: Step 16 of Appendix B states Place the Main Turbine turning gear in PULL TO LOCK, placing the MT on the turning gear is the normal evolution post trip.

C. Incorrect. Step 1 of Appendix B states Securing SG Blowdown. Directing Blowdown to the condenser will not remove the heat load.

D. Correct: Step 4 of Appendix B states placing the FWPT turning gear to PULL TO LOCK (PTL), this removes the heat load of the lube oil system while on the turning gear.

REV 0

ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Reactor Operator NRC Exam 9.

This Exam RO Level Appears on: RO EXAM 2012 Tier 1 Group 1 K/A # 2.4.45 Importance 4.1 Rating:

Given the following conditions:

x RCN-PIC-100 (PZR Press Master Controller), is in AUTO.

x RCN-HS-100 (PZR Press Control Channel X/Y selector), is selected to channel X .

x Pressure transmitter RCN-PT-100X fails low.

The following annunciators alarm on B04:

Which ONE of the following describes the appropriate response by the RO?

The RO will FIRST address the PZR ...

REV 0

ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Reactor Operator NRC Exam A. TRBL Alarm and Stop PZR Heaters.

B. PRESS HI-LO Alarm and Stop PZR Heaters.

C. TRBL Alarm and select 100Y on RCN-HS-100 (Pressurizer Pressure Control Selector).

D. PRESS HI-LO Alarm and select 100Y on RCN-HS-100 (Pressurizer Pressure Control Selector)

Answer: D Reference Id: Q43926 Difficulty: 2.00 Time to complete: 2 10CFR Category: CFR 55.41 55.41 (10) Administrative, normal, abnormal, and (10) emergency operating procedures for the facility.

Cognitive Level: Comprehension / Anal Question Source: New Comment:

Proposed reference to be provided to applicant during examination: NONE Technical

Reference:

LOIT Lesson Plan K&A: Ability to prioritize and interpret the significance of each annunciator or alarm. PPCS Malfunction Learning Objective: Describe the conditions required to generate the following annunciators: PZR TRBL, PZR PRES HI-LO.

Justification:

A. Incorrect: PZR TRBL Alarm is Amber, so the priority shall be given to the Green PZR Press Hi-Lo alarm. Stop PZR heaters is a correct action ONLY if both pressure instruments Fail Low.

B. Incorrect: PZR Press Hi-Lo alarm is Green, this is the correct Alarm to address. Stop PZR heaters is a correct action ONLY if both pressure instruments Fail Low.

C. Incorrect: PZR TRBL Alarm is Amber color so the priority shall be given to the Green PZR Press Hi-Lo alarm.

Selecting the other instrument is the correct response per the ARP.

D. Correct: PZR Press Hi-Lo alarm is Green, this is the correct Alarm to address. Selecting the other instrument is the correct response per the ARP.

REV 0

ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Reactor Operator NRC Exam 10.

This Exam Level: RO Appears on: RO EXAM 2009 RO EXAN 2012 Tier 1 Group 1 K/A # 4.1 029 EK1.03 Importance 3.6 Rating:

Given the following conditions:

x Unit 1 is at 30% power while shutting down in preparations for a refueling outage.

x Reactor Coolant pump 1A has tripped.

x The reactor did not automatically trip.

x All attempts to trip the reactor from the Control Room have failed.

Assuming NO other operator actions, initiating an 80 gpm boration would add...

A. positive reactivity to the core and cause RCS temperature to increase.

B. positive reactivity to the core and cause RCS temperature to decrease.

C. negative reactivity to the core and cause RCS temperature to increase.

D. negative reactivity to the core and cause RCS temperature to decrease.

Answer: D Reference Id: Q22491 Difficulty: 3.00 Time to complete: 3 10CFR Category: CFR 55.41 (1) 55.41 (1) Fundamentals of reactor theory, including fission process, neutron multiplication, source effects, control rod effects, criticality indications, reactivity coefficients, and poison effects.

Cognitive Level: Comprehension / Anal Question Source: PV Bank Not Modified Comment:

Proposed reference to be provided to applicant during examination: NONE Technical

Reference:

40DP-9AP06 (SPTA tech guideline)

K&A: Knowledge of the operational implications of the following concepts as they apply to the ATWS:

Effects of boron on reactivity REV 0

ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Reactor Operator NRC Exam Learning Objective: Given plant conditions following a reactor trip analyze whether the Reactivity Control Safety Function is met and what contingency actions are required if it is not in accordance with 40EP-9EO01.

Justification: The examinee may confuse the purpose of boron and dilution as to which will add negative reactivity. Another consideration is that there is a time in core life (BOL, high boron concentration and low power) when a positive MTC could exist where the effects of temperature change don't follow the normal core dynamics.

A. Incorrect:

B. Incorrect:

C. Incorrect:

D. Correct:

REV 0

ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Reactor Operator NRC Exam 11.

This Exam Level: RO Appears on: RO EXAM 2012 Tier 1 Group 1 K/A #: 038 2.2.44 Importance Rating: 4.2 Given the following conditions:

x Unit 2 was tripped due to a Steam Generator Tube Rupture.

x RCS pressure is 895 psia.

x RCS subcooling is 55°F.

x Steam Generator #1 pressure is 890 psia.

x RU-4 in high alarm.

x Steam generator #1 is isolated.

x Steam generator #1 level is 78% NR and rising slowly.

x Steam generator #2 level is 50% NR and steady.

Which ONE of the following is the preferred method to control level in the isolated steam generator and minimize the spread of contamination?

A. Steam the #1 steam generator to atmosphere via the ADVs.

B. Bypass the MSIV and steam the #1 steam generator to the condenser.

C. Line-up high rate blowdown to the condenser from #1 steam generator.

D. Lower RCS pressure below #1 steam generator pressure and allow backflow to the RCS.

Answer: D Reference Id: Q44015 Difficulty: 3.00 Time to complete: 3 10CFR Category: CFR 55.41 (10) 55.41 (10) Administrative, normal, abnormal, and emergency operating procedures for the facility.

Cognitive Level: Comprehension / Anal Question Source: New Comment:

Proposed reference to be provided to applicant during examination: NONE.

Technical

Reference:

40EP-9EO04 (SGTR) 40DP-9AP09 (SGTR Tech Guide)

K&A: Ability to interpret control room indications to verify the status and operation of a system, and understand how operator actions and directives affect plant and system conditions. SGTR Learning Objective: L11218 Given that the SGTR EOP is being implemented describe the SGTR EOP mitigation strategy in accordance with 40EP-9EO04.

REV 0

ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Reactor Operator NRC Exam Justification:

A. Incorrect: This will lower SG pressure to further below RCS pressure which will increase SG level and spread more contamination.

B. Incorrect: This will lower SG pressure to further below RCS pressure which will increase SG level, steaming to the condenser would minimize the chance of release to the environment, but still spread the contamination to the secondary.

C. Incorrect: Blowdown will lower level, but spread contamination to the secondary.

D. Correct: This will lower RCS pressure and reduce level of the SG by moving water into the RCS.

Contamination will be limited by putting the contaminated water back in the RCS.

REV 0

ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Reactor Operator NRC Exam 12.

This Exam Level RO Appears on: RO EXAM 2009 RO EXAM 2012 K/A # 4.1 055 EK3.01 Importance Rating:

2.7 Given the following conditions:

x Unit 1 has tripped from 100% power due to a Loss of Offsite power.

x The "B" DG is out of service for scheduled maintenance.

x The "A" DG failed to come up to speed.

Under these conditions, the class (PK) batteries are designed to maintain rated voltage for ...

A. 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months to provide continuous DC during a Design Basis Event.

B. 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to provide continuous DC during a Design Basis Event.

C. 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months to provide sufficient power for the protection and control of transformers and switchgear.

D. 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to provide sufficient power for the protection and control of transformers and switchgear.

Answer: A Reference Id: Q22493 Difficulty: 2.00 Time to complete: 2 10CFR Category: CFR 55.41 (5) 55.41 (5) Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.

Cognitive Level: Memory Question Source: PV Bank Not Modified Comment:

Proposed reference to be provided to applicant during examination: NONE Technical

Reference:

FSAR, LOIT Lesson plans PRA SIGNIFICANT QUESTION REV 0

ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Reactor Operator NRC Exam K&A: Knowledge of the reasons for the following responses as the apply to the Station Blackout: Length of time for which battery capacity is designed Learning Objective: Discuss the purpose and conditions under which the 125 VDC Class IE Power System is designed to function.

Justification:

A. Correct: 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and concurrent DBE-LOCA concurrent with BO as found in FSAR B. Incorrect: 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months is the old rating for the non-lass NK batteries C. Incorrect: power for the protection and control of transformers is for the non-class NK batteries, examinee may choose this believing that the ESF transformers use class power D. Incorrect: 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months is the old rating for the non-lass NK batteries REV 0

ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Reactor Operator NRC Exam

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Unit Differences Question ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Reactor Operator NRC Exam 14. This Exam Level RO Appears on: RO EXAM 2012 Tier 1 Group 1 K/A # 2.4.50 Importance Rating: 4.2 Given the following conditions: x Unit 1 is operating at 100% power. x 120VAC IE PNL D27 Inverter C Trouble Alarm was received in the Control Room. x The area operator reports that DC power to 120VAC Class IE Inverter PNC-N13 has been lost. Which ONE of the following describes the restoration of power to PNC? PNC 120VAC power is restored by... A. an auto shift to the battery. B. a manual shift to the battery. C. an auto shift to the voltage regulator. D. a manual shift to the voltage regulator. Answer: D Reference Id: Q43931 Difficulty: 2.00 Time to complete: 2 10CFR Category: CFR 55.41 (7) 55.41 (7) Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features. Cognitive Level: Comprehension / Anal Question Source: New Comment: Proposed reference to be provided to applicant during examination: NONE Technical

Reference:

LOIT Lesson Plan, 41AL-9RK1A (Unit 1 B01A ARP) K&A: Ability to verify system alarm setpoints and operate controls identified in the alarm response manual. PRA SIGNIFICANT QUESTION UNIT DIFFERENCES QUESTION REV 0

ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Reactor Operator NRC Exam Learning Objective: Describe the conditions required to generate the following annunciators:

120VAC IE PNL D25 INV A
120VAC IE PNL D26 INV B
120VAC IE PNL D27 INV C
120VAC IE PNL D28 INV D Justification:

A. Incorrect: The battery is the normal supply to the inverter. Unit 1 is not equipped with a static transfer switch. B. Incorrect: The battery is the normal supply to the inverter. If the normal power supply was the voltage regulator, a manual transfer to the battery would be required. C. Incorrect: This would be correct in Unit 2 or 3 which is equipped with a Static Transfer switch that would automatically transfer to the voltage regulator. D. Correct: Unit 1 is NOT supplied with a Static Transfer switch as in Unit 2 and Unit 3. Therefore on a loss of Power to the Inverter the operator must manually transfer the power supply from the inverter to the voltage regulator. REV 0

ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Reactor Operator NRC Exam 15. This Exam Level RO Appears on: RO EXAM 2012 Tier 1 Group 1 K/A # 4.2 058 AK1.01 Importance Rating: 2.8 Given the following conditions: x Unit 3 is operating at 2% power. x AFN-P01 (Non Essential Motor Driven Aux Feed Pump) is feeding both SGs. x AFB-P01 (Essential Motor Driven Aux Feed Pump) is out of service for maintenance. Which ONE of the following describes the operation of AFN-P01 following a Loss of PKA-M41 Control Power? AFN-P01 Control power must be shifted to the 'A' Battery... A. output to restore remote operation of the breaker. B. charger output to restore remote operation of the breaker. C. output to restore remote operation of both the breaker and suction valves. D. charger output to restore remote operation of both the breaker and suction valves. Answer: B Reference Id: Q43933 Difficulty: 2.00 Time to complete: 2 10CFR Category: CFR 55.41 (7) 55.41 (7) Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features. Cognitive Level: Memory Question Source: New Comment: Proposed reference to be provided to applicant during examination: NONE Technical

Reference:

40AO-9ZZ13 (Loss of Class Instrument and Control Power) PRA SIGNIFICANT QUESTION REV 0

ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Reactor Operator NRC Exam K&A: Knowledge of the operational implications of the following concepts as they apply to Loss of DC Power: Battery charger equipment and instrumentation. Learning Objective: Given a loss of PN or PK describe the availability of Auxiliary Feedwater in accordance with 40AO-9ZZ13. Justification: A. Incorrect: The normal supply is directly off of the PKA-M41, this is the correct action per the AOP. B. Correct: Per step 8b. IF AFB-P01 is NOT available, AND Battery Charger A is available, THEN perform the following: 1) Direct an operator to place PBA-U01 CONTROL POWER TRANSFER SWITCH FOR AFN-P01 to the ALTERNATE FEED FROM PKA-H11 position. C. Incorrect: PKA-M41 is the normal power supply from the battery. Suction valves are powered from PHA-M35 D. Incorrect: Switching to the output of the charger is correct but the suction valves are powered from PHA-M35. REV 0

ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Reactor Operator NRC Exam 16. This Exam Level RO Appears on: RO EXAM 2012 Tier 1 Group 1 K/A # 4.4 E05 EK2.2 Importance Rating: 3.7 Given the following conditions: Initial Conditions: x Unit 2 has tripped from 100% power. x SG #1 is 1000 psia and lowering. x SG #1 is 40% WR and lowering. x SG #2 is 800 psia and lowering. x SG #2 is 10% WR and lowering. x PZR level is at 30% and slowly lowering. x Containment Pressure is 1 psig and rising. At the time that the ORP is entered the conditions are as follows: x Containment pressure peaked and is stable at 9.8 psig. x Containment temperature is 185°F. x PZR level is 18% and rising. x RVUH level is 67%. x RCS subcooling is 98°F. x SG #1 is at 34% WR (rising) and being fed from AFW at 500 gpm. x SG #2 is below the indicated level. x Both HPSI pumps are injecting into the RCS. Based on these conditions, you should obtain CRS concurrence and throttle HPSI... A. immediately. B. when PZR level reaches 33%. C. when RVUH is equal to 100%. D. when SG #1 Level are 45%-60% NR. Answer: A REV 0

ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Reactor Operator NRC Exam Reference Id: Q43934 Difficulty: 3.00 Time to complete: 2 10CFR Category: CFR 55.41 55.41 (10) Administrative, normal, abnormal, and (10) emergency operating procedures for the facility. 10CFR Category: CFR 55.41 55.41 (10) Administrative, normal, abnormal, and (10)CFR emergency operating procedures for the facility.55.41 55.41 (8) (8) Components, capacity, and functions of emergency systems. Cognitive Level: Comprehension / Anal Question Source: Modified PV Bank Comment: Proposed reference to be provided to applicant during examination: NONE Technical

Reference:

40EP-9EO05, Excess Steam Demand, 40EP-9EO10 Appendix 2 SI Throttle Criteria K&A: Knowledge of the interrelations between the (Excess Steam Demand) and the following: Facility's heat removal systems, including primary coolant, emergency coolant, the decay heat removal systems, and relations between the proper operation of these systems to the operation of the facility. Learning Objective: Given conditions of an ESD describe the mitigating strategy outlined in the ESD EOP in accordance with 40EP-9EO05. Justification: A. Correct - PZR level requirement is > 15% for Harsh CTMT conditions. B. Correct - PZR level requirement for throttling HPSI is > 15% level when in Harsh CTMT conditions. 33% is the normal PZR Level Band per SPTAs C. Incorrect - RVUH level must be greater than 16% to throttle HPSI, which it is. Candidate may not understand RVUH and Plenum relationship. D. Incorrect - The SG requirement is RESTORING to 45-60% NR level. Candidate may believe that SG levels must be in the band. REV 0

Given the following conditions: Initial Conditions: x Unit 2 has tripped from 100% power. x SG #1 is 1000 psia and lowering. x SG #1 is 40% WR and lowering. x SG #2 is 800 psia and lowering. x SG #2 is 10% WR and lowering. x PZR level is at 30% and slowly lowering. x Containment Pressure is 1 psig and rising. At the time that the ORP is entered the conditions are as follows: x Containment pressure peaked and is stable at 9.8 psig. x Containment temperature is 185°F. x PZR level is 12% and rising. x RVUH level is 67%. x RCS subcooling is 98°F. x SG #1 is at 34% WR (rising) and being fed from AFW at 500 gpm. x SG #2 is below the indicated level. x Both HPSI pumps are injecting into the RCS. Based on these conditions, you should obtain CRS concurrence and throttle HPSI... A. immediately. B. when PZR level reaches 15%. C. when RVUH is equal to 100%. D. when SG #1 Level are 45%-60% NR. Answer: B OPTRNG_EXAM Page: 1 of 1 22 September 2011

ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Reactor Operator NRC Exam 17. This Exam Level: RO Appears on: RO EXAM 2012 Tier 1 Group 1 K/A # 4.4 E06 EA1.2 Importance 3.4 Rating: Given the following conditions: x Unit 1 is tripped from 100% power. x Containment Pressure is 1.7 psig and rising. 0 x Containment Temperature is 120 F and rising. x Containment Humidity is rising. x Containment sump levels are rising. x PZR Pressure is 2250 psia and rising. x PZR Level is 58% and rising. 0 x Tcold is 568 F and rising. 0 x Subcooled Margin is 58 F and lowering. x SG 1 and 2 levels are 30% WR and lowering. Which ONE of the following describes the ongoing event? A. RCS Cold Leg LOCA. B. PZR Steam Space LOCA. C. Feedline Break (ESD) inside containment. D. Steam Line Break (ESD) inside containment. Answer: C Reference Id: Q43935 Difficulty: 3.00 Time to complete: 3 10CFR Category: CFR 55.41 (5) 55.41 (5) Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics. REV 0

ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Reactor Operator NRC Exam Cognitive Level: Comprehension / Anal Question Source: New Comment: Proposed reference to be provided to applicant during examination: NONE Technical

Reference:

40EP-9EO05, ESD K&A: Ability to operate and / or monitor the following as they apply to the (Loss of Feedwater) Operating behavior characteristics of the facility. Learning Objective: Given conditions of an ESD analyze whether or not entry into the ESD EOP is appropriate in accordance with 40EP-9EO05. Justification: A. Incorrect: CTMT parameters changing are indicative of a LOCA inside the CTMT, Subcooling lowering is indicative of a LOCA. Tc and PZR parameters would lower. B. Incorrect: CTMT parameter and PZR level rising support the PZR Steam Space LOCA as does lowering subcooling. PZR Pressure would be lowering. C. Correct: All of these parameters support the Feedline Break inside CTMT. D. Incorrect: CTMT parameters support the Steam Line Break inside CTMT. Subcooling would rise, PZR Pressure and Level would lower. ESD procedure will mitigate both the Feedline and Steam Line breaks. REV 0

ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Reactor Operator NRC Exam 18. This Exam Level: RO Appears on: RO EXAM 2012 Tier 1 Group 1 K/A #: 4.2 065 AA1.03 Importance Rating: 2.9 Given the following conditions: x Unit 1 has experienced a Loss of Instrument Air (IA) to the Containment. x The CRS is implementing 40AO-9ZZ06 (Loss of Instrument Air). Which ONE of the following valves handswitches must be taken to CLOSE prior to restoring IA to Containment per 40AO-9ZZ06? A. CHA-HV-507 (RCP Bleedoff Isolation to RDT) B. CHA-UV-516 (Letdown to Regen Hx Isolation) C. WCB-UV-61 (CHW Return HDR Inside CNTMT Isol VLV) D. NCB-UV-403 (NCW CNTMT Downstream Return Isol VLV) Answer: B Reference Id: Q43990 Difficulty: 3.00 Time to complete: 2 10CFR Category: CFR 55.41 (7) 55.41 (7) Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features. Cognitive Level: Memory Question Source: New Comment: Proposed reference to be provided to applicant during examination: NONE Technical

Reference:

40AO-9ZZ06 (Loss of Instrument Air) OPERATING EXPERIENCE QUESTION K&A: Ability to operate and / or monitor the following as they apply to the Loss of Instrument Air: Restoration of systems served by instrument air when pressure is regained Learning Objective: Determine the mitigating strategies of the Loss of Instrument air AOP. Justification: A. Incorrect: This is an IA operated valve inside the CTMT that fails open to allow Seal Bleed Off to the RDT, it is not to be closed. B. Correct: Per step 4 of section 3.0, this valve will fail closed but if the handswitch is not taken to close the valve will open upon restoration of IA and possibly lead to damage of the letdown IXs. C. Incorrect: This valve is a Motor Operated Valve that will not be affected by the loss of IA, it is the inside CTMT isolation valve for WC. D. Incorrect: This valve is a Motor Operated Valve that will not be affected by the loss of IA, it is the inside CTMT isolation valve for NC. REV 0

ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Reactor Operator NRC Exam 19. This Exam Level RO Appears on: RO EXAM 2008 RO EXAM 2012 K/A # 4.2 001 AA1.07 Importance Rating: 3.3 Given the following conditions: x Unit 3 is operating at 80%. x Group 5 CEAs at 120 inches withdrawn. x All others CEAs at UEL. x Selected CEA is # 14. x Selected CEA Group is # 5. x A malfunction causes CEA 15 to move 12 steps out before STANDBY is selected and motion stops. Based on this event the pulse counter selected Group position reads... A. 120 inches. B. 122.25 inches. C. 124.5 inches. D. 129 inches. Answer: B Reference Id: Q43936 Difficulty: 2.00 Time to complete: 4 10CFR Category: CFR 55.41 (7) 55.41 (7) Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features. Cognitive Level: Comprehension / Anal Question Source: PV Bank Not Modified Comment: Proposed reference to be provided to applicant during examination: NONE Technical

Reference:

LOIT lesson plan REV 0

ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Reactor Operator NRC Exam K&A: Ability to operate and / or monitor the following as they apply to the Continuous Rod Withdrawal: RPI Learning Objective: Describe the required actions addressing a continuous rod motion accident. Justification: 12 steps times 3/4 inch equals 129 inches withdrawn A. Incorrect: examinee may believe that that the pulse counter uses lowest CEA position (CPCs) B. Correct: group position is the average position C. Incorrect: examinee may believe that the pulse counter uses average of high/low D. Incorrect: examinee may believe that pulse counter uses highest CEA position (CPCs) REV 0

ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Reactor Operator NRC Exam 20. This Exam Level RO Appears on: RO EXAM 2012 Tier 1 Group 2 K/A # 4.2 005 AK3.06 Importance 3.9 Rating: The CRS has directed the RO to open the supply breakers for L03 and L10 for a minimum of 5 seconds. Which ONE of the following describes the reason for this action? The 5 seconds allows time for the... A. motor generator stop contacts to close. B. CEAs to drop to the bottom of the core. C. trip coils to actuate to open L03 and L10 breakers. D. effects of the motor generator flywheel to taper off interrupting power to the CEAs. Answer: D Reference Id: Q43938 Difficulty: 2.00 Time to complete: 2 10CFR Category: CFR 55.41 (6) 55.41 (6) Design, components, and functions of reactivity control mechanisms and instrumentation. Cognitive Level: Memory Question Source: New Comment: Proposed reference to be provided to applicant during examination: NONE Technical

Reference:

EOP OPERATIONS EXPECTATIONS K&A: Knowledge of the reasons for the following responses as they apply to the Inoperable / Stuck Control Rod: Actions contained in EOP for inoperable/stuck control rod. Learning Objective: Given plant conditions following a reactor trip analyze whether the Reactivity Control Safety Function is met and what contingency actions are required if it is not in accordance with 40EP-9EO01. REV 0

ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Reactor Operator NRC Exam Justification: A. Incorrect: MG stop contact does not get a signal to actuate, these actions remove power from the MG set input, therefore no output. B. Incorrect: CEAs do require to be inserted within 4 seconds per Tech Specs, but this is not the reason for the 5 second wait. C. Incorrect: Trip coils inside the breaker have no time delay associated with them, they open instantaneously. D. Correct: As the Load Center supplying power to the MG sets is de-energized, the MG set flywheels will maintain the MG set output as inertial energy is dissipated. REV 0

ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Reactor Operator NRC Exam 21. This Exam Level RO Appears on: RO EXAM 2012 Tier 1 Group 2 K/A # 4.2 028 AK1.01 Importance 2.8 Rating: Given the following conditions: x Unit 3 operating at 100% power. x RCN-LIC-110 (Pressurizer Level Master Controller) is in "REMOTE-AUTO". x RCN-HS-110 (Level Control Selector Channel X/Y) is selected to channel 'Y'. x RCN-HS-100-3 (Pressurizer Heater Control Selector Level Trip Channel) is selected to 'X'. x A leak develops on the reference leg of RCN-LT-110Y (Level Transmitter 110Y). This leak exceeds the capacity of the condensing chamber's ability to keep the reference leg full. Assuming NO operator action, which ONE of the following describes the plant response? A. Letdown will be lost. B. The standby charging pump will start. C. Presssurizer heaters will cut-out on low level. D. Actual letdown flow will lower and stabilize at approximately 30 gpm. Answer: A Reference Id: Q43992 Difficulty: 2.00 Time to complete: 2 10CFR Category: CFR 55.41 (7) 55.41 (7) Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features. Cognitive Level: Comprehension / Anal Question Source: New Comment: Proposed reference to be provided to applicant during examination: NONE Technical

Reference:

LOIT Lesson Plan K&A: AK1.01 PZR reference leak abnormalities. Knowledge of the operational implications of the following concepts as they apply to Pressurizer Level Control Malfunctions: PZR reference leak abnormalities Learning Objective: Describe the response of the Pressurizer Level Control System to a failure of a Pressurizer Level Transmitter. REV. 0

ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Reactor Operator NRC Exam Justification: A. Correct: The level control system will sense a high level. Letdown flow increases to maximum.

     "Normally running" charging pump stops. Letdown will isolate due to the automatic closure of CHB-UV-0515 upon receipt of a hi-hi regenerative heat exchanger outlet temperature.

B. Incorrect:The level control system will sense a high level causing the standby charging pump to stop. C. Incorrect: The heaters cut out at 27% indicated level. D. Incorrect: The level control system will sense a high level. Letdown flow increases to maximum. REV. 0

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ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Reactor Operator NRC Exam 23. This Exam Level: RO Appears on: RO EXAM 2012 Tier 1 Group 2 K/A # 4.2 068 AK2.02 Importance Rating: 3.7 Given the following conditions: x Unit 2 Control Room is experiencing a fire. x The CRS has directed an evacuation of the Control Room. x 40AO-9ZZ19 (Control Room Fire) has been entered. Which ONE of the following describes the appropriate actions per the AOP? A. Initiate a RPCB Loss of Feed Pump from B04. B. Initiate a boration from the Remote Shutdown Panel. C. Trip the Reactor by opening the RTSG breakers locally. D. Trip the Reactor by depressing the RTSG Pushbuttons on B05. Answer: D Reference Id: Q43941 Difficulty: 2.00 Time to complete: 2 10CFR Category: CFR 55.41 55.41 (10) Administrative, normal, abnormal, and (10) emergency operating procedures for the facility. Cognitive Level: Memory Question Source: New Comment: Proposed reference to be provided to applicant during examination: NONE Technical

Reference:

40AO-9ZZ19 (Control Room Fire) K&A: Knowledge of the interrelations between the Control Room Evacuation and the following: Reactor trip system. Learning Objective: State the operator actions that are required to be performed prior to evacuation in the event of a Control Room fire. Justification: REV 0

ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Reactor Operator NRC Exam A. Incorrect: Numerous AOPs use the RPCB Loss of Feed Pump as a means of a rapid downpower. B. Incorrect: This is the correct action if after the trip is initiated from the CR, and a CEA doesn't fully insert into the core. C. Incorrect: Tripping the Reactor is the correct direction, just not the location. Candidate may think that due to the CR Fire that all actions must be taken outside of the control room. D. Correct: Per the not prior to and including Step 2a of the AOP, Steps 2-5 are expected to be performed in the control room and 2a. states Trip the Reactor. REV 0

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ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Reactor Operator NRC Exam 25. This Exam Level: RO Appears on: RO EXAM 2008 RO EXAM 2012 Tier 1 Group 2 K/A #: 4.4 A16 AK3.3 Importance Rating: 3.3 Given the following conditions: x Unit 1 is operating at 100% power. x Pressurizer level is slowly lowering. x RCS temperature is stable. x The in-service letdown control valve CHN-110P is slowly closing. x The CRS implements the appropriate AOP. x All available charging pumps are running. x Pressurizer level continues to lower. The AOP now directs... A. isolating letdown to quantify leakage for E-plan classification. B. an immediate reactor trip to minimize dose rates at the site boundary. C. an immediate reactor trip due to leakage is excess of Tech Spec limits. D. isolating letdown to determine if leakage exceeds CVCS makeup capacity. Answer: D Reference Id: Q22453 Difficulty: 3.00 Time to complete: 4 10CFR Category: CFR 55.41 (10) 55.41 (10) Administrative, normal, abnormal, and emergency operating procedures for the facility. Cognitive Level: Comprehension / Anal Question Source: PV Bank Not Modified Proposed reference to be provided to applicant during examination: NONE Technical

Reference:

40AO-9ZZ02, Excessive RCS Leakrate K&A: Knowledge of the reasons for the following responses as they apply to the (Excess RCS Leakage) Manipulation of controls required to obtain desired operating results during abnormal, and emergency situations. Learning Objective: Given indications of RCS or a Steam Generator Tube Leak, describe the basic procedure methodology, including Reactor Trip is thresholds, in accordance with 40AO-9ZZ02. Justification: A. Incorrect: The E-plan numbers are determined by performing appendix A/B of 40AO-9ZZ02. B. Incorrect: Tripping the Reactor is determined as thresholds are exceeded after completing the next step to isolate letdown then trip if Pzr level continues to lower. C. Incorrect: TS limits are defined and if not met to be in mode 3 within 6 hours, not to trip immediately. Candidate may think the TS limits are trip thresholds. The next step is to isolate letdown then trip if Pzr level continues to lower. D. Correct: Isolating letdown eliminates the Letdown system as a possible location of the leak, Plant operation is allowed if the leak is isolated as exhibited by the restoration of Pzr Level. The step of the procedure is to isolate letdown and determine if CVCS makeup capability is exceeded if so then trip reactor. REV 0

ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Reactor Operator NRC Exam 26. This Exam Level: RO Appears on: RO EXAM 2012 Tier 1 Group 2 K/A #: 4.4 A13 AK1.2 Importance Rating: 3.2 Given the following conditions: x Unit 1 has been in a Blackout condition for 3 hours. x The crew is performing actions of 40EP-9EO08 (Blackout). x PBA-S03 has been energized by ONE Station Blackout Generator (SBOG) per Standard Appendix 80. x Attempts to restore power from other sources have been unsuccessful. The following parameters exist: x REP CET indicated 579°F and stable. x RCS pressure indicates 1540 psia and slowly lowering. x Pressurizer level indicates 23% and slowly lowering. x SG1 and SG2 levels are 47% WR and slowly rising. x Train "A" ADVs are throttled open approximately 25%. x SG1 and SG2 pressures indicate 1150 psig and stable. x SIAS setpoints have been reset as primary pressure lowers. Which ONE of the following describes the action(s) that will be taken by the crew? A. Use Auxiliary Spray to lower RCS pressure. B. Commence a cooldown to shutdown cooling entry conditions. C. ENSURE Train "A" ADVs are throttled adequately to maintain RCS subcooling. D. OVERRIDE and ENERGIZE Train "A" class backup heater to stabilize RCS pressure. Answer: C Reference Id: Q43811 Difficulty: 4.00 Time to complete: 3 10CFR Category: CFR 55.41 (10) 55.41 (10) Administrative, normal, abnormal, and emergency operating procedures for the facility. Cognitive Level: Comprehension / Anal Question Source: New Comment: Proposed reference to be provided to applicant during examination: Steam Tables Technical

Reference:

40EP-9EO08, BLACKOUT / 40DP-9AP13. BO Tech Guideline K&A: Knowledge of the operational implications of the following concepts as they apply to the (Natural Circulation Operations) Normal, abnormal and emergency operating procedures associated with (Natural Circulation Operations). REV 0

ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Reactor Operator NRC Exam Learning Objective: L56411 Given conditions of a Blackout state the action necessary to maintain subcooling margin in accordance with 40EP-9EO08. Justification: A. Incorrect - Lowering RCS pressure will cause subcooled margin to lower, which will not promote natural circulation conditions. B. Incorrect - This step is not required be performed unless AC power is not restored. PBA-S03 has been energized with a SBOG. C. Correct - Per Step 21 Blackout EOP, if the conditions are met, ENSURE proper control of steam generator steaming and feeding. D. Incorrect - Raising pressure would improve subcooling and promote natural circulation conditions. But Pressurizer Level is below the heater cutout setpoint, therefore Heaters are not available. REV 0

ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Reactor Operator NRC Exam 27. This Exam Level: RO Appears on: RO EXAM 2007 RO EXAM 2012 Tier 1 Group 1 K/A #: 4.4 E09 EA1.1 Importance Rating: 4.2 Given the following conditions: x The Unit 2 CRS has entered the Functional Recovery procedure. x RWT level is 6.4%. x You have been directed to verify proper Recirculation Actuation Signal (RAS). Which ONE of the following actions must be manually performed given a proper "A" train RAS actuation? A. Stop SIA-P01, LPSI pump A B. Close SIA-UV-666, HPSI A pump Recirc valve C. Open SIA-UV-674, Cntmt Sump to Safety Injection Valve D. Close CHA-HV-531, RWT to Train A Safety Injection Valve Answer: D Reference Id: Q10333 Difficulty: 3.00 Time to complete: 2 10CFR Category: CFR 55.41 (7) 55.41 (7) Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features. Cognitive Level: Memory Question Source: PV Bank Not Modified Comment: Proposed reference to be provided to applicant during examination: NONE Technical

Reference:

40EP-9EO09 (FRP) 40AO-9ZZ17 (Inadvertant PPS actuations) K&A: Ability to operate and / or monitor the following as they apply to the (Functional Recovery) Components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features. Learning Objective: Given the FRP is being performed and IC is in progress describe how the FRP will maintain or recover the Inventory Control Safety Function in accordance with 40EP-9EO09. Justification: A. Incorrect: LPSI pump are tripped on a RAS actuation. B. Incorrect: All SI miniflow valves close on RAS actuation. C. Incorrect: RAS sump isolation valves open on RAS actuation. D. Correct: RWT isolation valves must be manually operated on RAS actuation. REV 0

ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Reactor Operator NRC Exam 28. This Exam Level RO Appears on: RO EXAM 2008 RO EXAM 2012 Tier 2 Group 1 K/A # 3.4 003 A1.05 Importance Rating: 3.4 Given the following conditions: x Unit 1 is operating at 100% power. x RCP 1A experiences a failure causing it to slow down at 1% per minute. Assuming that all other input parameters remained the same, the CPC calculated value of DNBR will ... Answer: C Reference Id: Q44016 Difficulty: 3.00 Time to complete: 3 10CFR Category: CFR 55.41 (7) 55.41 (7) Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features. Cognitive Level: Comprehension / Anal Question Source: PV Bank Not Modified Comment: Proposed reference to be provided to applicant during examination: NONE Technical

Reference:

LOIT lesson plan K&A: Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the RCPs controls including: RCS flow Learning Objective: L77427 Describe the function of the Reactor Coolant Pump Speed inputs to the Core Protection Calculators. Justification: A. Incorrect: Pump Speed is input to the Flow calculation which is used in the DNBR calculation. DNBR will reduce as speed drops. A DNBR trip will be generated when RCP speed reaches 95%. B. Incorrect: Pump Speed is input to the Flow calculation which is used in the DNBR calculation. DNBR will reduce as speed drops. The auxiliary trip monitoring RCPs is generated when less than 2 RCPs are running. C. Pump Speed is input to the Flow calculation which is used in the DNBR calculation. DNBR will reduce as speed drops. A DNBR trip will be generated when RCP speed reaches 95%. D. Incorrect: DNBR will reduce as speed drops then generate a DNBR trip. The auxiliary trip monitoring RCPs is generated when less than 2 RCPs are running. OPTRNG_EXAM Page: 1 of 1 2012/01/10

ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Reactor Operator NRC Exam 29. This Exam Level: RO Appears on: RO EXAM 2005 RO EXAM 2012 K/A #: 3.4 003 K6.04 Importance Rating: 2.8 Given the following conditions: x Nuclear Cooling Water (NC) has been lost due to a pipe rupture. x Train 'B' Essential Cooling Water (EW) has been cross-connected to NC. Which ONE of the following describes a condition that will isolate 'B' Essential Cooling Water to the RCPs? A. Containment pressure rises to 9.0 psig. B. Pressurizer pressure drops to 1800 psia. C. Instrument air header pressure drops to 60 psig. D. 'B' EW Surge Tank level drops to LO LEVEL setpoint. Answer: A Reference Id: Q43945 Difficulty: 3.00 Time to complete: 2 10CFR Category: CFR 55.41 (8) 55.41 (8) Components, capacity, and functions of emergency systems. Cognitive Level: Comprehension / Anal Question Source: PV Bank Not Modified Comment: Proposed reference to be provided to applicant during examination: NONE Technical

Reference:

LOIT Lesson Plans, 40AO-9ZZ17 (Inadvertant PPS-ESFAS Actuations) K&A: Knowledge of the effect of a loss or malfunction on the following will have on the RCPs: Containment isolation valves affecting RCP operation. Learning Objective: Describe the automatic features associated with the NC Containment Isolation Valves. Justification: A. Correct: Containment Spray Actuation Signal (CSAS) at 8.5 psig will close the CTMT Isolation Valves for the NC system which are downstream of the EW cross tie valves. B. Incorrect: EW 'A' will isolate on SIAS EW 'B' cross tie valves are manually operated valves with no automatic features. C. Incorrect: NC and EW valves are Motored Operated valves, the degraded Instrument Air Header pressure will not effect EW to RCPs. 40AO-9ZZ06 (Loss of IA) describes hundreds of components that are effected by the lowering IA header pressure. D. Incorrect: EW 'A' will isolate on LO 'A' EW Surge Tank Level. EW 'B' cross tie valves are manually operated valves with no automatic features. REV 0

ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Reactor Operator NRC Exam 30. This Exam Level: RO Appears on: RO EXAM 2012 Tier 2 Group 1 K/A #: 3.2 004 K1.04 Importance Rating: 3.4 Given the following conditions: x Unit 3 is operating at 100% power. x All RCP seal injection controllers (CHN-FIC-241-244) are in automatic. x The output SIGNAL of CHN-FIC-241, 1A RCP controller, is rising. x Disregard the response of the remaining Seal Injection controllers. Which ONE of the following describes the cause? A. NNN-D11 is de-energized. B. Inadvertent CSAS actuation. C. Actual Seal Injection flow is below setpoint. D. Regenerative Heat Exchanger outlet temperature has exceeded 413ºF. Answer: B Reference Id: Q10468 Difficulty: 3.00 Time to complete: 3 10CFR Category: CFR 55.41 (7) 55.41 (7) Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features. Cognitive Level: Comprehension / Anal Question Source: PV Bank Not Modified Comment: Proposed reference to be provided to applicant during examination: NONE Technical

Reference:

LOIT Lesson Plans K&A: Knowledge of the physical connections and/or cause-effect relationships between the CVCS and the following systems: RCPS, including seal injection flows. Learning Objective: L68108 Explain the operation of the RCP Seal Injection Flow Control Valves (CHE-FV-241,242,243, and 244), including their Control Room controls, under normal operating conditions. Justification: A. Incorrect: Loss of NNN-D11 will de-energize the controller therefore the output will be failed as is. B. Correct: CSAS actuation will isolate IA to the Containment and valves will slowly open, therefore controller will try to lower flow by raising output. These controllers are reverse acting. C. Incorrect: Actual Flow less than setpoint will cause the controller output to lower. Reverse Acting Controller. D. Incorrect: This will provide a close signal to CHB-UV-515, This Loss of Letdown will not effect seal injection flow. REV 0

ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Reactor Operator NRC Exam 31. This Exam Level: RO Appears on: RO EXAM 2012 Tier 2 Group 1 K/A #: 3.4 005 K3.07 Importance Rating: 3.2 Given the following plant conditions: x Refueling pool level is 137' 6' (>23 ft above the vessel flange). x Core RE-LOAD is in progress. x An irradiated fuel assembly is grappled and in the hoist box. x Train 'B' is under clearance for maintenance. x Train 'A' LPSI pump is gas bound. Which ONE of the following complies with Technical Specifications 3.9.4 (Shutdown Cooling (SDC) and Coolant Circulation - High Water Level) required actions ? A. Core re-load may continue. B. Immediately stop core re-load, leave the fuel assembly in the hoist box. C. Complete placing the fuel assembly in its designated core location, then suspend core re-load. D. Immediately stop core re-load until you have verification that all activities that could result in boron dilution have been suspended. Answer: B Reference Id: Q43947 Difficulty: 4.00 Time to complete: 2 10CFR Category: CFR 55.41 (10) 55.41 (10) Administrative, normal, abnormal, and emergency operating procedures for the facility. Cognitive Level: Comprehension / Anal Question Source: New Proposed reference to be provided to applicant during examination: NONE Technical

Reference:

Technical Specifications 3.9.4 (Shutdown Cooling (SDC) and Coolant Circulation - High Water Level) and Basis. K&A: Knowledge of the effect that a loss or malfunction of the RHRS will have on the following: Refueling operations. Learning Objective: L94060 Given a set of plant conditions identify whether or not LCO 3.9.4 is satisfied and any actions or surveillance requirements that would prevent core alterations per Tech Spec 3.9 and its Basis. Justification: A. Incorrect: Core Off Load would be permitted in this instance but Core Re Load would add energy to the core. B. Correct: Per TS 3.9.4 One SDC Cooling Loop shall be operable and in operation. The fact that B has no power and A is gas bound Condition A is not met and loading irradiated fuel must be suspended immediately. C. Incorrect: The fuel assembly would be placed back in its original position in the Spent Fuel Pool not the Core. D. Incorrect: Immediately suspending core reload is correct but once the boron concentration reduction is verified to not exist you may not restart the core re load. REV 0

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ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Reactor Operator NRC Exam 33. This Exam Level: RO Appears on: RO EXAM 2012 Tier 2 Group 1 K/A #: 3.5 007 A2.05 Importance Rating: 3.2 Given the following conditions: x Unit 2 is operating at 100% power. x PSV-203 (PZR safety valve) has seat leakage. x RDT level is rising. x RDT pressure is 9.8 psig and rising slowly. Which ONE of the following automatic actions will occur if NO operator action is taken? CHN-UV-540 (RDT Vent to Gas Surge Tank) will... A. OPEN and CHN-HV-923 (RDT Atmospheric Vent Isolation) will OPEN. B. CLOSE and CHN-HV-923 (RDT Atmospheric Vent Isolation) will OPEN. C. OPEN and CHA-UV-560 (RDT outlet containment isolation valve) will CLOSE. D. CLOSE and CHA-UV-560 (RDT outlet containment isolation valve) will CLOSE. Answer: D Reference Id: Q43950 Difficulty: 2.00 Time to complete: 52 10CFR Category: CFR 55.41 (3) 55.41 (3) Mechanical components and design features of the reactor primary system. Cognitive Level: Comprehension / Anal Question Source: New Comment: Proposed reference to be provided to applicant during examination: NONE Technical

Reference:

40AL-9RK4A (B04 ARP) K&A: Ability to (a) predict the impacts of the following malfunctions or operations on the PRTS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Exceeding PRT high-pressure Learning Objective: Describe automatic functions associated with the following Reactor Drain Tank Valves:* CHA-UV-560 (Reactor Drain Tank Outlet Isolation Valve)* CHB-UV-561 (Reactor Drain Tank Outlet Isolation Valve)* CHN-UV-540 (Reactor Drain Tank Vent Valve)* CHA-UV-580 (Reactor Drain Tank Makeup Supply Isolation Valve). Justification: REV 0

ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Reactor Operator NRC Exam A. Incorrect: CHN-UV-540 is the normal vent path with RDT Pressure greater than 5 psig but less than 10 psig. CHN-UV-540 will Auto Close at 10 psig. CHN-HV-923 is the correct vent path for RDT pressures greater than 10 psig, it has NO Auto Functions B. Incorrect: CHN-UV-540 will Auto Close at 10 psig. CHN-HV-923 is the correct vent path for RDT pressures greater than 10 psig, it has NO Auto Functions C. Incorrect:CHA-UV-560 will also Auto Close at 10 psig. D. Correct: CHN-UV-540 is the normal vent path with RDT Pressure greater than 5 psig but less than 10 psig. CHA-UV-560 will also Auto Close at 10 psig. REV 0

ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Reactor Operator NRC Exam 34. This Exam Level: RO Appears on: RO EXAM 2007 RO EXAM 2012 Tier 2 Group 1 K/A #: 3.8 008 K2.02 Importance Rating: 3.0 Given the following conditions: x Unit 1 is operating at 100% power. x NCN-P01A (NCW PUMP A) is in operation with NCN-P01B (NCW PUMP B) in standby. x The A Emergency Diesel Generator is under permit for maintenance. x NBN-X03 ESF Service Transformer fails. x This loss does NOT result in a Reactor Trip. Based on these conditions, the Nuclear Cooling Water system will... A. have no pumps running. B. be unaffected (no change in pump operation). C. remain in operation, however NCN-P01B is now running. D. remain in operation, with both NCN-P01A and NCN-P01B in operation. Answer: B Reference Id: Q5794 Difficulty: 3.00 Time to complete: 3 10CFR Category: CFR 55.41 (7) 55.41 (7) Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features. Cognitive Level: Comprehension / Anal Question Source: PV Bank Not Modified Comment: Proposed reference to be provided to applicant during examination: NONE Technical

Reference:

LOIT Lesson Plan K&A: Knowledge of bus power supplies to the following: CCW Pump, including emergency backup. Learning Objective: 64988 Explain the operation of the NC Pumps under normal operating conditions. REV 0

ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Reactor Operator NRC Exam Justification: A. Incorrect: Candidate may think that the NCW pumps are powered from PB buses and may think this situation has resulted in a loss of power to both. B. Correct: NCW pumps are powered from non-class 4160v busses NBN-S01 and NBN-S02. Losing transformer NBN-X03 with the A Diesel Generator tagged out will result in a loss of Class 4160v power on the A train, but will not affect power to the NCW pumps. C. Incorrect: May think that PBA has lost power and NCW A with it, NCW B would start on low header pressure. D. Incorrect: May think that the power transfer from off site to the EDG would result in both pumps running. REV 0

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ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Reactor Operator NRC Exam 36. This Exam Level: RO Appears on: RO EXAM 2012 Tier 2 Group 1 K/A #: 37 012 K4.06 Importance Rating: 3.2 The DNBR/LPD Reactor Protection System Operational Bypass is inserted ____(1)____ when the Excore NI Power decreases below ____(2)____ % A. (1) manually (2) 1E-2%. B. (1) manually (2) 1E-4%. C. (1) automatically (2) 1E-2%. D. (1) automatically (2) 1E-4%. Answer: B Reference Id: Q43995 Difficulty: 3.00 Time to complete: 2 10CFR Category: CFR 55.41 (7) 55.41 (7) Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features. Cognitive Level: Memory Question Source: PV Bank Not Modified Comment: Proposed reference to be provided to applicant during examination: NONE Technical

Reference:

LOIT Lesson Plan K&A: Knowledge of RPS design feature(s) and/or interlock(s) which provide: Automatic or manual enable/disable of RPS trips Learning Objective: L77084 Plant Protection System, Describe the RPS operating bypasses. Justification: A. Incorrect: It is inserted manually but is enabled below 1E-4%. 1E-2% is the Log Power Bypass. B. Correct:The bypass must be manually inserted from key switches at the remote CPC modules on B05 when ex-core safety channel NI power is less than 10-4% power. C. Incorrect: It is inserted manually. 1E-2% is the Log Power Bypass. D. Incorrect: It is inserted manually. REV 0

ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Reactor Operator NRC Exam 37. This Exam Level: RO Appears on: RO EXAM 2012 Tier 2 Group 1 K/A #: 3.7 012 A4.04 Importance Rating: 3.3 Given the following conditions: x Unit 1 is operating at 100% power x Channel 'D' PPS HI PZR PRESS is BYPASSED due to a failed high RCS pressure (Narrow Range) transmitter. x Channel 'B' PPS SG-2 level low has TRIPPED due to failed transmitter. x Channel 'A' RCS pressure (Narrow Range) transmitter now FAILS HIGH. Based on these conditions, which ONE of the following is correct? A. The operator can NOT physically bypass channel 'A' HI PZR PRESS bistable. B. The reactor would have tripped when the channel 'A' pressure transmitter failed. C. 2 Reactor Trip Circuit Breakers (RTCBs) would open when the channel 'A' RCS pressure transmitter failed, but the reactor would not trip. D. If the operator bypasses the 'A' HI PZR PRESS bistable, that channel would go into bypass, while removing the channel 'D' HI PZR PRESS bistable from bypass. Answer: D Reference Id: Q43953 Difficulty: 2.00 Time to complete: 2 10CFR Category: CFR 55.41 (5) 55.41 (5) Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics. Cognitive Level: Comprehension / Anal Question Source: New Comment: Proposed reference to be provided to applicant during examination: NONE Technical

Reference:

LOIT lesson plan K&A: Ability to manually operate and/or monitor in the control room: Bistable, trips, reset and test switches. Learning Objective: L77088 Describe the RPS Trip Channel bypass interlock. REV 0

ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Reactor Operator NRC Exam Justification: An electrical interlock prevents the operator from bypassing more than one trip channel at a time for any one type of trip. Different type trips may be bypassed simultaneously, either in one channel or in different channels. Attempting to insert a trip channel bypass in a second channel for the same type of trip will result in only the Highest priority channel being in bypass, with A being the highest, and D the lowest priority. If C channel Pressurizer pressure had tripped and was bypassed and A or B channel was subsequently bypassed, C would come out of bypass and trip. A. Incorrect: The operator CAN bypass the A RCS Press Transmitter. B. Incorrect: In this case the coincidence is 2/3 with the D channel bypassed. 2/4 is the normal coincidence which would result in a trip. C. Incorrect: This will not result in any RTSG breakers opening. RTSG breakers do not open on the specific parameter, only the Channel trip. D. Correct: Per the explanation above, the hierarchy of the system would cause the D channel to come out of bypass when the A channel is placed in bypass. REV 0

ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Reactor Operator NRC Exam 38. This Exam Level: RO Appears on: RO EXAM 2007 RO EXAM 2012 Tier 2 Group 1 K/A #: 3.2 013 A4.03 Importance Rating: 3.9 Given the following conditions: x Unit 1 is operating at 100% power. x The CRS directs an RO to initiate a MSIS from the Aux Relay Cabinets. x The RO performs the following actions: x Depresses the 1-3 and 2-4 MSIS trip pushbuttons simultaneously on the "A" train. x Depresses the 1-3 and 2-4 MSIS trip pushbuttons sequentially (push then release) on the "B" train. Assuming that SG pressures remains above the MSIS setpoint, you would expect an "A" train MSIS full initiation with... A. no initiation of the "B" train, "A" MSIS can be reset by depressing either reset pushbutton. B. a half leg initiation of the "B" train, "A" MSIS can be reset by depressing either reset pushbutton. C. no initiation of the "B" train, "A" MSIS can only be reset by depressing both reset pushbuttons simultaneously. D. a half leg initiation of the "B" train, "A" MSIS can only be reset by depressing both reset pushbuttons simultaneously. Answer: A Reference Id: Q44012 Difficulty: 4.00 Time to complete: 3 10CFR Category: CFR 55.41 (7) 55.41 (7) Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features. Cognitive Level: Memory Question Source: PV Bank Not Modified Comment: Proposed reference to be provided to applicant during examination: NONE Technical

Reference:

73ST-9DG01(ISG testing) K&A: Knowledge of ESFAS design feature(s) and/or interlock(s) which provide for the following: Main Steam Isolation System. Learning Objective: Describe how an ESFAS subsystem can be manually actuated and manually reset from the Aux Relay Cabinets. Justification: REV 0

ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Reactor Operator NRC Exam A. Correct: To initiate an ESFAS actuation both buttons must be pushed sim. pushing and releasing gives no initiation half leg or otherwise, power is still available to all relays. Resetting requires that either reset button on the train be depressed. B. Incorrect: No initiation of the B train will occur, the MSIS can be reset by pushing either Aux Relay Cabinet Pushbutton. C. Incorrect: No initiation is correct for the B train, but you don't have to press both Aux Relay Cabinet Pushbuttons to reset. D. Incorrect: No initiation of the B train will occur, but you don't have to press both Aux Relay Cabinet Pushbuttons to reset. REV 0

ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Reactor Operator NRC Exam 39. This Exam Level: RO Appears on: RO EXAM 2012 Tier 2 Group 1 K/A #: 3.5 022 A4.01 Importance Rating: 3.6 Given the following conditions: x Unit 3 is operating at 100% power. x An Inadvertent SIAS has occurred. Which ONE of the following describes the status of the Containment Normal ACUs? The Containment Normal ACUs... A. continue to run. B. are load shed and must be manually started by an operator. C. are load shed and will sequence back on after 120 seconds. D. shift to take suction on elevations 100' and below in containment. Answer: B Reference Id: Q43955 Difficulty: 4.00 Time to complete: 3 10CFR Category: CFR 55.41 (7) 55.41 (7) Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features. Cognitive Level: Memory Question Source: New Comment: Proposed reference to be provided to applicant during examination: NONE Technical

Reference:

40AO-9ZZ17(Inadvertent PPS ESFAS), LOIT Lesson Plans K&A: Ability to manually operate and/or monitor in the control room: CCS fans Learning Objective: Describe the automatic functions associated with the Containment Building Normal ACU Fans (HCN-A01-A, B, C, & D) . Justification: A. Incorrect: Not all HVAC system respond to a SIAS, the AUX Building HVAC system does not respond to a SIAS. B. Correct: This is correct, the Containment Normal ACUs will receive a Load Shed signal on the SIAS and need to be manually restarted by a operator. C. Incorrect: The Load Shed portion is correct but the 120 Seconds is the time delay associated with the CEDM ACUs. D. Incorrect: On a SIAS the Fuel Building HVAC system will shift suctions to the Aux Building 100 foot elevation and below. REV 0

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ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Reactor Operator NRC Exam 41. This Exam Level: RO Appears on: RO EXAM 2012 Tier 2 Group 1 K/A #: 3.4 039 K3.03 Importance Rating: 3.2 Given the following conditions: x Unit 2 has tripped from 100% power. x S/G #1 level is 23% WR and lowering rapidly. x S/G #1 pressure is 780 psia and lowering rapidly. x S/G #2 level is 28% WR and lowering slowly. x S/G #2 pressure is 1050 psia and stable. Assuming NO operator action, AFA-P01 (Essential Turbine Driven Aux Feed Pump) is... A. still in standby. B. operating and aligned to receive steam from BOTH SGs. C. operating and aligned to receive steam from SG #1 ONLY. D. operating and aligned to receive steam from SG #2 ONLY. Answer: B Reference Id: Q43957 Difficulty: 3.00 Time to complete: 3 10CFR Category: CFR 55.41 (4) 55.41 (4) Secondary coolant and auxiliary systems that affect the facility. Cognitive Level: Comprehension / Anal Question Source: New Comment: Proposed reference to be provided to applicant during examination: NONE Technical

Reference:

LOIT Lesson Plan K&A: Knowledge of the effect that a loss or malfunction of the MRSS will have on the following: AFW pumps. Learning Objective: Explain the operation of the AFW Pump Turbine Main Steam Supply Valves (SGA-UV-134 and -138) under normal operating conditions. Justification: A. Incorrect: Both MOVs will open on the AFAS signal that was received at 25.8% WR on the #1 SG. Candidate may not know the AFAS setpoint. Also, Candidate may think the D/P lockout of 185 psid will not allow the lower pressure SG to supply steam to AFA-P01. B. Correct: Both Main Steam Supply valves AUTO open on an AFAS actuation, regardless of which SG has experienced the low level. In addition, the D/P lockout does NOT impact the operation of the steam supply valves. C. Incorrect: Candidate may think only the SG that is below the AFAS setpoint will supply steam to AFA-P01. D. Incorrect: Candidate may think only the SG that is INTACT will supply steam to AFA-P01 due to the D/P lockout. REV 0

ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Reactor Operator NRC Exam 42. This Exam Level: RO Appears on: RO EXAM 2012 Tier 2 Group 1 K/A #: 3.4 059 A1.03 Importance Rating: 2.7 Which ONE of the following describes the operation of the Main Feedwater Pumps during a power ascension above 20% Power. In accordance with 40OP-9ZZ05 (Power Operations) the second Main Feedwater Pump must be started prior to... A. exceeding 60% reactor power. B. placing 2nd stage reheat in service. C. MFWP suction pressure lowering below 300 psia. D. MFWP discharge pressure and SG pressure delta P dropping below 100 psid. Answer: A Reference Id: Q43958 Difficulty: 3.00 Time to complete: 2 10CFR Category: CFR 55.41 (4) 55.41 (4) Secondary coolant and auxiliary systems that affect the facility. Cognitive Level: Memory Question Source: New Comment: Proposed reference to be provided to applicant during examination: NONE Technical

Reference:

40OP-9ZZ05 (Power Operations) K&A: Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the MFW controls including: Power level restrictions for operation of MFW pumps and valves. Learning Objective: L82548 Explain the operation of the MFWPs under normal operating conditions. Justification: A. Correct: Per NOTE after 4.3.43 the 2nd MFWP must be started to prevent damage to the 1st MFWP turbine. B. Incorrect: The minimum suction pressure for the MFWP is 300 psig. This threshold has you start the 3rd condensate pump. C. Incorrect:Placing the 2nd stage reheat is done after reaching 15% power. This is not a milestone for placing the 2nd MFWP in service. Starting a second MFWP would cause suction pressure to lower. D. Incorrect: 100 psid is the lower limit at 100% power and is not a parameter used for starting a 2nd MFWP. REV 0

ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Reactor Operator NRC Exam 43. This Exam Level: RO Appears on: RO EXAM 2005 RO EXAM 2012 Tier 2 Group 1 K/A # 3.4 005 K5.03 Importance Rating: 2.9 Given the following conditions: x Unit 2 is in Mode 5 following refueling. x Shutdown Cooling in service using LPSI 'A'. x Shutdown Cooling Train B is lined up for SDC but has not been recirculated. x LPSI Pump 'A' trips due to a fault condition. x RCS Pressure is 360 psia. Which ONE of the following describes the potential concern with swapping the SDC alignment to 'B' Train at this time? A. The LTOP could lift when the "B" SDC Loop is exposed to the RCS. B. The colder water in Loop B could cause the 19°F per minute heatup rate limit on the SDC loop to be exceeded. C. The minimum temperature limit of 350°F will be violated by swapping the SDC loops at this time without first recirculating the standby loop. D. The "B" Shutdown Cooling loop may have a different boron concentration than the RCS and may have to be equalized to prevent an unacceptable RCS boron concentration change. Answer: D Reference Id: Q10202 Difficulty: 3.00 Time to complete: 3 10CFR Category: CFR 55.41 (10) 55.41 (10) Administrative, normal, abnormal, and emergency operating procedures for the facility. Cognitive Level: Comprehension / Anal Question Source: New Comment: Proposed reference to be provided to applicant during examination: NONE Technical

Reference:

40OP-9SI01 (Shutdown Cooling Initiation) K&A: Knowledge of the operational implications of the following concepts as they apply the RHRS: Reactivity effects of RHR fill water. Learning Objective: L79915 Discuss the concerns with boron concentration associated with the Shutdown Cooling System. Justification: A. Incorrect: This is incorrect, the LTOP lift pressure is 467 psia which is greater than the 360 psia of the RCS currently. B. Incorrect: The water will be colder which would result in a cooldown not a heatup. C. Incorrect: The average bulk temperature should lower when introduced into the system, therefore the 350 F limit will not be approached. D. Correct: The Precautions and Limitations of the OP describe the fact that an Idle SDC loop may have a different boron concentration. REV 0

ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Reactor Operator NRC Exam 44. This Exam Level: RO Appears on: RO EXAM 2012 Tier 2 Group 1 K/A #: 3.4 061 K4.02 Importance Rating: 4.5 Given the following conditions: Initial Conditions: x Unit 1 is in Mode 3 following an automatic reactor trip.. x AFN-P01 (Non-Essential Motor Driven Aux Feed Pump) is feeding both SGs at 350 gpm. x AFB-P01 (Essential Motor Driven Aux Feed Pump) is in standby. x AFA-P01 (Essential Turbine Driven Aux Feed Pump) is in standby. Subsequently: x Pressurizer pressure lowers to 1700 psia. Which ONE of the following describes the status of the Auxiliary Feedwater System One minute after the Pzr Pressure reaches 1700 psia? AFN-P01... A. has tripped, AFB-P01 starts and feeds the SGs. B. is running and feeding both SGs. AFB-P01 is in standby status. C. has tripped, AFB-P01 starts but must be manually aligned to feed the SGs. D. is running with its feedpath isolated, AFB-P01 starts but must be manually aligned to feed the SGs. Answer: C Reference Id: Q44004 Difficulty: 3.00 Time to complete: 3 10CFR Category: CFR 55.41 (7) 55.41 (7) Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features. Cognitive Level: Comprehension / Anal Question Source: New Comment: Proposed reference to be provided to applicant during examination: NONE Technical

Reference:

01-M-AFP-001 (Auxiliary Feedwater System Print) K&A: Knowledge of AFW design feature(s) and/or interlock(s) which provide for the following: AFW automatic start upon loss of MFW pump, S/G level, blackout, or safety injection. Learning Objective: Describe the Control Room controls associated with the Essential Auxiliary Feedwater Pump AFB-P01 including it's indications. OPTRNG_EXAM Page: 1 of 2 2011/11/03

ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Reactor Operator NRC Exam Justification: A. Incorrect: AFN will trip on the load shed stop signal initiated by the SIAS, AFB does automatically start on the SIAS, only AFAS will open the Feed valves therefore AFB will not be feeding the SGs. B. Incorrect: AFN will trip on the load shed stop signal initiated by the SIAS, therefore the no feed will be supplied to the SGs. C. Correct: AFN will trip on the load shed stop signal initiated by the SIAS, AFB does automatically start on the SIAS, only AFAS will open the Feed valves therefore AFB will have to be manually aligned to feed the SGs. D. Incorrect: AFN will trip on the load shed stop signal initiated by the SIAS, the downcomer isolations will remain open so the AFN feedpath is not isolated. AFB does automatically start on the SIAS, only AFAS will open the Feed valves therefore AFB will have to be manually aligned to feed the SGs. OPTRNG_EXAM Page: 2 of 2 2011/11/03

ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Reactor Operator NRC Exam 45. This Exam Level: RO Appears on: RO EXAM 2012 Tier 2 Group 1 K/A #: 3.6 062 K4.03 Importance Rating: 2.8 Given the following list of conditions:

1. The BUS XFR SWITCH must be in AUTO.

2. A generator trip must have occurred.

3. The synchroscope must be on.

4. The synch check relay must be satisfied.

5. A Unit Aux Transformer trip must have occurred.

6. A lockout of the Normal Supply breaker must have occurred.

Which ONE of the following describes the conditions that must be met for an automatic Fast Bus Transfer of NAN-S01 to NAN-S03 to occur? This is not an all inclusive list. A. 1, 2 and 4 B. 1, 4 and 6 C. 2, 3 and 6 D. 3, 4 and 5 Answer: A Reference Id: Q43959 Difficulty: 3.00 Time to complete: 2 10CFR Category: CFR 55.41 (7) 55.41 (7) Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features. Cognitive Level: Memory Question Source: New Comment: Proposed reference to be provided to applicant during examination: NONE Technical

Reference:

LOIT Lesson Plans K&A: Knowledge of ac distribution system design feature(s) and/or interlock(s) which provide for the following: Interlocks between automatic bus transfer and breakers Learning Objective: Explain the operation of Switchgear NAN-S01 and NAN-S02 under normal operating conditions. REV 0

ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Reactor Operator NRC Exam Justification: The NAN-S01 and NAN-S02 buses are designed with the ability to allow a fast bus transfer from the unit auxiliary transformer source to the NAN-S03 and NAN-S04 source. The feature allows the 13.8 kV bus loads to remain energized in the event of a loss of the main generator, the normal in-house supply. If the main turbine/generator trips at 100% power, the reactor can remain critical following the load rejection as the reactor coolant pumps remain powered. The sequence of events and associated interlocks that initiate a 13.8 kV NA fast bus transfer is listed below. In order to allow a NA fast bus transfer, the manual/auto transfer switch on the control room B01 panel must be in auto. The initiating event for a NA fast bus transfer is always a main generator trip . The activation of this lockout initiates the opening of the unit auxiliary transformer supply breakers, NAN-S01A and NAN-S02A. An automatic sync check is performed between the NAN-S01 to NAN-S03 and NAN-S02 to NAN-S04 bus. If the two sources are in sync, this contact is closed. Buses NAN-S03/S04 are checked for normal voltage and frequency. Both the unit auxiliary supply breaker and the bus tie breakers are checked for tripped 86 lockout relays. If both are reset, the close signal is allowed to pass on to the bus tie breakers, NAN-S03B/S04B. A. Correct: These 3 are required to have an automatic Fast Bus Transfer. B. Incorrect: 1 and 4 are correct but 6 is not. Lockout on the normal supply breaker would prevent the FBT from occurring. Candidate may believe a UAT Trip is required vice a Main Turbine Trip. C. Incorrect: 2 is correct, but 3 and 6 are not. Synch Check is automatically performed the synchroscope is not required for this check. Lockout on the normal supply breaker would prevent the FBT from occurring. D. Incorrect: 4 is correct but 3 and 5 are not. Synch Check is automatically performed the synchroscope is not required for this check. UAT trip may be confused for the Main Turbine Trip requirement. REV 0

ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Reactor Operator NRC Exam 46. This Exam Level: RO Appears on: RO EXAM 2012 Tier 2 Group 1 K/A #: 3.6 062 A3.01 Importance Rating: 3.0 Given the following conditions: x Unit 2 has tripped from 100% power. x NAN-X01 (S/U XFMR #1) has faulted. x SIAS has actuated. x EDG 'A' is at 60.1 Hz and 4200 VAC. x No 86 Lockouts on PBA-S03. x Normal/Alternate Supply Breakers to PBA-S03 have operated as designed. Which ONE of the following describes the status of the... (1) EDG 'A' output breaker? (2) Amperage Indication on Load Centers supplied by PBA-S03? (3) NHN-M71Energized/Not Energized? A. (1) OPEN (2) AMPS INDICATED (3) NOT ENERGIZED B. (1) OPEN (2) AMPS NOT INDICATED (3) ENERGIZED C. (1) CLOSED (2) AMPS INDICATED (3) NOT ENERGIZED D. (1) CLOSED (2) AMPS NOT INDICATED (3) ENERGIZED Answer: C Reference Id: Q43962 Difficulty: 3.00 Time to complete: 2 10CFR Category: CFR 55.41 (7) 55.41 (7) Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features. Cognitive Level: Comprehension / Anal Question Source: New Comment: Proposed reference to be provided to applicant during examination: NONE Technical

Reference:

LOIT Lesson Plan, Electrical Distribution Drawing K&A: Ability to monitor automatic operation of the ac distribution system, including: Vital ac bus amperage Learning Objective: Describe the Local and Control Room indications associated with the Class IE AC Electrical Distribution System. REV 0

ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Reactor Operator NRC Exam Justification: NAN-X01 (Startup Transformer #1) is the normal supply to NAN-S05 which supplies PBA-S03 thru its associated ESF Transformer. This fault will cause an undervoltage condition on PBA-S03. Candidate may not know the S/U XFMR arrangement and believe that PBA-S03 is still being powered from off site power. EDG 'A' meets the requirements to automatically close in on PBA-S03 and power the bus. These requirements are Frequency between 59.9 and 60.5 Hz. Voltage between 4080 and 4300 Volts. No lockouts on the bus. Normal and Alternate supply breakers are open. Due to the EDG 'A' supplying PBA-S03 Amps will be indicated. NHN-M71 is a SIAS Load Shed Panel that will be de-energized due to the SIAS. A. Incorrect: (1) EDG 'A' meets the requirements to automatically close in on PBA-S03 and power the bus. (2) Due to the EDG 'A' supplying PBA-S03 Amps will be indicated. (3) NHN-M71 is a SIAS Load Shed Panel that will be de-energized due to the SIAS. B. Incorrect: (1) EDG 'A' meets the requirements to automatically close in on PBA-S03 and power the bus. (2) Due to the EDG 'A' supplying PBA-S03 Amps will be indicated. (3) NHN-M71 is a SIAS Load Shed Panel that will be de-energized due to the SIAS. C. Correct: These are all correct. D. Incorrect: (1) EDG 'A' meets the requirements to automatically close in on PBA-S03 and power the bus. (2) Due to the EDG 'A' supplying PBA-S03 Amps will be indicated. (3) NHN-M71 is a SIAS Load Shed Panel that will be de-energized due to the SIAS. REV 0

ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Reactor Operator NRC Exam 47. This Exam Level RO Appears on: RO EXAM 2012 Tier 2 Group 1 K/A # 3.6 063 K2.01 Importance Rating: 2.9 Which ONE of the following valves are powered from a vital 125 VDC control centers? A. SIA-UV-644, SIT Isolation B. SID-UV-654, Shutdown Cooling Isolation C. SIE-HV-661, Combined SIT Drain to RDT D. SIB-HV-690, Shutdown Cooling Loop 1 Warm-up Bypass Answer: B Reference Id: Q43972 Difficulty: 2.00 Time to complete: 2 10CFR Category: CFR 55.41 (7) 55.41 (7) Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features. Cognitive Level: Memory Question Source: New Comment: Proposed reference to be provided to applicant during examination: NONE Technical

Reference:

LOIT Lesson Plan K&A: Knowledge of bus power supplies to major DC loads. Learning Objective: Knowledge of major DC loads Justification: A. Incorrect: The SIT Isolation valves are powered by class 480v MCCs. B. Correct: Class DC electrical distribution trains "C" and "D" provide power to the Shutdown Cooling Isolation Valves through inverters PKC-N43 and PKD-N44. C. Incorrect:The SIT Drains are air operated. D. Incorrect: The Shutdown Cooling Loop Warm-up Bypasses are powered class 480 v MCCs. REV 0.

ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Reactor Operator NRC Exam 48. This Exam Level: RO Appears on: RO EXAM 2007 RO EXAM 2012 Tier 2 Group 1 K/A #: 063 2.1.27 Importance Rating: 3.9 Given the following conditions: x Unit 1 is in Mode 5. x Battery Charger "A" (PKA-H11) has tripped. x Battery Charger "AC" (PKA-H15) is connected to the "C" Battery bus (PKC-M43). Can the "AC" Battery Charger be aligned to both PKA-M41 and PKC-M43 at this time? A. YES, provided the Unit remains in Mode 5. B. NO, a mechanical interlock prevents this alignment. C. YES, provided that the "A" battery is disconnected from PKA-M41. D. NO, this action may only occur while restoring the MVDC safety functions as implemented by the Lower Mode Functional Recovery Procedure. Answer: B Reference Id: Q44002 Difficulty: 2.00 Time to complete: 2 10CFR Category: CFR 55.41 (7) 55.41 (7) Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features. Cognitive Level: Memory Question Source: PV Bank Not Modified Comment: Proposed reference to be provided to applicant during examination: NONE Technical

Reference:

LOIT Lesson Plan K&A: Knowledge of system purpose and or function: DC Electrical Distribution Learning Objective: L74205 Explain the operation of the Class IE 125 VDC Battery Chargers under normal operating conditions. PRA SIGNIFICANT QUESTION Justification: A. Incorrect: Tech Specs 3.8.1 do not allow for the class busses to be cross tied in Modes 1-4. Candidate may think that since the unit is in mode 5 this may not apply. B. Correct: PVNGS has a mechanical interlock that prevents the Swing chargers from connecting to multiple DC buses simultaneously C. Incorrect: Batteries are not allowed to be crosstied to the same bus, if the A battery was disconnected this would remove that obstacle to crosstying the busses, but the mechanical interlock is not disabled when the battery is disconnected from the bus. D. Incorrect: LMFRP has many instances where DC busses are restored. Candidate may believe that the crosstying is one of them. REV 0

ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Reactor Operator NRC Exam 49. This Exam Level RO Appears on: RO EXAM 2012 Tier 2 Group 1 K/A # 3.6 064 K6.07 Importance Rating: 2.7 Given the following list of conditions: x Unit 1 is operating at 100% power. x The DG A right bank Starting Air Receiver is tagged out. x There was an Inadvertent Containment Spray System Actuation. The remaining left bank receiver and starting air subsystem will apply air to ____(1)____ diesel cylinder bank(s) and the diesel starts in the ____(2)_____ mode. A. (1) both (2) Test Run B. (1) both (2) Emergency C. (1) only the left (2) Test Run D. (1) only the left (2) Emergency Answer: A Reference Id: Q43971 Difficulty: 3.00 Time to complete: 2 10CFR Category: CFR 55.41 (7) 55.41 (7) Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features. Cognitive Level: Comprehension / Anal Question Source: New Proposed reference to be provided to applicant during examination: NONE Technical

Reference:

LOIT Lesson Plan PRA SIGNIFICANT QUESTION K&A: K6.07 Knowledge of the effect of a loss or malfunction of the following will have on the ED/G system: Air receivers Learning Objective: Describe the operation of the Diesel Generator Air Starting Sub-system under normal conditions. REV 0

ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Reactor Operator NRC Exam Justification: A. Correct: Crossover piping allows starting air to to be supplied to both banks of diesel cylinders. The diesel starts in the test run mode of operation on an inadvertent Containment Spray System actuation. B. Incorrect:Crossover piping allows starting air to to be supplied to both banks of diesel cylinders. The diesel does not start in the Emergency run mode of operation on an inadvertent Containment Spray System actuation. C. Incorrect:Crossover piping allows starting air to to be supplied to both banks of diesel cylinders. The diesel starts in the test run mode of operation on an inadvertent Containment Spray System actuation. D. Incorrect:Crossover piping allows starting air to to be supplied to both banks of diesel cylinders. The diesel does not start in the Emergency run mode of operation on an inadvertent Containment Spray System actuation. REV 0

Palo Verde Operating Experience ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Reactor Operator NRC Exam 50. This Exam Level: RO Appears on: RO EXAM 2012 Tier 2 Group 1 K/A #: 3.6 064 A1.08 Importance Rating: 3.1 While setting up a Diesel Generator to be paralleled with off-site power the following parameters are noted just before the output breaker is closed; x The synchroscope is moving slowly in the fast direction. x Grid frequency 59.9 Hz x Diesel RPM 600 x Bus Voltage 4160v x Generator Voltage 4150v Upon closure of the Diesel Generator output breaker, the operator must immediately raise ____(1)____ to avoid a ____(2)____ trip. A. (1) speed (2) over current B. (1) speed (2) reverse power C. (1) voltage (2) over current D. (1) voltage (2) reverse power Answer: B Reference Id: Q43968 Difficulty: 3.00 Time to complete: 2 10CFR Category: CFR 55.41 (7) 55.41 (7) Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features. Cognitive Level: Comprehension / Anal Question Source: New Comment: Proposed reference to be provided to applicant during examination: NONE Technical

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LOIT Lesson Plan, 40OP-9DG01(Emergency Diesel Generator) K&A: Maintaining minimum load on ED/G (to prevent reverse power) Learning Objective: Manually start, load, and unload the 'A' Diesel Generator Justification: A. Incorrect: Going to raise on the speed controller with the generator output breaker closed will raise load and is directed by procedure however, this will also raise output current. B. Correct: Going to raise on the speed controller with the generator output breaker closed will raise load and is directed by procedure. The basis for this step is to avoid a reverse power trip. C. Incorrect: Raising voltage setpoint will change reactive loading however, under the conditions stated an overcurrent condition will not be approached. D. Incorrect: Raising voltage setpoint will change reactive loading however, raising voltage will not mitigate a reverse power condition. REV 0

Palo Verde Operating Experience ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Reactor Operator NRC Exam 51. This Exam Level RO Appears on: RO EXAM 2012 Tier 2 Group 1 K/A # 2.2.39 Importance Rating: 3.9 Which ONE of the following pair of inoperable components would require entry into a ONE hour or less LCO condition while in Mode 1, steady state conditions? A. HPSI "A" and LPSI "B". B. AFW Pumps "A" and "B". C. Control Room Ventilation Intake Monitors RU-29 and 30. D. Both Atmospheric Dump Valves on Steam Generator #1. Answer: C Reference Id: Q43960 Difficulty: 3.00 Time to complete: 3 mins 10CFR Category: CFR 55.41 55.41 (10) Administrative, normal, abnormal, and (10) emergency operating procedures for the facility. Cognitive Level: Memory Question Source: New Comment: Proposed reference to be provided to applicant during examination: NONE Technical

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Tech Specs OPERATING EXPERIENCE QUESTION K&A: Knowledge of less than or equal to one hour Technical Specification action statements for systems. Justification: A. Incorrect - This is a 72 hour action per TS 3.5.3 B. Incorrect - This is a 6 hour action per TS 3.7.5. C. Correct - This is a 1 hour action per TS 3.3.9. D. Incorrect - This is a 24 hour action per TS 3.7.4. REV 0

ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Reactor Operator NRC Exam 5. This Exam Level: RO Appears on: RO EXAM 2012 Tier 2 Group 1 K/A #: 3.4 076 A2.01 Importance Rating: 3.5 Given the following conditions: x Unit 1 is operating at 100% power. x The Plant Cooling Water System develops a large unisolable leak in the common pump discharge header. x Plant Cooling Water Header Pressure Low Alarm Annunciates in the Control Room. x Essential Cooling Water train "A" is crosstied and supplying Nuclear Cooling Water priority loads. x 40AO-9ZZ03 Loss Of Cooling Water has been entered. Which ONE of the following systems are affected and what actions should the crew take? A. Turbine Cooling Water System, Trip the Reactor. B. Essential Cooling Water System, Trip the Reactor. C. Turbine Cooling Water System, Trip the Main Turbine. D. Essential Cooling Water System, Trip the Main Turbine. Answer: A Reference Id: Q43961 Difficulty: 3.00 Time to complete: 2 10CFR Category: CFR 55.41 55.41 (10) Administrative, normal, abnormal, and (10) emergency operating procedures for the facility. Cognitive Level: Comprehension / Anal Question Source: New Comment: Proposed reference to be provided to applicant during examination: NONE Technical

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40AO-9ZZ03, Loss of Cooling Water K&A: Ability to (a) predict the impacts of the following malfunctions or operations on the SWS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Loss of SWS. Learning Objective: Given plant conditions determine if the Loss of Cooling Water AOP should be executed in accordance with 40AO-9ZZ03. Justification: A. Correct: Plant Cooling Water System cools the Turbine Cooling Water Heat Exchanger and, 40AO-9ZZ03 Loss of Cooling Water requires a Reactor Trip. B. Incorrect: The loss of Essential Cooling Water System in this case would require a Reactor Trip however, the loss of Plant Cooling Water will not affect Essential Cooling Water. C. Incorrect: It is true that the Plant Cooling Water System cools the Turbine Cooling Water Heat Exchanger however, 40AO-9ZZ03, Loss of Cooling Water requires a Reactor Trip. D. Incorrect: The loss of Essential Cooling Water System in this case would require a Reactor Trip however, the loss of Plant Cooling Water will not affect Essential Cooling Water. 40AO-9ZZ03, Loss of Cooling Water requires a Reactor Trip. REV 0

ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Reactor Operator NRC Exam 53. This Exam Level: RO Appears on: RO EXAM 2012 Tier 2 Group 1 K/A #: 3.4 076 K1.19 Importance Rating: 3.6 Given the following conditions: x Unit 1 is operating at 100% power. x Both Nuclear Cooling Water Pumps are unavailable. x Essential Cooling Water (EW) is cross tied to supply Nuclear Cooling Water (NC). Which ONE of the following describes the NC priority heat load that will be supplied from EW? A. Normal Chillers. B. Letdown heat exchanger. C. Waste Gas Compressors. D. Containment Normal AHUs. Answer: A Reference Id: Q43965 Difficulty: 2.00 Time to complete: 2 10CFR Category: CFR 55.41 (4) 55.41 (4) Secondary coolant and auxiliary systems that affect the facility. Cognitive Level: Memory Question Source: New Comment: Proposed reference to be provided to applicant during examination: NONE Technical

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LOIT Lesson Plan K&A: Knowledge of the physical connections and/or cause- effect relationships between the SWS and the following systems: SWS emergency heat loads. Learning Objective: L65468 Describe the Nuclear Cooling Water Priority loads that can be supplied by the Essential Cooling Water system. Justification: A. Correct: Normal Chillers are a Priority Heat Load. B. Incorrect: Waste Gas compressors are not a Priority Heat Load. C. Incorrect: Letdown heat exchanger are not a Priority Heat Load. D. Incorrect: Containment Normal AHUs are not a Priority Heat Load. REV 0

ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Reactor Operator NRC Exam 54. This Exam Level: RO Appears on: RO EXAM 2007 RO EXAM 2012 Tier 2 Group 1 K/A #: 3.8 078 K3.02 Importance Rating: 3.6 Which ONE of the following is true regarding an Instrument Air pipe rupture in the Main Steam Support Structure (MSSS)? A. Service Air will supply all loads B. Accumulator will provide ADV operation C. Low Pressure Nitrogen will supply all loads D. Economizer Feedwater Isolation valves fast closure and slow mode of operation are available via the accumulator Answer: B Associated KA: L56751 Determine the major effects on plant operation as instrument air pressure degrades. Reference Id: Q44003 Difficulty: 3.00 Time to complete: 3 10CFR Category: CFR 55.41 (7) 55.41 (7) Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features. Cognitive Level: Comprehension / Anal Question Source: New Comment: Proposed reference to be provided to applicant during examination: NONE Technical

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40AO-9ZZ06 (Loss of Instrument Air) K&A: Knowledge of the effect that a loss or malfunction of the IAS will have on the following: Systems having pneumatic valves and controls. Learning Objective: Determine the major effects on plant operation as instrument air pressure degrades. Justification: A. Incorrect:The break will prevent backup sources supplying loads, Service Air no longer is a backup. B. Correct: Accumulator will allow ADV operation for up to 8 hours. C. Incorrect: Nitrogen backup may open on low pressure but the pipe break makes this useless. D. Incorrect: Accumulator provides fast closure but not slowmode of operation. REV 0

ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Reactor Operator NRC Exam 55. This Exam Level RO Appears on: RO EXAM 2007 RO EXAM 2012 Tier 2 Group 1 K/A # 3.5 103 A1.01 Importance 3.7 Rating: Given the following conditions: x Unit 1 has tripped due to a LOCA inside Containment. x SIAS/CIAS/MSIS/CSAS have initiated. x Both Containment Spray trains have failed to actuate. x The CRS has entered the Functional Recovery procedure. x CTPC-2 is being implemented to supply CS flow using LPSI pump A. Which ONE of the below listed sets of parameters will be monitored to satisfy CPTC-2? Containment... A. humidity and CS flow. B. pressure and CS flow. C. humidity and LPSI pump amps D. pressure and LPSI pump amps. Answer: D Reference Id: Q43989 Difficulty: 3.00 Time to complete: 2 10CFR Category: CFR 55.41 (10) 55.41 (10) Administrative, normal, abnormal, and emergency operating procedures for the facility. Cognitive Level: Memory Question Source: PV Bank Not Modified Comment: Proposed reference to be provided to applicant during examination: NONE Technical

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40EP-9EO09, CTPC-2 K&A: Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the containment system: Containment pressure, temperature, and humidity. Learning Objective: L65087 Describe the design basis associated with the Containment system. Justification: A. Incorrect: When LPSI is cross tied to CS, CS header flow is not available. (40EP-9EO09, CTPC-2, note by step 3). B. Incorrect: When LPSI is cross tied to CS, CS header flow is not available. (40EP-9EO09, CTPC-2, note by step 3). C. Incorrect: Humidity will be high initially from the LOCA, so a change would not be seen. D. Correct: 40EP-9EO09, CTPC-2 step 3.1.f limits amps to ensure continued operation of the LPSI pump. Containment pressure will drop if the section is performed correctly. REV 0

ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Reactor Operator NRC Exam 56. This Exam Level: RO Appears on: RO EXAM 2008 RO EXAM 2012 Tier 2 Group 2 K/A #: 3.1 001 A4.03 Importance Rating: 4.0 Given the following conditions: x Unit 3 is operating at 55% power following a Large Load Reject event. x The CRS has implemented 40AO-9ZZ08 (Load Rejection). x CEDMCS has been placed in standby. x Reg. Group 3 CEAs are at 135 inches withdrawn. x Reg. Group 4 CEAs are fully inserted. Proper CEA group overlap will be restored by ... A. withdrawing Reg group 4 CEAs in manual group mode. B. withdrawing Reg group 4 CEAs in manual sequential mode. C. withdrawing Reg. group 4 CEAs in manual individual mode while maintaining CEAs within 6.6 inches. D. lowering the load limit pot until the "Load Limiting" light illuminates then allow the Reg group 4 CEAs to withdraw in auto sequential mode. Answer: A Reference Id: Q22484 Difficulty: 3.00 Time to complete: 3 10CFR Category: CFR 55.41 (7) 55.41 (7) Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features. Cognitive Level: Memory Question Source: PV Bank Not Modified Comment: Proposed reference to be provided to applicant during examination: NONE Technical

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40AO-9ZZ08 (Large Load reject), 40OP-9SF01 (CEDMCS operations) K&A: Ability to manually operate and/or monitor in the control room: CRDS mode control Learning Objective: L78790 Describe the CEDMCS Remote Operator Module located in the Control Room to include all switches and the meaning of each switch position. Justification: A. Correct: RPCB LLR procedure directs withdraw in manual group. B. Incorrect: Manual Sequential would cause group 3 to withdraw to UGS while moving group 4 C. Incorrect: this would work but not directed by procedure, 6.6 inches is the CWP/CEDMCS Alarm limit. D. Incorrect: Lowering the pot is procedurally directed but to clear the RPCB signal not to withdraw CEAs. REV 0

ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Reactor Operator NRC Exam 57. This Exam Level: RO Appears on: RO EXAM 2012 Tier 2 Group 2 K/A #: 3.2 002 K3.03 Importance Rating: 4.2 Given the following conditions: x Unit 1 has tripped due to a Large Break LOCA. Which ONE of the following describes when the operating crew will consider the CTMT to be HARSH? CTMT Temperature >____(1)____ 0F OR CTMT Radiation level > ____(2)____ mR/hr. A. (1) 170 (2) 105 B. (1) 170 (2) 108 C. (1) 235 (2) 105 D. (1) 235 (2) 108 Answer: B Reference Id: Q43966 Difficulty: 2.00 Time to complete: 2 10CFR Category: CFR 55.41 55.41 (10) Administrative, normal, abnormal, and (10) emergency operating procedures for the facility. Cognitive Level: Memory Question Source: New Comment: Proposed reference to be provided to applicant during examination: NONE Technical

Reference:

40EP-9EO03 (LOCA), 40DP-9AP08 (Tech Guide) K&A: Knowledge of the effect that a loss or malfunction of the RCS will have on the following: Containment Learning Objective: Given conditions of LOCA analyze Containment Temperature and Pressure Control to determine if the SFSC acceptance criteria is satisfied in accordance with 40EP-9EO03. Justification: A. Incorrect: 170 0F is correct but 10 5 is the Rem value, the procedure specifically state mR/hr. B. Correct: 170 0F is correct and 10 8 is correct. C. Incorrect: 235 0F is the temperature that the CSAS pressure corresponds to. 10 5 is the Rem value, the procedure specifically state mR/hr. D. Incorrect: 235 0F is the temperature that the CSAS pressure corresponds to. 10 8 is correct. REV 0

ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Reactor Operator NRC Exam 58. This Exam Level: RO Appears on: RO EXAM 2008 RO EXAM 2012 Tier 2 Group 2 K/A #: 3.7 016 A3.01 Importance Rating: 2.9 Given the following conditions: x Unit 1 is operating at 100% power. x SG #1 level transmitter LT-1111 is within the normal band. x SG #1 level transmitter LT-1112 is within the normal band. Which ONE of the following describes the level transmitter signal(s)? SG #1 DFWCS automatically uses the... A. lower output of LT-1111 and LT1112. B. higher output of LT-1111 and LT-1112. C. average output of LT-1111 and LT-1112. D. output of LT-1111, unless it is out of range then LT-1112 will be selected. Answer: B Reference Id: Q43967 Difficulty: 3.00 Time to complete: 2 10CFR Category: CFR 55.41 (7) 55.41 (7) Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features. Cognitive Level: Memory Question Source: PV Bank Not Modified Comment: Proposed reference to be provided to applicant during examination: NONE Technical

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LOIT lesson plan K&A: Ability to monitor automatic operation of the NNIS, including: Automatic selection of NNIS inputs to control systems Learning Objective: L82151 Describe the NR steam generator level inputs to DFWCS and their function. Justification: A. Incorrect: DFWCS uses the higher output, candidate may think that the system uses the lower. B. Correct: DFWCS uses the higher output. C. Incorrect: DFWCS uses the higher output, candidate may think that the system uses the average. D. Incorrect: This would be true if LT-1111 is placed in maintenance. REV 0

ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Reactor Operator NRC Exam 59. This Exam Level: RO Appears on: RO EXAM 2010 RO EXAM 2012 Tier 2 Group 2 K/A #: 3.7 017 K1.01 Importance Rating: 3.2 Core Exit Thermocouples (CETs) provide a DIRECT input to which ONE of the following? A. COLSS. B. QSPDS. C. ERFDADS. D. B02 Post Accident Meters. Answer: B Reference Id: Q43753 Difficulty: 3.00 Time to complete: 3 10CFR Category: CFR 55.41 (7) 55.41 (7) Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features. Cognitive Level: Memory Question Source: New Comment: Proposed reference to be provided to applicant during examination: NONE Technical

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LOIT Lesson Plan K&A: Knowledge of the physical connections and/or cause effect relationships between the ITM system and the following systems: Plant computer. Learning Objective: L77368 Explain the operation of the Core Exit Thermocouples (CETs) associated with the Incore Instrumentation System. Justification: A. Correct: CET detectors are connected to the QSPDS cabinet by a chromel aluminum lead which removes the need for a temperature controlled environment junction box. B. Incorrect: COLSS receives inputs from the Incore detectors which are on the same instrument string as the CETs. C. Incorrect: ERFDADS receives CET data from QSPDS. D. Incorrect: B02 Post Accident Monitors receive data from QSPDS to display Core Exit Temps and Saturation Margins. REV 0

ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Reactor Operator NRC Exam 60. This Exam Level RO Appears on: RO EXAM 2012 Tier 2 Group 2 K/A # 3.5 028 A1.02 Importance Rating: 3.4 Given the following conditions: x Unit 2 has experienced a small LOCA resulting in a containment pressure of 2 psig. x PZR pressure is steady at 2100 psia. x The Hydrogen Recombiners are in operation. x Containment hydrogen concentration is 3.5%. x The break suddenly propagates resulting in dropping PZR pressure and containment pressure rising to 7 psig. Which ONE of the following describes the impact on the Hydrogen Recombiners? The Hydrogen Recombiners... A. will still be aligned. B. must be isolated to prevent exceeding its design pressure. C. must be isolated to prevent exceeding its design hydrogen concentration. D. have isolated and can be realigned from the control room using its override feature. Answer: D Reference Id: Q44009 Difficulty: 3.00 Time to complete: 3 10CFR Category: CFR 55.41 (7) 55.41 (7) Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features. Cognitive Level: Comprehension / Anal Question Source: New Comment: Proposed reference to be provided to applicant during examination: NONE Technical

Reference:

System Technical Manual, LOCA Procedure Technical Guide K&A: Ability to predict and/or monitor changes in parameter (to prevent exceeding design limits) associated with operating the HRPS controls including: Containment pressure. Learning Objective: Describe the automatic functions associated with the Hydrogen Control System Containment Isolation Valves. REV. 0

ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Reactor Operator NRC Exam Justification: A. Incorrect: The containment isolation valves for the hydrogen control system close on a Containment Isolation Signal actuated at 3.0 psig. B. Incorrect: The Hydrogen Recombiners can withstand maximum design containment pressure. In the LOCA procedure there is a limit imposed to ensure containment pressure is less than < 8.5 psig before aligning the hydrogen recombiners. The Hydrogen Control operating procedure has a maximum containment pressure of 10 psig. C. Incorrect: There is a hydrogen concentration lower limit of operation for the PURGE Units of at least 2.8%. The hydrogen control procedure does not have an upper limit on hydrogen concentration however, there is a caution to assume an explosive mixture is present when placing the hydrogen control system in operation. D. Correct: The containment isolation valves for the hydrogen control system close on a Containment Isolation Signal actuated at 3.0 psig and will be overriden and opened to re-establish hydrogen control. REV. 0

ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Reactor Operator NRC Exam 61. This Exam Level: RO Appears on: RO EXAM 2012 Tier 2 Group 2 K/A #: 3.8 029 A1.03 Importance Rating: 3.0 Which ONE of the following describes the interlock associated with Power Access Purge Containment Inlet Isolation valves. Containment ____(1)____ must be ____(2)____ the setpoint before the dampers will OPEN. A. (1) pressure (2) above B. (1) pressure (2) below C. (1) temperature (2) above D. (1) temperature (2) below Answer: B Reference Id: Q43969 Difficulty: 2.00 Time to complete: 2 10CFR Category: CFR 55.41 (7) 55.41 (7) Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features. Cognitive Level: Memory Question Source: New Comment: Proposed reference to be provided to applicant during examination: NONE Technical

Reference:

LOIT Lesson Plan K&A: Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the Containment Purge System controls including: Containment pressure, temperature, and humidity. Learning Objective: 75092 Describe the automatic functions and interlocks associated with the Power Access Purge Containment Isolation Dampers (CPA-UV-4A & 4B, and CPB-UV-5A & 5B). Justification: A. Incorrect: Candidate may confuse the inlet isolation valve with the vent valve CPN-PV-43 which has an interlock to remain closed so that flow will be directed through the vent orifice when pressure is above .5 psig B. Correct: The Power Access Purge Containment Inlet Isolation Valves are interlocked such that Containment Pressure must be below 0.03 psig as measured by HC-PT-493, before the dampers will open. C. Incorrect: Temperature provides interlocks to the CTMT Purge AHUs to determine if the Heaters or Chill Water will be used to adjust the temperature. CTMT Temperature is a Tech Spec monitored parameter. Candidate may confuse the inlet isolation valve with the vent valve CPN-PV-43 which has an interlock to remain closed so that flow will be directed through the vent orifice when pressure is above .5 psig D. Incorrect: Temperature provides interlocks to the CTMT Purge AHUs to determine if the Heaters or Chill Water will be used to adjust the temperature. CTMT Temperature is a Tech Spec monitored parameter. REV 0

ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Reactor Operator NRC Exam 62. This Exam Level: RO Appears on: RO EXAM 2012 Tier 2 Group 2 K/A #: 3.8 033 K4.01 Importance Rating: 2.7 Given the following conditions: x Unit 1 is at 100% power x Spent Fuel Pool (SFP) level is 137' 10" and has been noted by the AO to be slowly losing level over the past several shifts. x Chemistry has just reported SFP Boron Concentration at 1900 ppm. x The crew is investigating the loss of level at this time. x You are directed by the CRS to add water to the SFP. Which ONE of the following is the appropriate source of makeup water to the SFP? A. Recycle Monitor Tank. B. Refueling Water Tank. C. Condensate Storage Tank. D. Reactor Makeup Water Tank. Answer: B Reference Id: Q43970 Difficulty: 2.00 Time to complete: 2 10CFR Category: CFR 55.41 55.41 (10) Administrative, normal, abnormal, and (10) emergency operating procedures for the facility. Cognitive Level: Comprehension / Anal Question Source: New Comment: Proposed reference to be provided to applicant during examination: NONE Technical

Reference:

40AO-9ZZ23 (Loss of SFP Level), Tech Spec 3.714 & 3.7.15 K&A: Knowledge of design feature(s) and/or interlock(s) which provide for the following: Maintenance of spent fuel level. Learning Objective: Explain the operation of the Spent Fuel Pool under normal operating conditions. Justification: Tech Spec 3.7.15 states that SFP Boron Concentration must be > 2150 ppm. Therefore a Borated source must be used for make up. Normal losses from the SFP are from evaporation, therefore the normal makeup is a NON Borated Source. Candidate must know the Tech Spec Limit and that the loss is due to a leak which is not evaporation. These conditions require a Borated Makeup. A. Incorrect: RMT is a source of make up to the SFP, but it is NOT Borated. B. Correct: RWT is borated to >4400 ppm and is the correct source. C. Incorrect: CST is the normal source of make up for losses due to evaporation. It is NOT borated. D. Incorrect: RMWT is an available makeup source to the SFP, but it is NOT Borated. REV 0

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ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Reactor Operator NRC Exam 64. This Exam Level: RO Appears on: RO EXAM 2012 Tier 2 Group 2 K/A #: 3.7 072 K5.01 Importance Rating: 2.7 Given the following conditions: x Unit 1 operating at 100% power. x The core is at 250 EFPD. x A containment purge is in progress. x The reactor trips with indications of a large break LOCA. x A CIAS fails to actuate. x Core damage is indicated. The Power Access Purge Area Monitors, SQA-RU-37 and SQB-RU-38 will sense rising _______ radiation levels. A. beta B. alpha C. gamma D. neutron Answer: C Reference Id: Q44013 Difficulty: 2.00 Time to complete: 2 10CFR Category: CFR 55.41 (11) 55.41 (11) Purpose and operation of radiation monitoring systems, including alarms and survey equipment. Cognitive Level: Comprehension / Anal Question Source: New Comment: Proposed reference to be provided to applicant during examination: NONE Technical

Reference:

STM K&A: Knowledge of the operational implications of the following concepts as they apply to the ARM system: Radiation theory, including sources, types, units, and effects Learning Objective: L66723 Given a Area Radiation Monitor number and name describe the purpose Justification: A. Incorrect: Beta radiation will not be able to penetrate the piping B. Incorrect: Alpha radiation will not be able to penetrate the piping C. Correct: The radiation levels sensed by this detector would be coming from inside the purge lines, gamma being the most penetrating. D. Incorrect: There would be no significant neutron radiation levels due to the trip, containment shielding, and detector design. REV 0

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Reference:

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ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Reactor Operator NRC Exam 66. This Exam Level: RO Appears on: RO EXAM 2012 Tier 3 K/A #: 2.0 2.1 2.1.26 Importance Rating: 3.4 Which ONE of the following is the lower oxygen concentration limit which establishes confined space entry requirements? A. 16.0% B. 19.5% C. 21.0% D. 23.5% Answer: B Reference Id: Q43977 Difficulty: 3.00 Time to complete: 2 10CFR Category: CFR 55.41 55.41 (10) Administrative, normal, abnormal, and (10) emergency operating procedures for the facility. Cognitive Level: Memory Question Source: New Comment: Proposed reference to be provided to applicant during examination: NONE Technical

Reference:

LOIT Lesson Plan K&A:.Conduct of Operations: Knowledge of non-nuclear safety procedures (e.g. rotating equipment, electrical, high temperature, high pressure, caustic, chlorine, oxygen and hydrogen). Learning Objective: L62991 From memory state the required oxygen levels in a confined space Justification: A. Incorrect: This is the lethal limit. B. Correct:An oxygen deficient atmosphere exists when the oxygen concentration is less than 19.5%. C. Incorrect:This is the normal concentration in air. D. Incorrect:This is the upper limit. REV 0

ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Reactor Operator NRC Exam 67. This Exam Level: RO Appears on: RO EXAM 2012 Tier 3 K/A #: 2.1.37 Importance Rating: 4.3 Which ONE of the following describes the control room personnel that MUST attend a reactivity brief for a normal shiftly dilution per ODP-1 (Operations Principles and Standards)? The CRS, RO... A. and CO. B. and SM. C. and STA D. SM, STA and CO. Answer: A Reference Id: Q43988 Difficulty: 2.00 Time to complete: 2 10CFR Category: CFR 55.41 55.41 (10) Administrative, normal, abnormal, and (10) emergency operating procedures for the facility. Cognitive Level: Memory Question Source: New Comment: Proposed reference to be provided to applicant during examination: NONE Technical

Reference:

ODP-1 (Operations Principles and Standards) K&A: Conduct of Operations: Knowledge of procedures, guidelines or limitations associated with Reactivity Management Learning Objective: ODP-1 Reactivity Management Justification: A. Correct: Per ODP-1 The CRS, RO and CO WIll attend the Reactivity Brief. The SM and STA(s) should attend but are not required per the ODP-1 guidance. B. Incorrect: SM should attend but is not required. C. Incorrect: STA should attend but is not required. D. Incorrect: CO is required to attend but the SM and STA are not. REV 0

ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Reactor Operator NRC Exam 68. This Exam Level: RO Appears on: RO EXAM 2012 Tier 3 K/A #: 2.1.29 Importance Rating: 4.1 Given the following conditions: x Unit 1 is operating at 100% power. x The operating crew is performing a lineup to Drain the Safety Injection Tank (SIT) 1A . x SIE-V463 (SIT Fill and Drain Line Containment Isolation Valve) is to be opened to support the evolution. x The CRS has verified this to be a normally locked closed Containment Isolation valve. Per guidance found in 40DP-9OP19 (Locked Valve, Breaker, and Component Tracking), this valve ... A. is prohibited from being operated while in Mode 1. B. may be opened provided the the four hour action for an inoperable containment penetration is entered when the valve is opened. C. may be opened provided an Operator is identified in the Control Room log with the responsibility to close the valve with in 1 (ONE) hour. D. may be opened provided a dedicated Operator is stationed at the valve who must be in constant communication with the Control Room. Answer: D Reference Id: Q5219 Difficulty: 4.00 Time to complete: 3 10CFR Category: CFR 55.41 55.41 (10) Administrative, normal, abnormal, and (10) emergency operating procedures for the facility. Cognitive Level: Memory Question Source: PV Bank Not Modified Comment: Proposed reference to be provided to applicant during examination: NONE Technical

Reference:

40DP-9OP19 (Locked Valve, Breaker, and Component Tracking) K&A: Knowledge of how to conduct system lineups, such as valves, breakers, switches, etc. Learning Objective: describe the administrative controls required when intermittently opening of locked closed manual containment isolation valves in accordance with 40DP-9OP19. REV 0

ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Reactor Operator NRC Exam Justification: A. Incorrect: Candidate may think that this is a containment penetration that can not be opened in Mode 1. This may be performed in Mode 1. B. Incorrect: Entering the 4 hour action of 3.6.3 is not required to entered. Also, this will not eliminate the need for a dedicated operator or 60 second operation. C. Incorrect: The designated operator will be identified in the control room log, but this does not meet the requirements to close the valve with in 60 seconds. Tech specs has many instances of one hour requirements. D. Correct: This is correct per 40DP-9OP19 REV 0

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ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Reactor Operator NRC Exam 72. This Exam Level: RO Appears on: RO EXAM 2012 Tier 3 K/A #: 2.3.5 Importance Rating: 2.9 Given the following conditions: x You are preparing to enter the RCA on an approved Radiological Exposure Permit (REP). x Electronic Personnel Dosimeter (EPD) dose alarm setting is 500 mrem. x Electronic Personnel Dosimeter (EPD) dose rate alarm setting is 1000 mrem/hr. x Assigned RP work area dose rate is 1000 mr/hr. Based on the conditions above, which ONE of the following describes the Alarm you will receive and when you would be required to exit the Radiological Control Area (RCA)? You must leave the RCA ____(1)____ due to an EPD ____(2)____ alarm. A. (1) immediately (2) Dose B. (1) immediately (2) Dose Rate C. (1) in 30 minutes (2) Dose D. (1) in 30 minutes (2) Dose Rate Answer: B Reference Id: Q43985 Difficulty: 2.00 Time to complete: 2 10CFR Category: CFR 55.41 (11) 55.41 (11) Purpose and operation of radiation monitoring systems, including alarms and survey equipment. Cognitive Level: Comprehension / Anal Question Source: Industry Bank Comment: Proposed reference to be provided to applicant during examination: NONE Technical

Reference:

Radworker Training Handout K&A: Ability to use radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc. Learning Objective: 67126 Explain the operation of the Field Units under normal operating conditions. Justification: A. Incorrect. A dose alarm would be received in 60 minutes. Dose = 500 mrem/1000 mr/hr. B. Correct. A dose rate alarm would be received immediately since the work area dose rate is 1000 mr/hr; which is equal to the rate alarm setting. You are required by the ALARA program to exit the RCA upon receiving an ED alarm. C. Incorrect. A dose alarm would be received in 30 minutes. Dose = 500 mrem/1000 mr/hr. D. Incorrect. A dose rate alarm would be received immediately since the work area dose rate is 1000 mr/hr; which is equal to the rate alarm setting. REV 0

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ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Reactor Operator NRC Exam 73. This Exam Level: RO Appears on: RO EXAM 2012 Tier 3 K/A #: 2.4.16 Importance Rating: 3.5 Given the following conditions: x Unit 1 is operating at 100% power. x An Abnormal Operating procedure has been implemented. x During performance of the AOP the Reactor Trips. The next AOP step directs the following:

a. GO TO the appropriate procedure for the current plant conditions.

Which statement below best describes the use of Abnormal Operating Procedures (AOPs) after the crew has entered the Emergency Operating procedures (EOPs)? A. Immediately exit the AOP being performed. B. No further AOP actions are permitted until after the SPTAs are completed. C. Continue through the AOP until a step is reached that directs exiting the procedure. D. Any AOP that has been started prior to a reactor trip must be performed through completion. Answer: A Reference Id: Q43786 Difficulty: 3.00 Time to complete: 2 10CFR Category: CFR 55.41 55.41 (10) Administrative, normal, abnormal, and (10) emergency operating procedures for the facility. Cognitive Level: Memory Question Source: Modified PV Bank Comment: Proposed reference to be provided to applicant during examination: NONE Technical

Reference:

40DP-9AP18 (AOP Users Guide) K&A: Knowledge of EOP implementation hierarchy and coordination with other support procedures or guidelines such as, operating procedures, abnormal operating procedures, and severe accident management guidelines. Learning Objective: L82065 Given indications for entry into an Abnormal Operating Procedure define the required actions for the conditions given in accordance with the applicable Abnormal Operating Procedure. Justification: A. Correct - This is true for a "GO TO" step in the AOPs As found in section 17 of the users guide. B. Incorrect - No actions are permitted until the Reactivity Safety Function is complete. C. Incorrect - Some AOPs must be completed concurrently such as the "PERFORM" direction. D. Incorrect - AOPs must be completed unless directed to exit. REV 0

ORIGINAL QUESTION 1 ID: Q8781 Points: 1.00

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ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Reactor Operator NRC Exam 75. This Exam Level: RO Appears on: RO EXAM 2008 RO EXAM 2012 Tier 3 K/A #: 2.4.5 Importance Rating: 3.7 A plant perturbation is in progress that if not properly addressed could result in a manual or automatic Unit trip. Which ONE of the following sets of procedures would be used to mitigate this event? A. Normal Operating Procedures. B. General Operating Procedures. C. Abnormal Operating Procedures. D. Emergency Operating Procedures. Answer: C Reference Id: Q22410 Difficulty: 2.00 Time to complete: 1 10CFR Category: CFR 55.41 55.41 (10) Administrative, normal, abnormal, and (10) emergency operating procedures for the facility. Cognitive Level: Memory Question Source: New Comment: Proposed reference to be provided to applicant during examination: None Technical

Reference:

AOP/EOP Users Guides K&A: Emergency Procedures / Plan Knowledge of the organization of the operating procedures network for normal, abnormal, and emergency evolutions. Learning Objective: Given that an ORP is being implemented describe the use of an AO or OP when the reactor trips or when performing an EOP Justification: A. Incorrect: Intended normal conditions not transients B. Incorrect: For general operations, not transients C. Correct: AOPs restore normal conditions following a transient D. Incorrect: Place the plant in a safe condition after a Reactor trip event REV 0

ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Senior Reactor Operator NRC Exam 1. This Exam Level: SRO Appears on: SRO EXAM 2012 Tier 1 Group 1 K/A #: 011 2.4.41 Importance 4.6 Rating: Given the following conditions: x Unit 3 has tripped from 100% power. x Containment hydrogen concentration per HPA-AI-9 indicates 3.8%. x Containment hydrogen concentration per HPB-AI-10 indicates 4.2%. x Estimated reactor coolant system leakage is 500 gpm. x Highest Rep CET reading is 587°F. x RCS chemistry sample dose equivalent Iodine 131 indicates 308 uCi/gm. x Containment pressure - 37 psig and slowly lowering. x Pressurizer pressure - 610 psia. x RVLMS - upper head level - 16%. x All equipment has properly actuated. Which ONE of the following describes the appropriate classification and code for this event? A. Unusual Event - FU1 B. Alert - FA1 C. Site Area Emergency - FS1 D. General Emergency - FG1 Answer: C Reference Id: Q43902 Difficulty: 3.00 Time to complete: 3 10CFR Category: CFR 55.43 (5) 55.43 (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. Cognitive Level: Comprehension / Anal Question Source: New Comment: Proposed reference to be provided to applicant during examination: NEI 99-01 HOT/COLD EAL CHART Technical

Reference:

NEI99-01 HOT EAL CHART K&A: Knowledge of the emergency action level thresholds and classifications. Learning Objective: L58622 Given an Emergency Plan condition, use the EAL tables and basis document to determine the emergency plan classification REV 0

ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Senior Reactor Operator NRC Exam Justification: A. Incorrect: NUE is met but it is not the highest EAL classification of the event. Candidate may confuse any of the indications and not properly apply them to the EAL chart. B. Incorrect: Alert is met but it is not the highest EAL classification of the event. Candidate may confuse any of the indications and not properly apply them to the EAL chart. C. Correct: SAE is met and is the highest EAL classification of the event. D. Incorrect: GE is not met. Candidate may confuse any of the indications and not properly apply them to the EAL chart. REV 0

RADIOLOGICAL SYSTEM MALFUNCTIONS HAZARDS FISSION PRODUCT BARRIERS EFFLUENTS RX and CORE AC/DC POWER ALARMS / COMMUNICATIONS NATURAL / DESTRUCTIVE FIRE / EXPLOSION TOXIC / FLAMMABLE SECURITY CR EVACUATION EC DISCRETION RG1 - Off-site dose resulting from an actual or IMMINENT release of gaseous MG2 - Automatic Trip and all manual actions HG1 - HOSTILE ACTION resulting in HG2 - Other conditions exist which in the radioactivity greater than 1000 mrem TEDE or 5000 mrem Thyroid CDE for the actual fail to shutdown the reactor and indication of MG1 - Prolonged loss of all Off-site and all loss of physical control of the facility. judgment of the EC warrant declaration of a or projected duration of the release using actual meteorology. POTENTIAL POTENTIAL POTENTIAL an extreme challenge to the ability to cool the On-Site AC power to emergency busses. General Emergency. LOSS LOSS LOSS core exists. Modes 1 & 2 LOSS LOSS LOSS Note: The EC should not wait until the applicable time has elapsed, but should declare the 1. A HOSTILE ACTION has occurred such FUEL CLAD RCS CONTAINMENT 1. Other conditions exist which in the event as soon as it is determined that the condition will likely exceed the applicable time. If that plant personnel are unable, either 1.a. Plant Protection System failed to 1.a. Loss of all off-site and all on-site judgment of the EC indicate that events are dose assessment results are available, declaration should be based on dose assessment instead remotely or locally, to operate equipment shutdown the reactor. AC power to PBA-S03 and PBB-S04. in progress or have occurred which involve of radiation monitor values. Do not delay declaration awaiting dose assessment results. required to maintain safety functions. AND OR actual or IMMINENT substantial core AND

b. All Manual actions do NOT shutdown 2. A HOSTILE ACTION has caused failure of degradation or melting with potential for

1. VALID reading on ANY of the following radiation monitors greater than the value for 15 the reactor as indicated by: b. EITHER of the following: loss of containment integrity or HOSTILE Spent Fuel Cooling Systems and minutes or longer: Reactor power is NOT dropping to ACTION that results in an actual loss of 3/3 less than 5% power Restoration of at least one emergency bus IMMINENT fuel damage is likely for a Plant Vent RU-144 CH-1 >1.04E+00 uCi/cc in less than 4 hours is not likely. physical control of the facility. Releases can All full strength CEAs are NOT freshly off-loaded reactor core in pool.

Fuel Building RU-146 CH-2 >3.50E+01 uCi/cc Loss of at least 2 FG1 - Loss of ANY Two Barriers AND Loss or Potential inserted RCS and Core Heat Removal Safety be reasonably expected to exceed EPA OR -- YES -- Protective Action Guideline exposure levels Barriers? Loss of the Third Barrier AND Function Acceptance Criteria NOT

2. Dose assessment using actual meteorology indicates doses greater than 1000 mrem TEDE c. Rep CET greater than 1200 oF. Satisfied per 40EP-9EO08, BLACKOUT. off-site for more than the immediate site GENERAL EMERGENCY GENERAL EMERGENCY OR 5000 mrem thyroid CDE at or beyond the site boundary. area.

OR

3. Field survey results indicate closed window dose rates greater than 1000 mR/hr expected to continue for 60 minutes or longer; or analyses of field survey samples indicate thyroid CDE greater than 5000 mrem for one hour of inhalation, at or beyond site boundary.

RS1 - Off-site dose resulting from an actual or IMMINENT release of gaseous radioactivity POTENTIAL POTENTIAL POTENTIAL MS2 - Automatic Trip fails to shutdown the HS2 - Control room evacuation has been HS3 - Other conditions exist which in the LOSS LOSS LOSS MS1 - Loss of all Off-site and all On-Site AC HS4 - HOSTILE ACTION within the greater than 100 mrem TEDE or 500 mrem Thyroid CDE for the actual or projected duration of LOSS LOSS LOSS reactor and manual actions taken at the reactor MS6 - Inability to monitor a significant initiated and plant control cannot be judgment of the EC warrant declaration of a power to emergency busses for 15 minutes or PROTECTED AREA. the release. control console are not successful in shutting transient in Progress. established. Site Area Emergency. FUEL CLAD RCS CONTAINMENT longer. down the reactor. 1.a. Control Room evacuation has been Note: The EC should not wait until the applicable time has elapsed, but should declare the Modes 1 & 2 Note: The EC should not wait until the Note: The EC should not wait until the initiated. event as soon as it is determined that the condition will likely exceed the applicable time. If 1. A HOSTILE ACTION is occurring or has 1. Other conditions exist which in the applicable time has elapsed, but should applicable time has elapsed, but should occurred within the PROTECTED AREA as AND dose assessment results are available, declaration should be based on dose assessment instead judgment of the EC indicate that events are declare the event as soon as it is determined declare the event as soon as it is determined reported by the Security Team. b. Control of the plant cannot be established of radiation monitor values. Do not delay declaration awaiting dose assessment results. 1.a. Plant Protection System failed to in progress or have occurred which involve that the condition has exceeded, or will likely that the condition has exceeded, or will likely at the Remote Shutdown Panel

                                                                                                                                                                                                                       -- NO --                                                                                                         shutdown the reactor.                                                                                                                                                                                                                                                                                                                                                                                  actual or likely major failures of plant exceed, the applicable time                       exceed, the applicable time                                                                                                                                                                                                                           within 15 minutes.

AND functions needed for protection of the public 2/3

b. Manual actions taken on Panels B05 and 1. a. Loss of annunciators on ANY 4 of the or HOSTILE ACTION that results in

1. VALID reading on ANY of the following radiation monitors greater than the value for 15 FS1 - Loss or Potential Loss of ANY Two Barriers B01 do NOT shut down the reactor as 1. Loss of all Off-Site and all On-Site AC following B01, B02, B04, B05, B06 intentional damage or malicious acts; (1) minutes or longer: indicated by: power to PBA-S03 and PBB-S04 or toward site personnel or equipment that could Plant Vent RU-144 CH-1 >1.04E-01 uCi/cc Reactor power is NOT dropping to for 15 minutes or longer. SESS for 15 minutes or longer.

OR lead to the likely failure of or; (2) that prevent Fuel Building RU-146 CH-1 >3.50E+00 uCi/cc less than 5% power Loss of either PNA-D25 or PNB-D26 effective access to equipment needed for the OR POTENTIAL POTENTIAL All full strength CEAs are NOT inserted LOSS LOSS MS3 - Loss of all Vital DC Power for 15 for 15 minutes or longer. protection of the public. Any releases are not

2. Dose assessment using actual meteorology indicates doses greater than 100 mrem TEDE OR LOSS LOSS minutes or longer. AND expected to result in exposure levels which 500 mrem thyroid CDE at or beyond the site boundary.

exceed EPA Protective Action Guideline OR FUEL CLAD RCS b. ANY of the following: Note: The EC should not wait until the exposure levels beyond the site boundary. SITE AREA EMERGENCY

3. Field survey results indicate closed window dose rates greater than 100 mR/hr expected to SITE AREA EMERGENCY applicable time has elapsed, but should Automatic turbine setback/runback continue for 60 minutes or longer; or analyses of field survey samples indicate thyroid CDE greater than 25% thermal reactor declare the event as soon as it is determined greater than 500 mrem for one hour of inhalation, at or beyond the site boundary power that the condition has exceeded, or will likely exceed, the applicable time. Reactor Trip VALID ESFAS Actuation

1. Less than 112 VDC on all PKA-M41, AND 1/2 PKB-M42, PKC-M43, and PKD-M44 for 15 minutes or longer. c. Plant computer indications are unavailable.

FA1 - ANY Loss or ANY Potential Loss of EITHER Fuel Clad OR RCS MA2 - Automatic Trip fails to shutdown the MA4 - UNPLANNED Loss of safety system HA2 - FIRE or EXPLOSION affecting the HA3 - Access to a VITAL AREA is HA6 - Other conditions exist which in the EFFLUENTS RAD LEVELS MA5 - AC power capability to emergency HA1 - Natural or destructive phenomena HA4 - HOSTILE ACTION within the Owner HA5 - Control Room evacuation has been reactor and the manual actions taken from the annunciation or indication in the Control operability of plant safety systems required to prohibited due to release of toxic, corrosive, judgment of the EC warrant declaration of an POTENTIAL busses reduced to a single power source for 15 affecting VITAL AREAS. Controlled Area or airborne attack threat initiated. LOSS reactor control console are successful in Room with EITHER (1) a significant transient establish or maintain safe shutdown. asphyxiant or flammable gases which Alert. LOSS minutes or longer such that ANY additional RA2 - Damage to irradiated fuel or loss of shutting down the reactor. Modes 1 and 2 in progress, or (2) compensatory indicators are jeopardize operation of systems required to 1. Control Room evacuation is required by: 1. Other conditions exist which in the RA1 - ANY release of gaseous radioactivity single failure would result in station blackout. maintain safe operations or safely shutdown water level that has resulted or will result in CONTAINMENT unavailable. 1. A HOSTILE ACTION is occurring or has 40AO-9ZZ18, Shutdown Outside Control judgment of the EC indicate that events are to the environment greater than 20 times the the uncovering of irradiated fuel outside the 1.a. Plant Protection System failed to 1.a. Seismic event greater than 1. FIRE or EXPLOSION resulting in VISIBLE the reactor. occurred within the Owner Controlled Area in progress or have occurred which involve ODCM for 15 minutes or longer. Operating Basis Earthquake (OBE) DAMAGE to ANY POWER BLOCK Room reactor vessel. shutdown the reactor. Note: The EC should not wait until the Note: The EC should not wait until the as reported by the Security Team. an actual or potential substantial as indicated by ANY Force Balance structure or Control Room indication of Note: If the equipment in the stated area was OR AND applicable time has elapsed, but should applicable time has elapsed, but should degraded performance of safety systems. OR degradation of the level of safety of the Note: This EAL does not apply to the cask Accelerometer reading greater than 0.10g. already inoperable, or out of service, before 40AO-9ZZ19, Control Room Fire. Note: The EC should not wait until the b. Manual shutdown actions taken on declare the event as soon as it is determined declare the event as soon as it is determined AND 2. A validated notification from NRC of an plant or a security event that involves loading pit during cask loading operations. that the condition has exceeded, or will likely that the condition has exceeded, or will likely the event occurred, then this EAL should not applicable time has elapsed, but should 1/1 Panels B05 or B01 are successful as b. Earthquake confirmed by ANY of the airliner attack threat within 30 minutes of probable life threatening risk to site exceed, the applicable time. exceed, the applicable time. following: be declared as it will have no adverse impact declare the event as soon as it is determined 1. A water level drop in the reactor refueling indicated by all of the following: the site. personnel or damage to site equipment Earthquake felt in plant on the ability of the plant to safely operate or that the release duration has exceeded, or will cavity, spent fuel pool, cask loading pit, or Reactor Power is dropping to OR because of HOSTILE ACTION. Any FU1 - ANY Loss OR ANY Potential Loss of 1.a. AC power capability to 1. a. UNPLANNED Loss of annunciators on National Earthquake Center safely shutdown beyond that already allowed likely exceed, the applicable time. In the fuel transfer canal that will result in less than 5% power ANY 4 of the following 3. A HOSTILE ACTION directed toward the releases are expected to be limited to small Containment PBA-S03 and PBB-S04 reduced to a Control Room indication of degraded by Technical Specifications at the time of the absence of data to the contrary, assume that uncovering irradiated fuel. Negative Startup rate B01, B02, B04, B05, B06 or SESS ISFSI. fractions of the EPA Protective Action single power source for 15 minutes or performance of systems required for the event. the release duration has exceeded the OR All full strength CEAs are inserted for 15 minutes or longer Guideline exposure levels. or Boration in progress longer. OR safe shutdown of the plant. applicable time if an ongoing release is 2. A VALID High Alarm on ANY of the 1. Access to a VITAL AREA is prohibited due Note: Multiple events could occur which result in the conclusion that exceeding the loss or The Containment Barrier should not be declared lost or potentially lost based on AND UNPLANNED Loss of either OR detected and the release start time is unknown. following due to damage to irradiated fuel 2. Tornado touching down or high winds to toxic, corrosive, asphyxiant or flammable potential loss thresholds is IMMINENT. exceeding Technical Specification action statement criteria, unless there is an b. Any additional single power source PNA-D25 or PNB-D26 gases which jeopardize operation of systems ALERT or loss of water level: event in progress requiring mitigation by the Containment barrier. When no for 15 minutes or longer. reaching 100 mph resulting in ALERT In this IMMINENT loss situation use judgment and failure will result in station blackout. RU-16 Containment Operating Level Area event is in progress (Loss or Potential Loss of either Fuel Clad and/or RCS) the VISIBLE DAMAGE to ANY required to maintain safe operations or classify as if the thresholds are exceeded. AND POWER BLOCK structure OR safely shutdown the reactor.

1. VALID reading on ANY of the following RU-17 Incore Instrument Area Containment Barrier status is addressed by Technical Specifications.

RU-19 New Fuel Area b. ANY of the following: Control Room indication of degraded radiation monitors greater than the value performance of safety systems. for 15 minutes or longer: RU-31 Spent Fuel Pool Area Automatic turbine setback/runback OR RU-33 Refueling Machine Area Fuel Clad Barrier RCS Barrier Containment Barrier greater than 25% thermal reactor 3. Internal flooding in ANY POWER BLOCK Plant Vent RU-143 CH-1 > 1.22E-02 uCi/cc RU-143 Plant Vent power structure resulting in an electrical shock Fuel Bldg RU-146 CH-1 >1.13E-01 uCi/cc Reactor Trip hazard that precludes access to operate or RU-145 Fuel Building Vent Loss Potential Loss Loss Potential Loss Loss Potential Loss VALID ESFAS Actuation monitor safety equipment OR OR Control Room indication of degraded

2. Confirmed sample analyses for gaseous RA3 - Rise in radiation levels within the Plant computer unavailable

1. A. Coolant activity 1. A. RCS leak rate greater 1. A. RCS leak rate greater 1. A. A containment 1. A. Containment pressure performance of those safety systems.

releases indicates concentrations or release facility that impedes operation of systems greater than 300 Ci/gm than charging capacity pressure rise followed by greater than 60 psig OR than available makeup rates greater than 20 times the ODCM required to maintain plant safety functions. Dose Equivalent I-131. capacity as indicated by with Letdown isolated. a rapid unexplained drop and rising. 4. Vehicle crash resulting in Section 3.0 limits for 15 minutes or longer. a loss of RCS subcooling OR in containment pressure. OR VISIBLE DAMAGE to ANY to saturation (0 oF). OR B. 4.5% H2 inside POWER BLOCK structure OR

1. Dose rate greater than 15 mR/hr in the B. RCS Pressure Control Control Room indication of degraded Safety Function Status B. Containment pressure containment.

Control Room Area OR Secondary Alarm or sump level response OR performance of safety systems Station. Not Satisfied. not consistent with C. a. Pressure greater than OR LOCA or MSLB 8.5 psig. C. RCS and Core Heat HU2 - FIRE within the PROTECTED AREA HU5 - Other conditions exist which in the conditions. AND MU2 - Inability to reach required shutdown MU1 - Loss of all Off-site AC power to MU3 - UNPLANNED loss of safety system HU1 - Natural or destructive phenomena HU3 - Release of toxic, corrosive, asphyxiant, HU4 - Confirmed SECURITY CONDITION RU1 - ANY release of gaseous radioactivity RU2 - UNPLANNED rise in plant radiation Removal Safety Function b. Less than one full not extinguished within 15 minutes of judgment of the EC warrant declaration of a to the environment greater than 2 times the within Technical Specification limits. emergency busses for 15 minutes or longer. annunciation or indication in the Control affecting the PROTECTED AREA. or flammable gases deemed detrimental to or threat which indicates a potential ODCM for 60 minutes or longer. levels. Status Not Satisfied. train of Containment detection or EXPLOSION within the UE. Room for 15 minutes or longer. NORMAL PLANT OPERATIONS. degradation in the level of safety of the plant. Spray operating. PROTECTED AREA. Note: The EC should not wait until the 1. a. A VALID Alert Alarm on ANY of the 2. A. Rep CET reading 2. A. Rep CET reading 2. A. a. Rep CET greater Note: The EC should not wait until the 1. Seismic event identified by ANY 2 of the Note: The EC should not wait until the applicable time has elapsed, but should currently or previously currently or previously than 1200ºF. 1. Plant is not brought to required operating applicable time has elapsed, but should Note: The EC should not wait until the applicable time has elapsed, but should 1. Toxic, corrosive, asphyxiant or flammable 1. Other conditions exist which in the following: following: 1. A SECURITY CONDITION that does declare the event as soon as it is determined greater than 1200 oF greater than 700 oF AND mode within Technical Specifications declare the event as soon as it is determined applicable time has elapsed, but should declare the event as soon as it is determined gases in amounts that have or could NOT involve a HOSTILE ACTION as judgment of the EC indicate that events are

b. Restoration not VALID Seismic Event alarm that the release duration has exceeded, or will RU-16 Containment Operating Level Area LCO Action Statement Time. that the condition has exceeded, or will likely declare the event as soon as it is determined that the duration has exceeded, or will likely adversely affect NORMAL PLANT reported by the Security Team. in progress or have occurred which indicate effective within Earthquake felt in plant likely exceed, the applicable time. In the RU-17 Incore Instrument Area exceed, the applicable time. that the condition has exceeded, or will likely exceed the applicable time. OPERATIONS. a potential degradation of the level of safety 15 minutes. National Earthquake Center OR absence of data to the contrary, assume that RU-19 New Fuel Area OR exceed, the applicable time. OR OR of the plant or indicate a security threat to RU-31 Spent Fuel Pool Area 2. A credible PVNGS security threat facility protection has been initiated. No the release duration has exceeded the B. a. Rep CET greater than MU5 - RCS Leakage. 1. Loss of all off-site AC power to 2. Tornado touching down within the 1. FIRE in the POWER BLOCK or Turbine 2. Report by local, county or state officials for notification.

RU-33 Refueling Machine Area 700 oF. PBA-S03 and PBB-S04 1. UNPLANNED Loss of annunciators on Building not extinguished within 15 minutes evacuation or sheltering of site personnel releases of radioactive material requiring applicable time if an ongoing release is ANY 4 of the following PROTECTED AREA or high winds AND for 15 minutes or longer. of a FIRE alarm or Control Room based on an off-site event. OR off-site response or monitoring are expected detected and the release start time is unknown. b. RVLMS less than 21% B01, B02, B04, B05, B06 or SESS reaching 100 mph. AND OR notification. 3. A validated notification from NRC unless further degradation of safety systems

b. UNPLANNED water level drop in the plenum. 1. Unidentified or pressure boundary for 15 minutes or longer. providing information of an aircraft threat.

1. VALID reading on ANY of the following AND OR 3. Internal flooding in the POWER BLOCK OR occurs.

reactor refueling cavity, fuel transfer LEAKAGE greater than 10 gpm. radiation monitors greater than the value for c. Restoration not OR UNPLANNED Loss of either canal, cask loading pit, or spent fuel pool that has the potential to affect safety related 2. EXPLOSION within the 60 minutes or longer: as indicated by ANY of the following: effective within PNA-D25 or PNB-D26 equipment required by Technical

2. Identified LEAKAGE greater than 25 gpm. PROTECTED AREA.

Plant Vent RU-143 CH-1 >1.22E-03 uCi/cc 15 minutes. for 15 minutes or longer. Visual observation Specifications for the current operating Fuel Bldg RU-145 CH-1 >1.13E-02 uCi/cc SFP LEVEL HI - LOW (EO204A) 3. A. RUPTURED SG is mode.

3. A. RVLMS level 3. A. RUPTURED SG OR on PCN-E02 currently or previously results in an SIAS. also FAULTED outside MU4 - Fuel Clad Degradation. MU6 - Loss of all On-site or Off-site OR

2. Confirmed sample analyses for gaseous RWLIS less than 21% plenum. of containment. communications capabilities 4. Main Turbine failure resulting in casing releases indicates concentrations or release OR penetration or damage to turbine or Pressurizer level B.a. Primary-to-Secondary rates greater than 2 times the ODCM Section 3 OR 1. RU-155D High Alarm Main Generator seals.

leakrate greater than 1. Loss of all of the following on-site limits for 60 minutes or longer. 10 gpm. OR

2. UNPLANNED VALID Area Radiation 2.a. DOSE EQUIVALENT I-131 communication methods affecting the Monitor readings or survey results indicate AND greater than 1.0 Ci/gm for 48 hours. ability to perform routine operations.

a rise by a factor of 1000 over normal* UNUSUAL EVENT UNUSUAL EVENT

b. UNISOLABLE steam PBX levels. release from the OR affected SG to the Plant Page System environment. b. Coolant Gross Specific Activity Two-Way Radio

*Normal levels can be considered as the greater than 100/ Ci/gm. OR highest reading in the past twenty-four hours 4. A. a. Failure of all 2. Loss of all of the following off-site excluding the current peak value. valves in any one communication methods affecting the line to close MU8 - Inadvertent Criticality. Mode 3 or 4 AND ability to perform off-site notifications.

b. Direct downstream PBX pathway to the 1. UNPLANNED sustained source range FTS ISFSI environment exists Cellular Phones after containment count rise observed on nuclear isolation signal. instrumentation.

5. A. Containment radiation 5. A. Containment radiation 5. A. Containment radiation E-HU1 - Damage to a loaded cask monitor monitor monitor CONFINEMENT BOUNDARY RU-148 > 2.1E+05 mR/hr RU-148 > 5.0E+04 mR/hr RU-148 > 6.8E+06 mR/hr OR OR OR RU-149 > 2.4E+05 mR/hr RU-149 > 5.6E+04 mR/hr. RU-149 > 7.8E+06 mR/hr

1. Damage to a loaded cask CONFINEMENT

6. A. Any condition in the opinion of the EC that indicates 6. A. Any condition in the opinion of the EC that indicates 6. A. Any condition in the opinion of the EC that indicates BOUNDARY.

Loss or Potential Loss of the Fuel Clad Barrier. Loss or Potential Loss of the RCS Barrier. Loss or Potential Loss of the Containment Barrier. CONTAINMENT CLOSURE: The procedurally defined actions taken to secure containment and its associated IMMINENT: Mitigation actions have been ineffective, additional actions are not expected to be successful, and trended POWER BLOCK: Structures, systems or components listed below that contain equipment necessary for safe operation structures, systems, and components as a functional barrier to fission product release in Mode 6. information indicates that the event or condition will occur. and/or shutdown of the reactor. UNISOLABLE: A breach or leak that cannot be isolated from the Control Room. A. Containment CONFINEMENT BOUNDARY: The dry storage cask barriers between areas containing radioactive substances and the LEAKAGE shall be: B. Auxiliary Building environment. UNPLANNED: A parameter change or an event that is not the result of an intended evolution and requires corrective or

a. Identified LEAKAGE C. Refueling Water Tank (RWT) mitigative actions.

EXPLOSION: A rapid, violent, unconfined combustion, or catastrophic failure of pressurized/energized equipment that 1. LEAKAGE, such as that from pump seals or valve packing (except reactor coolant pump (RCP) seal water D. Diesel Generator Building imparts energy of sufficient force to potentially damage permanent structures, systems, or components. injection or leakoff), that is captured and conducted to collection systems or a sump or collecting tank; E. Diesel Generator Fuel Oil Storage Tanks FAULTED: in a steam generator, the existence of secondary side leakage that results in an uncontrolled drop in steam 2. LEAKAGE into the containment atmosphere from sources that are both specifically located and known either not F. Fuel Building generator pressure or the steam generator being completely depressurized to interfere with the operation of leakage detection systems or not to be pressure boundary LEAKAGE; or G. Spray Pond VALID: An indication, report, or condition, is considered to be VALID when it is verified by (1) an instrument channel

3. Reactor Coolant System (RCS) LEAKAGE through a steam generator (SG) to the Secondary System (primary to H. Condensate Storage Tank (CST) check, (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt FIRE: Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical secondary LEAKAGE). I. Control Building related to the indicators operability, the conditions existence, or the reports accuracy is removed.

equipment do not constitute FIRES. Observation of flame is preferred but is NOT required if large quantities of smoke and b. Unidentified LEAKAGE J. Corridor Building heat are observed. All LEAKAGE that is not identified LEAKAGE; K. MSSS Definitions HOSTILE ACTION: An act toward a NPP or its personnel that includes the use of violent force to destroy equipment, c. Pressure Boundary LEAKAGE Revision 0 take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, LEAKAGE (except primary to secondary LEAKAGE) through a nonisolable fault in an RCS component body, pipe PROTECTED AREA: The area which encompasses all controlled areas within the security PROTECTED AREA fence. VISIBLE DAMAGE: Damage to equipment or structure that is readily observable without measurements, testing, or 10/01/09 explosives, PROJECTILEs, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall wall, or vessel wall. analysis. Damage is sufficient to cause concern regarding the continued operability or reliability of the affected structure, RUPTURED: in a steam generator, existence of primary-to-secondary leakage of a magnitude sufficient to require or system, or component. Example damage includes: deformation due to heat or impact, denting, penetration, rupture, intent may be included. HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts cause a reactor trip and safety injection cracking, and paint blistering. Surface blemishes (e.g., paint chipping, scratches) should not be included. that are not part of a concerted attack on the NPP. Non-terrorism-based EALs should be used to address such activities (i.e., NORMAL PLANT OPERATIONS: Activities at the plant site, excluding the Water Reclamation Facility, associated this may include violent acts between individuals in the owner controlled area). with routine testing, maintenance, or equipment operations, in accordance with normal operating or administrative SECURITY CONDITION: Any Security Event as listed in the approved security contingency plan that constitutes a HOSTILE FORCE: One or more individuals who are engaged in a determined assault, overtly or by stealth and procedures. Entry into abnormal or emergency operating procedures, or deviation from normal security or radiological threat/compromise to site security, threat/risk to site personnel, or a potential degradation to the level of safety of the plant. controls posture, is a departure from NORMAL PLANT OPERATIONS. VITAL AREAS: Areas, within the PROTECTED AREA, that contains equipment vital to the operations of the plant. EP-0801 A deception, equipped with suitable weapons capable of killing, maiming, or causing destruction. A SECURITY CONDITION does not involve a HOSTILE ACTION.

ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Senior Reactor Operator NRC Exam 2. This Exam Level: SRO Appears on: SRO EXAM 2012 Tier 1 Group 1 K/A #: 025 2.4.6 Importance 4.7 Rating: Given the following conditions: Initial Conditions: x Unit 1 is in Mode 6. x Core off-load is in progress. x SDC is in service using LPSI pump "B". x "A" EW heat exchanger is tagged out for tube leak repair. Subsequently: x A large piece of tarp has lodged in the "B" train SDC suction piping. x LPSI pump "B" has been secured. The CRS should restore SDC flow by use of which ONE of the following? A. CS pump "A" with "A" train auxiliaries per Lower Mode Functional Recovery (40EP-9EO11). B. LPSI pump "A" with "B" train auxiliaries per Lower Mode Functional Recovery (40EP-9EO11). C. CS pump "A" with "A" train auxiliaries per Recovery from Shutdown Cooling to Normal Operating Lineup (40OP-9SI02). D. LPSI pump "A" with "B" train auxiliaries per Recovery from Shutdown Cooling to Normal Operating Lineup (40OP-9SI02). Answer: B Reference Id: Q43900 Difficulty: 3.00 Time to complete: 3 10CFR Category: CFR 55.43 (5) 55.43 (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. Cognitive Level: Comprehension / Anal Question Source: New Comment: Proposed reference to be provided to applicant during examination: NONE Technical

Reference:

40EP-9EO11 (LMFRP) K&A: Knowledge of EOP mitigation strategies: Loss of RHR Learning Objective: L56595 Given the LMFRP HR-2 is being performed, and SDC is in service describe how adequate SDC flow is determined and what actions may be taken if adequate flow cannot be maintained in accordance with 40EP-9EO11. REV 1

ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Senior Reactor Operator NRC Exam Justification: A. Incorrect: Appendix 241 allows the use of either LPSI or CS pump. Candidate must understand that the 'A' Auxiliaries are unavailable due to the A EW HX being out of service. B. Correct: For the current lineup, Appendix 241 directs per step 2 to use LPSI A as the SDC pump. C. Incorrect: Appendix 241 allows for the use of the CS pump and 40OP-9SI02 addresses the use of CS pumps for emergency operations from a SDC Train B lineup. D. Incorrect: This is the correct action, but 40OP-9SI01 does not address the cross tie for LPSI pumps only CS pumps. REV 1

ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Senior Reactor Operator NRC Exam 3. This Exam Level: SRO Appears on: SRO EXAM 2012 Tier 1 Group 1 K/A #: 4.1 038 EA2.15 Importance 4.4 Rating: Given the following conditions: x Unit 2 has tripped from 100% power. x SG 1 AFW Flow is 0 gpm. x SG 1 pressure is 1165 psia and stable. x SG 1 level is 10% NR and rising. x SG 2 AFW Flow is 150 gpm. x SG 2 pressure is 1170 psia and stable. x SG 2 level is 60% WR and rising. x Pressurizer level is 35% and stable. x RCS pressure is 1300 psia and stable. x RCPs 1A & 2A are operating. x Thot is 500°F and stable. x Tcold is 497°F and stable. x HPSI has been throttled. x SPTAs are complete. Which ONE of the following describes the appropriate procedure and action needed to mitigate this event? The CRS will enter ____(1)____ AND reduce RCS pressure to less than ____(2)____ psia. A. (1) 40EP-9EO03 (LOCA) (2) 960 B. (1) 40EP-9EO04 (SGTR) (2) 960 C. (1) 40EP-9EO03 (LOCA) (2) 1135 D. (1) 40EP-9EO04 (SGTR) (2) 1135 Answer: D Reference Id: Q43905 Difficulty: 3.00 Time to complete: 3 10CFR Category: CFR 55.43 (5) 55.43 (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. Cognitive Level: Comprehension / Anal Question Source: New Comment: Proposed reference to be provided to applicant during examination: NONE Technical

Reference:

40EP-9EO04 (SGTR) REV 1

ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Senior Reactor Operator NRC Exam K&A: Ability to determine or interpret the following as they apply to a SGTR: Pressure at which to maintain RCS during S/G cooldown. Learning Objective: L11226 Given the SGTR EOP is being used and given plant conditions determine an appropriate pressure target for depressurization and state the basis for this value. Justification: A. Incorrect: LOCA is the incorrect procedure due to the indications of SGTR. Candidate may select LOCA based on the Low PZR Pressure and Level. 960 psia is the MSIS setpoint pressure but the correct pressure is < 1135 psia and 1165 +/- 50 psia. B. Incorrect: SGTR is the correct procedure but 960 psia is the MSIS setpoint pressure but the correct pressure is < 1135 psia and 1165 +/- 50 psia. C. Incorrect: LOCA is the incorrect procedure due to the indications of SGTR. Candidate may select LOCA based on the Low PZR Pressure and Level.correct pressure is < 1135 psia and 1165 +/- 50 psia. D. Correct: Per Step 12 of 40EP-9EO04 (SGTR), Maintain pressurizer pressure within ALL of the following criteria:

Less than 1135 psia
Approximately equal to the pressure of the Steam Generator with the tube rupture (+/- 50 psi) correct pressure is < 1135 psia and 1165 +/- 50 psia.

REV 1

ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Senior Reactor Operator NRC Exam 4. This Exam Level SRO Appears on: SRO EXAM 2008 SRO EXAM 2012 Tier 1 Group 1 K/A # 4.2 056 AA2.09 Importance 2.9 Rating: Given the following conditions: x Unit 2 is operating at 100% power. x NAN-S02 Fast Bus Transfer is blocked due to SWYD maintenance. x The "A" and "C" Containment Normal ACUs are running. x The "B" and "D" Containment Normal ACUs are in standby. x The "A" and "B" Normal Chillers are running. x NAN-S01 bus faults and de-energizes. x All equipment actuates as expected. Which ONE of the following describes the appropriate procedure the CRS should implement? A. 40EP-9EO07 (LOOP) due to a loss of 4 RCPs. Normal containment cooling can be restored by energizing NAN-S02 from Offsite power. B. 40EP-9EO07 (LOOP) due to a loss of 4 RCPs. Normal containment cooling will be restored by the auto start of the "B" and "D" ACU units. C. 40EP-9EO02 (Reactor Trip) due to a loss of 2 RCPs. Normal containment cooling can be restored by energizing NAN-S02 from Offsite power. D. 40EP-9EO02 (Reactor Trip) due to the loss of 2 RCPs. Normal containment cooling will be restored by the auto start of the "B" and "D" ACU units. Answer: A Reference Id: Q43903 Difficulty: 3.00 Time to complete: 3 10CFR Category: CFR 55.43 (5) 55.43 (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. Cognitive Level: Comprehension / Anal Question Source: PV Bank Not Modified Comment: Proposed reference to be provided to applicant during examination: None Technical

Reference:

40EP-0EO07 (LOOP) K&A: Ability to determine and interpret the following as they apply to the Loss of Offsite Power: Operational status of reactor building cooling unit. Learning Objective: 74452 Describe the automatic functions associated with the Containment Building Normal ACU Fans (HCN-A01-A, B, C, & D) REV 0

ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Senior Reactor Operator NRC Exam Justification: A. Correct: LOOP/LOFC due to the loss of 4 RCPs and both NAN-S01/S02 de-energized even though the switchyard is still energized. The operators have the ability to manually energize S02 following the event to restore normal containment cooling. B. Incorrect:LOOP is correct but the B/D units have no power for the auto start and the "A" (PBA-S03) normal chiller will have to be manually started in addition NC pumps have no power till S02 is energized. C. Incorrect: all 4 RCPs trip due the loss of NAN-S01 tripping 2 RCPs causing a Rx trip/Turbine trip and a subsequent loss of NAN-S02 due fast bus transfer blocked on the 2 side. Core Heat Removal Safety Function will not be met due to Natural Circulation Delta T being > 10 0F. D. Incorrect: 4 RCPs trip and the B/D units have no power for the auto start and the "A" (PBA-S03) normal chiller will have to be manually started in addition NC pumps have no power till S02 is energized. Core Heat Removal Safety Function will not be met due to Natural Circulation Delta T being > 10 0F. REV 0

ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Senior Reactor Operator NRC Exam 5. This Exam Level: SRO Appears on: SRO EXAM 2012 Tier 1 Group 1 K/A #: 4.2 057 AA2.19 Importance 4.3 Rating: Given the following conditions: x Unit 1 is operating at 100% power. x PPS TRBL/GRND alarm on B05. x All initiation relay lights are extinguished on Channel A and Channel C. x PKA, PKB, PKC, and PKD are energized. x All initiation relays on Channels B and D are energized. Which ONE of the following describes the impact on the plant and the procedure entry required? A. No RTSG breakers have tripped, enter 40AO-9ZZ13(Loss of Class Control Power). B. Two RTSG breakers are tripped, enter 40AO-9ZZ13(Loss of Class Control Power). C. No RTSG breakers have tripped, enter 40AO-9ZZ17(Inadvertent ESFAS Initiation). D. Two RTSG breakers are tripped, enter 40AO-9ZZ17(Inadvertent ESFAS Initiation). Answer: B Reference Id: Q43887 Difficulty: 2.00 Time to complete: 2 10CFR Category: CFR 55.43 (5) 55.43 (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. Cognitive Level: Comprehension / Anal Question Source: New Comment: Proposed reference to be provided to applicant during examination: NONE Technical

Reference:

40AO-9ZZ13 (Loss of Class Instrument and Control Power) K&A: L11089 Ability to determine and interpret the following as they apply to the Loss of Vital AC Instrument Bus: The plant automatic actions that will occur on the loss of a vital ac electrical instrument bus Learning Objective: L11089 Given a loss of PK and/or PN describe how the RPS responds to the power loss in accordance with 40AO-9ZZ13. Justification: A. Incorrect: PNA and PNC have tripped which will result in RTSGs 1 and 3 opening due to the loss of PNA and PNC. 40AO-9ZZ13 Loss of Class instrument or control power is the correct procedure. REV 0

ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Senior Reactor Operator NRC Exam B. Correct: De-energizing initiation relays on Channel A and C will result in RTSGs 1 and 3 opening due to the loss of PNA OR PNCeither loss will send the crew to 40AO-9ZZ13 Loss of Class instrument or control power.Entry conditions for Inadvertent ESFAS are not met and will not correct this condition. C. Incorrect: PNA and PNC have tripped which will result in RTSGs 1 and 3 opening due to the loss of PNA and PNC. D. Incorrect: De-energizing initiation relays on Channel A and C will result in RTSGs 1 and 3 opening due to the loss of PNA OR PNC. Entry conditions for Inadvertent ESFAS are not met and will not correct this condition. REV 0

ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Senior Reactor Operator NRC Exam 6. This Exam Level SRO Appears on: SRO EXAM 2009 SRO EXAM 2012 Tier 1 Group 1 K/A # 4.4 E05 EA2.2 Importance 4.2 Rating: Given the following conditions: x Pressurizer pressure is 1600 psia and stable. x RCS temperature is being controlled with SG 2 x Loop 1 T-cold is 362°F and stable. x Loop 1 T-hot is 390°F and stable. x Loop 2 T-cold is 380°F and stable.. x Loop 2 T-hot is 395°F and stable. x REP CET is 397°F and stable. x SIAS, CIAS, MSIS, and CSAS have automatically actuated. x Safety Injection flow is adequate. x There is no activity present in the steam plant or containment. x SG 1 WR level is 0%. x SG 2 WR level is 65% and rising. The CRS should implement ____(1)____ AND ____(2)____. A. (1) 40EP-9EO05 (ESD) (2) equalize loop T-colds at 362 °F then initiate a cooldown. B. (1) 40EP-9EO05 (ESD) (2) lower RCS pressure to within Pressure/Temperature limits. C. (1) 40EP-9EO09 (FRP) HR is jeopardized (2) equalize loop T-colds at 362 °F then initiate a cooldown. D. (1) 40EP-9EO09 (FRP) HR is jeopardized (2) lower RCS pressure to within Pressure/Temperature limits. Answer: B Reference Id: Q43904 Difficulty: 4.00 Time to complete: 4 10CFR Category: CFR 55.43 (5) 55.43 (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. Cognitive Level: Comprehension / Anal Question Source: PV Bank Not Modified Comment: Proposed reference to be provided to applicant during examination: 40OP-9EO010 (Standard Appendix) 2 Pages 1 and 2 Technical

Reference:

ESD, 40EP-9EO06 / Tech guide and standard appendices K&A: Adherence to appropriate procedures and operation within the limitations in the facility*s license and amendments. Excess Steam Demand REV 0

ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Senior Reactor Operator NRC Exam Learning Objective: L11210 Given that the EOPs are being performed and specific plant conditions are given, determine whether or not the plant is over subcooled, and if it is what actions must be taken in accordance with the appropriate procedure. Justification: A. Incorrect: Equalizing Loop Tcolds is required per step 14 of ESD, but a 2 hour soak is required. B. Correct: Per Step 14 a and Step 14 e. of ESD Maintain Tc within the P/T limits of App 2 and if PT limits were exceeded and RCPs are secured then a 2 hour soak is required at current conditions. C. Incorrect: Equalizing Loop Tcolds is required per step 14 of ESD, but a 2 hour soak is required. The FRP is not the appropriate ORP due to single event in progress and HR is not jeopardized. D. Incorrect: Action is correct but the FRP is not the appropriate ORP due to single event in progress and HR is not jeopardized. REV 0

RCS Press Temp Limits Normal CTMT Conditions 2500 100 °F/hr Cooldown 200 °F Subcooled 2000 RCP NPSH 1500 STANDARD APPENDICES Appendix A di 2, 2 PALO VERDE NUCLEAR GENERATING STATION 1000 Figures 350 psia transition line RCS Pressure (psia) QSPDS no longer useful Minimum Subcooled 500 Appendix 2 40EP-9EO10 Revision: 65 SDC Region 0 0 50 100 150 200 250 300 350 400 450 500 550 600 RCS Temperature (Th °F) Page 1 of 3 Page 18 of 1280 Forced Circulation - Th indication used Natural Circulation - REP CET used

RCS Press Temp Limits Harsh CTMT Conditions 2500 200 °F Subcooled 100 °F/hr Cooldown 2000 RCP NPSH 1500 STANDARD APPENDICES PALO VERDE NUCLEAR GENERATING STATION 350 psia transition line 1000 QSPDS no longer useful RCS Pressure (psia) Minimum Subcooled 500 Appendix 2 40EP-9EO10 Revision: 65 SDC Region 0 0 50 100 150 200 250 300 350 400 450 500 550 600 RCS Temperature (Th °F) Page 2 of 3 Page 19 of 1280 Forced Circulation - Th indication used Natural Circulation - REP CET used

ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Senior Reactor Operator NRC Exam 7. This Exam Level: SRO Appears on: SRO EXAM 2012 Tier 1 Group 2 K/A #: 003 2.2.38 Importance Rating: 4.5 Given the following conditions: x Unit 2 is operating at 100% ARO. x A Regulating Group 5 CEA has dropped completely into the core. x All required actions are complete. Which ONE of the following describes Technical Specification 3.1.5 (CEA Alignment)? CEA alignment must be restored within a maximum of ______ hour(s). A. 1 B. 2 C. 6 D. 12 Answer: B Reference Id: Q43907 Difficulty: 3.00 Time to complete: 3 10CFR Category: CFR 55.43 (1) 55.43 (1) Conditions and limitations in the facility license. Cognitive Level: Memory Question Source: New Comment: Proposed reference to be provided to applicant during examination: NONE Technical

Reference:

Technical Specification 3.1.5 K&A: Knowledge of conditions and limitations in the facility license Learning Objective: L89763 Given plant conditions and Technical Specification action statements that are greater than one hour apply the action statements that are greater than one hour for T.S. 3.1 in accordance with Tech Spec 3.1. Justification: A. Incorrect: 1 hour applies to reducing THERMAL POWER in accordance with the COLR. B. Correct: TS 3.1.5 Condition A.2 requires CEA alignment to be restored within 2 hours. C. Incorrect: 6 hours applies to being in Mode 3 within 6 hours if the CEA alignment or Power limit if condition A can not be met. D. Incorrect: 12 hours applies to the frequency that CEAs with inoperable position indicators be verified. REV 0

ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Senior Reactor Operator NRC Exam 8. This Exam Level: SRO Appears on: SRO EXAM 2012 Tier 1 Group 2 K/A #: 4.2 024 AA2.06 Importance 3.7 Rating: Given the following conditions: Initial Conditions: x Unit 1 is operating at 100%. x CEDMCS is in Automatic. x A New Purification Letdown Ion Exchanger was just placed in service at the end of last shift. x Tavg is 591 0F and rising slowly. Subsequently: x A Low Rate CEA insertion demand exists. x CEAs begin inserting. Which ONE of the following would cause this condition and what procedure will be used to respond? A. RWT to CVCS gravity feed isolation (CHE-HV-536) is leaking by, isolate per 40OP-9CH02 (Purification System). B. New letdown IX not appropriately borated prior to placing in service, isolate per 40OP-9CH02 (Purification System). C. New letdown IX not appropriately borated prior to placing in service, borate the RCS per 40OP-9CH01 (CVCS Normal Operations). D. RWT to CVCS gravity feed isolation (CHE-HV-536) is leaking by, borate the RCS per 40OP-9CH01 (CVCS Normal Operations). Answer: C Reference Id: Q43890 Difficulty: 3.00 Time to complete: 3 10CFR Category: CFR 55.43 (5) 55.43 (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. Cognitive Level: Comprehension / Anal Question Source: New Comment: Proposed reference to be provided to applicant during examination: NONE Technical

Reference:

40OP-9CH01 (CVCS Normal Operations). K&A: Ability to determine and interpret the following as they apply to the Emergency Boration: When boron dilution is taking place. Learning Objective: L63180 Given that a dilution of the RCS is occurring and 40AO9ZZ01 has been entered identify how the dilution will be mitigated REV 0

ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Senior Reactor Operator NRC Exam Justification: A. Incorrect: CHE-HV-536 is the RWT gravity feed isolation valve, the RWT is a borated source of water that if it were to leak by RCS Temperature would lower. 40OP-9CH02 is the procedure that provides direction to borate the IX prior to placing in service, but doesn't provide direction to isolate and borate the RCS to remedy to situation. B. Incorrect: An IX that has not been appropriately borated will resulting in the RCS temperature rise and the CEA insertion, but 40OP-9CH02 is the procedure that provides direction to borate the IX prior to placing in service, but doesn't provide direction to isolate and borate the RCS to remedy to situation. C. Correct: An IX that has not been appropriately borated will result in the RCS temperature rise and the CEA insertion. 40OP-9CH01 is the procedure that directs borating the RCS to maintain Tc on program. D. Incorrect: CHE-HV-536 is the RWT gravity feed isolation valve, the RWT is a borated source of water that if it were to leak by RCS Temperature would lower. 40OP-9CH01 is the procedure that directs borating the RCS to maintain Tc on program. REV 0

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ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Senior Reactor Operator NRC Exam 10. This Exam Level: SRO Appears on: SRO EXAM 2007 SRO EXAM 2012 Tier 1 Group 2 K/A #: 4.4 E09 EA2.2 Importance Rating: 4.0 Given the following conditions: Radiation Monitor status just prior to Reactor trip is as follows: x RU-139 (Main Steam Line SG #1) is in ALERT alarm. x RU-140 (Main Steam Line SG #2) is in HIGH alarm. x RU-142 (Main Steam Line N-16) channels 1/2 are ALERT alarm. x RU-142 (Main Steam Line N-16) channels 3/4 are in HIGH alarm. Current plant conditions: x SG #1level is 51% WR and rising. x SG #1 pressure is 1200 psi and stable. x SG #2 level is 28% WR and lowering. x SG #2 pressure 1070 psi and lowering. x Containment temperature is 195°F. x Containment pressure 9.0 psig. x RCPs have been tripped. x All expected ESFAS actuations have initiated. x RU-16, Containment Operating Level Monitor, is in ALERT alarm. x SPTAs are complete. Which ONE of the following mitigation strategies would the CRS direct? A. Feed #1 SG at 1360 - 1600 gpm to 45% NR B. Feed #2 SG at 1360 - 1600 gpm to 45% NR C. Feed #1 SG to 45% NR, Secure feed to #2 SG D. Feed #2 SG to 45% NR, Secure feed to #1 SG Answer: C Reference Id: Q10294 Difficulty: 3.00 Time to complete: 3 10CFR Category: CFR 55.43 (5) 55.43 (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. Cognitive Level: Comprehension / Anal Question Source: PV Bank Not Modified Comment: Proposed reference to be provided to applicant during examination: NONE Technical

Reference:

40EP-9EO09 (FRP) REV 0

ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Senior Reactor Operator NRC Exam K&A: Ability to determine and interpret the following as they apply to the (Functional Recovery): Adherence to appropriate procedures and operation within the limitations in the facility's license and amendments Learning Objective: L90459 Diagnose FRP event in progress Justification: A. Incorrect: 1360-1600 gpm is the strategy for a SGTR with steam releasing to atmosphere for the Ruptured (SGTR) #1 SG. SG #2 is the Faulted (ESD) SG. Candidate may feed the Ruptured (SGTR) #1 SG since a Dual Event ESD/SGTR is in progress. B. Incorrect: 1360-1600 gpm is the strategy for a SGTR with steam releasing to atmosphere. Candidate may feed the Faulted (ESD) #2 SG since a Dual Event ESD/SGTR is in progress. C. Correct: SG #1 is not faulted so it should be restored to 45 -60% NR, we are not expected feed a faulted SG with another available for Heat Removal. D. Incorrect: SG #2 is faulted; feeding would add to the cooldown, SG #1 is available for HR. REV 0

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ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Senior Reactor Operator NRC Exam 12. This Exam Level: SRO Appears on: SRO EXAM 2012 Tier 2 Group 1 K/A #: 3.2 006 A2.04 Importance 3.8 Rating: Given the following conditions: Initial Conditions: x Unit 1 automatically tripped from 100% power. x SPTAs are in progress. x The crew has manually initiated SIAS/CIAS. x Adequate SI flow has been verified. x 525 KV East and West Bus Voltage meters indicate 0 Vac. x Pressurizer pressure is 1450 psia and lowering. x Pressurizer level is 20% and lowering. x SG 1 & 2 pressures being controlled at 1180 psia with ADVs. x PBA-S03 is energized by DG "A". x DG "B" has tripped on "overspeed". Subsequently: x HPSI pump "A" discharge pressure degrades to 1000 psig. Which ONE of the following describes the impact on Safety Injection and the appropriate procedure to be used to mitigate? HPSI flow lowers to .. A. zero (0) gpm, utilize 40EP-9EO03 (LOCA). B. half its original value, utilize 40EP-9EO03 (LOCA). C. zero (0) gpm, utilize 40EP-9EO09 (FRP) MVAC-2 DGs. D. half its original value, utilize 40EP-9EO09 (FRP) MVAC-2 DGs. Answer: C Reference Id: Q44014 Difficulty: 3.00 Time to complete: 3 10CFR Category: CFR 55.43 (5) 55.43 (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. Cognitive Level: Comprehension / Anal Question Source: New Comment: Proposed reference to be provided to applicant during examination: NONE Technical

Reference:

40OP-9EO09 (FRP) REV 0

ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Senior Reactor Operator NRC Exam K&A: Ability to (a) predict the impacts of the following malfunctions or operations on the ECCS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Improper discharge pressure. Learning Objective: 65106 Describe how the FRP will maintain or recover the Maintenance of Vital Auxiliaries. Justification: A. Incorrect: Per Standard Appendix 2, HPSI Pump Delivery Curves, 0 gpm does not meet the acceptable region of the curve, due to the Loss of Offsite Power (LOOP) and the HPSI A degraded condition, HPSI A is not available, the Loss of the EDG B results in HPSI B not being available. The LOCA procedure does not provide direction to Crosstie PB busses to restore electrical power to the undamaged HPSI B. B. Incorrect: HPSI Flow will drop to 0 not half. HPSI B is not available and HPSI A is operating below the pressure of the RCS. Due to the Loss of Offsite Power (LOOP) and the HPSI A degraded condition, HPSI A is not available, the Loss of the EDG B results in HPSI B not being available. The LOCA procedure does not provide direction to Crosstie PB busses to restore electrical power to the undamaged HPSI B. C. Correct: Due to the Loss of Offsite Power (LOOP) and the loss of PBA-S03 along with the HPSI A degraded condition HPSI A is not available, the Loss of the EDG B results in HPSI B not being available. FRP MVAC-2 will provide direction to restore electrical power to PBB-S04 and start HPSI B to restore adequate HPSI delivery. D. Incorrect: HPSI Flow will drop to 0 not half. HPSI B is not available and HPSI A is operating below the pressure of the RCS Per Standard Appendix 2, HPSI Pump Delivery Curves, 0 gpm does not meet the acceptable region of the curve. FRP MVAC-2 is the correct procedure. REV 0

ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Senior Reactor Operator NRC Exam 13. This Exam Level: SRO Appears on: SRO EXAM 2012 Tier 2 Group 1 K/A #: 3.3 010 A2.02 Importance 3.9 Rating: Given the following conditions: x Unit 1 is operating at 100% power. x PZR pressure was reported as 2230 psia and lowering. x Main spray valves 100E & 100F indicate full open. x All attempts to close Main Spray valves have failed. x Pressurizer pressure is 2050 psia and continuing to lower. This will cause the RCN-PIC-100 (PPCS master controller) output to go to _____(1)____ and the CRS should _____(2)_____. A. (1) minimum, (2) trip the Reactor, stop all 4 RCPs and enter 40EP-9EO07 (LOOP/LOFC). B. (1) maximum, (2) trip the Reactor, stop the Loop 1 RCPs only and enter 40EP-9EO02 (Reactor Trip). C. (1) minimum, (2) trip the Reactor, stop two RCPs when SIAS/CIAS initiates and enter 40EP-9EO02 (Reactor Trip). D. (1) maximum (2) close IAA-UV-2 (IA CTMT Isolation) per 40AL-9RK4A (B04A ARP), Main Spray valves will close immediately. Answer: A Reference Id: Q43920 Difficulty: 2.00 Time to complete: 2 10CFR Category: CFR 55.43 (5) 55.43 (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. Cognitive Level: Comprehension / Anal Question Source: New Comment: Proposed reference to be provided to applicant during examination: NONE Technical

Reference:

40AL-9RK4A (Panel B04A ARP), 40EP-9EO07 (LOOP/LOFC) K&A: Ability to (a) predict the impacts of the following malfunctions or operations on the PZR PCS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Spray valve failures Learning Objective: L75344 Describe the response of the Pressurizer Pressure Control System to a failure of an input transmitter. REV 0

ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Senior Reactor Operator NRC Exam Justification: A. Correct:The Pressurizer Pressure Master Controller is a reverse acting controller. A decrease in controller output results in an increase in system pressure. The B04A Alarm Response procedure directs stopping all RCPs. The Crew must trip the reactor to stop all 4 RCPs. Tripping all 4 RCPs will result in a LOFC. B. Incorrect: The Pressurizer Pressure Master Controller is a reverse acting controller. Examine may pick this distracter since spray valves come off the Loop 1 cold legs but will not completely stop the pressure decrease. Reactor Trip is not the appropriate procedure. C. Incorrect: Shutting IAA-UV-2 was previously an option in the B04A Alarm Response procedure. PVNGS experienced a plant event where IA was isolated to CTMT and IA pressure maintained Spray Valves open well past the expected response time. D. Incorrect: This is a strategy for decreasing pressure when a LOCA is diagnosed. Reactor Trip is not the appropriate procedure. REV 0

ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Senior Reactor Operator NRC Exam 14. This Exam Level SRO Appears on: SRO EXAM 2012 Tier 2 Group 1 K/A # 059 2.4.11 Importance 4.2 Rating: Given the following conditions: Initial conditions: x Unit 1 is operating at 80% power. Subsequently: x The B Main Feedwater Pump Trips. x CEA Subgroups 4, 5, and 22 drop to the bottom of the core. x CEA 67 (Regulating Group 2, 4 Finger CEA) slips 3 inches, to 147 inches withdrawn. x Turbine Load is approximately 940 MW. Which ONE of the following describes the actions directed by 40AO-9ZZ09 (RPCB Loss of Feedpump)? A. Trip the reactor. B. Adjust turbine load to 65% or less (~ 890 MW). C. Borate the RCS to reduce reactor power to ~ 12%. D. Manually insert CEAs to match reactor and turbine power. Answer: B Reference Id: Q43913 Difficulty: 3.00 Time to complete: 3 10CFR Category: CFR 55.43 (5) 55.43 (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. Cognitive Level: Comprehension / Anal Question Source: New Comment: Proposed reference to be provided to applicant during examination: NONE Technical

Reference:

40AO-9ZZ09,RPCB (loss of Feedpump) K&A: Knowledge of abnormal condition procedures. Learning Objective: L56804 Describe the contingency action(s) that the operator would be required to take if RPCB does not operate properly. OPTRNG_EXAM Page: 1 of 2 2012/01/12

ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Senior Reactor Operator NRC Exam Justification: A. Incorrect: The AOP directs tripping the reactor if CEA deviation is > 6.6 inches. B. Correct: Contingency action for step 4 is Reduce the load limit potentiometer until the Main Turbine load is 65% or less (~890 MW). C. Incorrect: Manually inserting CEAs to match turbine load is only an action if Initial Rx Power was less than 74%. D. Incorrect: This action is for a RPCB due to a Load Rejection. OPTRNG_EXAM Page: 2 of 2 2012/01/12

ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Senior Reactor Operator NRC Exam 15. This Exam Level: SRO Appears on: SRO EXAM 2008 SRO EXAM 2012 Tier 2 Group 1 K/A #: 3.8 078 A2.01 Importance 2.9 Rating: Given the following conditions: x Unit 1 is operating at 100% power. x Instrument Air System (IA) is aligned for normal operation. x The following alarm is received on B07. INSTAIRHDR PRESSLO x IA system pressure is continuing to trend down slowly. x Instrument Air Dryer IAN-M01C is in service. x A large differential pressure exists between air receiver pressure and pressure downstream of Instrument Air Dryer IAN-M01C. Which ONE of the following describes the impact to the IA system and the appropriate procedural action? A. Instrument Air Dryers will automatically shift at 80 psig, implement 40AO-9ZZ06 (Loss of Instrument Air), to verify the shift. B. Instrument Air Dryers will automatically shift at 80 psig, implement 40AL-9RK7B (Window 01B INST AIR HDR PRESS LO), to verify the shift. C. The IA header pressure will continue to LOWER until the nitrogen backup valve opens, implement 40AO-9ZZ06 (Loss of Instrument Air), to valve in another air dryer. D. The IA header pressure will continue to LOWER until the nitrogen backup valve opens, implement 40AL-9RK7B (Window 01B INST AIR HDR PRESS LO), to valve in another air dryer. Answer: C Reference Id: Q43888 Difficulty: 2.00 Time to complete: 2 10CFR Category: CFR 55.43 (5) 55.43 (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. Cognitive Level: Comprehension / Anal Question Source: PV Bank Not Modified Comment: Proposed reference to be provided to applicant during examination: NONE Technical

Reference:

40AO-9ZZ06 (Loss of Instrument Air) 40AL-9RK7B (B07B ARP) OPTRNG_EXAM Page: 1 of 2 2012/01/12

ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Senior Reactor Operator NRC Exam K&A: Ability to (a) predict the impacts of the following malfunctions or operations on the IAS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Air dryer and filter malfunction Learning Objective: L56781 Determine the mitigating strategies of the Loss of Instrument air AOP. Justification: A. Incorrect: IA Dryers have no automatic shift, the automatic features associated with the IA system is the N2 Backup alignment and Stby IA comp starting. 40AO-9ZZ06 (Loss of IA) is the correct procedure. B. Incorrect: IA Dryers have no automatic shift, the automatic features associated with the IA system is the N2 Backup alignment and Stby IA comp starting. 40AL-9RK7B will not direct actions to shift the air dryers. C. Correct: N2 Backup valve automatically opens at 85 psig to maintain system pressure and 40AO-9ZZ06 (Loss of IA) should be implemented to align the other IA dryer to mitigate the effects. D. Incorrect: N2 Backup valve automatically opens at 85 psig to maintain system pressure but 40AL-9RK7B will not direct actions to shift the air dryers. OPTRNG_EXAM Page: 2 of 2 2012/01/12

ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Senior Reactor Operator NRC Exam 16. This Exam Level: SRO Appears on: SRO EXAM 2012 Tier 2 Group 1 K/A #: 2.2.40 Importance 4.7 Rating: Given the following conditions: x Unit 2 is in Mode 4 during a refueling outage. x RCS Pressure is 450 psia and stable. The STA has determined that RCA-HV-106 (PZR/RV HEAD VENT TO CTMT) is INOPERABLE. Given the supplied references, which ONE of the following describes the required action (if any) per Technical Specification 3.4.12 (Pressurizer Vents)? A. No action required due to only ONE (1) path is INOPERABLE. B. Restore ONE (1) additional pressurizer vent paths to OPERABLE within 6 hours. C. Restore TWO (2) additional pressurizer vent to OPERABLE status within 72 hrs. D. Restore THREE (3) additional pressurizer vent paths to OPERABLE within 7 days. Answer: C Reference Id: Q43813 Difficulty: 3.00 Time to complete: 3 10CFR Category: CFR 55.43 (2) 55.43 (2)Facility operating limitations in the technical specifications and their bases. Cognitive Level: Comprehension/Anal Question Source: New Proposed reference to be provided to applicant during examination: Tech Spec 3.4.12, Tech Spec Basis for 3.4.12 and Diagram of PZR Vents from 40OP-9RC04 (RCGVS) Technical

Reference:

Tech Specs OPERATING EXPERIENCE QUESTION K&A: Ability to apply Technical Specifications for a system: RCS Learning Objective: Given conditions when an LCO is not met, apply Tech Spec Section 3.4.12 (PZR Vents) in accordance with Tech Spec 3.4.12. Justification: A. Incorrect - Candidate may read the Tech Spec as No Action due to only one vent path INOPERABLE. B. Incorrect - This would be correct if the candidate does not understand that one valve RCN-HV-106 being INOPERABLE actually results in two vent paths being INOPERABLE. In this Case 2 are INOPERABLE. C. Correct - This will ensure that all 4 vent paths are OPERABLE and the LCO can be exited. D. Incorrect - When RCN-HV-106 is INOPERABLE 2 vent paths are then INOPERABLE. Therefore 3 is incorrect, and the time requirement is 72 hrs. REV 0

Pressurizer Vents 3.4.12 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.12 Pressurizer Vents LCO 3.4.12 Four pressurizer vent paths shall be OPERABLE. APPLICABILITY: MODES 1, 2, and 3. MODE 4 with RCS pressure 385 psia. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Two or three required A.1 Restore required 72 hours pressurizer vent paths pressurizer vent inoperable. paths to OPERABLE status. B. All pressurizer vent B.1 Restore one 6 hours paths inoperable. pressurizer vent path to OPERABLE status. C. Required Action and C.1 Be in MODE 3. 6 hours associated Completion Time of Condition A, AND or B not met. C.2 Be in MODE 4 with RCS 24 hours pressure < 385 psia. SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.12.1 Perform a complete cycle of each 18 months Pressurizer Vent Valve. SR 3.4.12.2 Verify flow through each 18 months pressurizer vent path. PALO VERDE UNITS 1,2,3 3.4.12-1 AMENDMENT NO. 117

PVNGS NUCLEAR ADMINISTRATIVE AND TECHNICAL MANUAL Page 16 of 16 Revision Reactor Coolant Gas Vent System (RCGVS) 40OP-9RC04 11 Appendix C Page 1 of 1 Appendix C - RV Head and Pressurizer Vent System RCN-V392 RCB-HV-109 CONTAINMENT ATMOS S S RCA-HV-106 RCE-V006 VENT TO GAS SURGE HDR ATMOS RCB- S HV-102 S RCB-RCA- HV-108 RCB- S S HV-103 CHN- S HV-105 RCA-UV-540 HV-101 S RCN-RCN-V212 V090 CHN- RCE-V007 HV-923 Reactor Vessel Reactor Head Pressurizer Drain Tank This diagram is only a simplified likeness of system diagrams M-RCP-001 and M-CHP-003 End of Appendix C

Pressurizer Vents B 3.4.12 B 3.4 REACTOR COOLANT SYSTEM (RCS) B 3.4.12 Pressurizer Vents BASES BACKGROUND The pressurizer vent is part of the reactor coolant gas vent system (RCGVS) as described in UFSAR 18.II.B.1 (Ref. 1). The pressurizer can be vented remotely from the control room through the following four paths (see UFSAR Figure 18.II.B-1):

1. From the pressurizer vent through SOV HV-103, then through SOV HV-105 to the reactor drain tank (RDT).

2. From the pressurizer vent through SOV HV-103, then through SOV HV-106 directly to the containment atmosphere.

3. From the pressurizer vent through SOVs HV-108 and HV-109, then through SOV HV-105 to the reactor drain tank (RDT).

4. From the pressurizer vent through SOVs HV-108 and HV-109, then through SOV HV-106 directly to the containment atmosphere.

The RCGVS also includes the reactor head vent, which can be used along with the pressurizer vent to remotely vent gases that could inhibit natural circulation core cooling during post accident situations. However, this function does not meet the criteria of 10 CFR 50.36(c)(2)(ii) to require a Technical Specification LCO, and therefore the reactor head vent is not included in these Technical Specifications. (continued) PALO VERDE UNITS 1,2,3 B 3.4.12-1 REVISION 1

Pressurizer Vents B 3.4.12 BASES APPLICABLE The requirement for the pressurizer vent path to be SAFETY ANALYSES OPERABLE is based on the steam generator tube rupture (SGTR) with loss of offsite power (SGTRLOP) and SGTR with loss of offsite power and single failure (SGTRLOPSF) analysis, as described in UFSAR 15.6.3 (Ref. 4). It is assumed that the auxiliary pressurizer spray system (APSS) is not available for this event. Instead, RCS depressurization is performed by venting the RCS via a pressurizer vent path and throttling HPSI flow. The analysis assumes venting to the containment atmosphere via path 4 as described below. The results of the CENTS based analysis for SGTRLOP and SGTRLOPSF forwarded to the NRC in Reference 2 states that the auxiliary spray was assumed to be unavailable and use of pressurizer head vents was credited for de-pressurization. The staff has reviewed and accepted the results of the analysis. The staff's detailed evaluation has been reported in Amendment No. 149, which increases power to 3990 MWt for Unit 2 and incorporates replacement steam generator (Ref. 3). The pressurizer vent paths satisfy Criterion 3 of 10 CFR 50.36 (c)(2)(ii). LCO The LCO requires four pressurizer vent paths be OPERABLE. The four vent paths are:

1. From the pressurizer vent through SOV HV-103, then through SOV HV-105 to the reactor drain tank (RDT).

2. From the pressurizer vent through SOV HV-103, then through SOV HV-106 directly to the containment atmosphere.

3. From the pressurizer vent through SOVs HV-108 and HV-109, then through SOV HV-105 to the reactor drain tank (RDT).

4. From the pressurizer vent through SOVs HV-108 and HV-109, then through SOV HV-106 directly to the containment atmosphere.

(continued) PALO VERDE UNITS 1,2,3 B 3.4.12-2 REVISION 34

Pressurizer Vents B 3.4.12 BASES LCO A vent path is flow capability from the pressurizer to the (continued) RDT or from the pressurizer to containment atmosphere. Loss of any single valve in the pressurizer vent system will cause two flow paths to become inoperable. A pressurizer vent path is required to depressurize the RCS in a SGTR design basis event which assumes LOP and APSS unavailable. APPLICABILITY In MODES 1, 2, 3, and MODE 4 with RCS pressure 385 psia the four pressurizer vent paths are required to be OPERABLE. The safety analysis for the SGTR with LOP and a Single Failure (loss of APSS) credits a pressurizer vent path to reduce RCS pressure. In MODES 1, 2, 3, and MODE 4 with RCS pressure 385 psia the SGs are the primary means of heat removal in the RCS, until shutdown cooling can be initiated. In MODES 1, 2, 3, and MODE 4 with RCS pressure 385 psia, assuming the APSS is not available, the pressurizer vent paths are the credited means to depressurize the RCS to Shutdown Cooling System entry conditions. Further depressurization into MODE 5 requires use of the pressurizer vent paths. In MODE 5 with the reactor vessel head in place, temperature requirements of MODE 5 (< 210°F) ensure the RCS remains depressurized. In MODE 6 the RCS is depressurized. ACTIONS A.1 If two or three pressurizer vent paths are inoperable, they must be restored to OPERABLE status. Loss of any single valve in the pressurizer vent system will cause two flow paths to become inoperable. Any vent path that provides flow capability from the pressurizer to the RDT or to the containment atmosphere, independent of which train is powering the valves in the flow path, can be considered an operable vent path. The Completion Time of 72 hours is reasonable because there is at least one pressurizer vent path that remains OPERABLE. (continued) PALO VERDE UNITS 1,2,3 B 3.4.12-3 REVISION 48

Pressurizer Vents B 3.4.12 BASES B.1 If all pressurizer vent paths are inoperable, then restore at least one pressurizer vent path to OPERABLE status. The Completion Time of 6 hours is reasonable to allow time to correct the situation, yet emphasize the importance of restoring at least one pressurizer vent path. If at least one pressurizer vent path is not restored to OPERABLE within the Completion Time, then Action C is entered. C.1 If the required Actions, A and B, cannot be met within the associated Completion Times, the plant must be brought to a MODE in which the requirement does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours, and to MODE 4 with RCS pressure < 385 psia within 24 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner without challenging plant systems. SURVEILLANCE SR 3.4.12.1 REQUIREMENTS SR 3.4.12.1 requires complete cycling of each pressurizer vent path valve. The vent valves must be cycled from the control room to demonstrate their operability. Pressurizer vent path valve cycling demonstrates its function. The frequency of 18 months is based on a typical refueling cycle and industry accepted practice. This surveillance test must be performed in Mode 5 or Mode 6. SR 3.4.12.2 SR 3.4.12.2 requires verification of flow through each pressurizer vent path. Verification of pressurizer vent path flow demonstrates its function. The frequency of 18 months is based on a typical refueling cycle and industry accepted practice. This surveillance test must be performed in Mode 5 or Mode 6. (continued) PALO VERDE UNITS 1,2,3 B 3.4.12-4 REVISION 0

Pressurizer Vents B 3.4.12 BASES REFERENCES 1. UFSAR, Section 18.

2. "Palo Verde Nuclear Generating Station (PVNGS) Unit 2 Docket No. STN 50-529 Request for a License Amendment to Support Replacement of Steam Generators and Uprated Power Operations," Letter 102-046141-CDM/RAB, C, D.

Mauldin (APS) to the NRC, December 21, 2001.

3. "Palo Verde Nuclear Generating Station, Unit 2 (PVNGS-

2) - Issuance of Amendment on Replacement of Steam Generators and Uprated Power Operations (TAC NO.

MB3696", B.M. Pham (NRC) to G. R. Overbeck (APS), September 29, 2003.

4. UFSAR, Section 15.

PALO VERDE UNITS 1,2,3 B 3.4.12-5 REVISION 31

ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Senior Reactor Operator NRC Exam 17. This Exam Level: SRO Appears on: SRO EXAM 2008 SRO EXAM 2012 Tier 2 Group 2 K/A #: 3.4 041 A2.02 Importance 3.9 Rating: Given the following conditions: x Unit 1 was operating at 100% power. x SBCV #6 failed 100% open. x A reactor trip and MSIS have both automatically initiated. x T-avg dropped to 570°F on the reactor trip. x T-cold dropped to 546°F before the MSIS was initiated. Which ONE of the following describes the impact to the SBCS and the appropriate response? A. SBCS "Quick Open" was blocked on the trip, direct the crew to maintain T-cold at 556°F and implement 40EP-9EO05 (ESD) B. SBCS "Quick Open" was blocked on the trip, direct the crew to restore T-cold to 560-570°F and implement 40EP-9EO02 (Rx Trip) C. SBCS "Quick Open" functioned normally on the trip, direct the crew to maintain T-cold at 556°F and implement 40EP-9EO05 (ESD) D. SBCS "Quick Open" functioned normally on the trip, direct the crew to restore T-cold to 560-570°F and implement 40EP-9EO02 (Rx Trip) Answer: A Reference Id: Q22473 Difficulty: 3.00 Time to complete: 4 10CFR Category: CFR 55.43 (5) 55.43 (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. Cognitive Level: Comprehension / Anal Question Source: PV Bank Not Modified Comment: Proposed reference to be provided to applicant during examination: NONE Technical

Reference:

40EP-9EO05 (ESD) Simplified drawings, LOIT lesson plan K&A: Ability to (a) predict the impacts of the following malfunctions or operations on the SDS; and (b) based on those predictions or mitigate the consequences of those malfunctions or operations: Steam valve stuck open Learning Objective: L65641 Describe the interrelationship between the Steam Bypass Control System and the Main Steam System REV 0

ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Senior Reactor Operator NRC Exam Justification: A. Correct: Quick Open is blocked on Rx trip with T-avg < 573.5°F, OPS expectations requires that ESD be entered if T-cold goes below 560°F due to an ESD event. B. Incorrect: Quick Open is blocked on Rx trip with T-avg < 573.5°F. Examinee may pick any of these others based on lack of system understanding. Rx Trip is not the correct procedure. ESD will stabilize Tcold and Rx Trip will not. C. Incorrect: Quick Open does not function normally due to the low Tavg, ESD is the Correct Procedure. D. Incorrect: Quick Open does not function normally due to the low Tavg, Rx Trip is not the correct procedure. ESD will stabilize Tcold and Rx Trip will not. REV 0

ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Senior Reactor Operator NRC Exam 18. This Exam Level: SRO Appears on: SRO EXAM 2012 Tier 2 Group 2 K/A #: 3.7 072 A2.03 Importance 2.9 Rating: Given the following conditions: Initial Conditions: x Unit 2 is in an outage. x Core Off Load is in progress. x RU-37 ( Power Access Purge Area Monitor Train A) is inoperable and in bypass on BOP-ESFAS. Subsequently: x RU-38 ( Power Access Purge Area Monitor Train B) power supply fuses blow. Which ONE of the following predicts the expected plant response and appropriate actions? CPIAS actuates and provides a cross trip to ____(1)____. IF the CPIAS did not actuate properly the CRS must suspend ____(2)____. A. (1) FBEVAS (2) movement of irradiated fuel assemblies in the fuel building per TRM 3.9.104 (FBEVAS). B. (1) CREFAS (2) movement of irradiated fuel assemblies in the fuel building per Tech Spec 3.3.9 (CREFAS). C. (1) FBEVAS (2) core alterations and movement of irradiated fuel assemblies in the CTMT per Tech Spec 3.3.8 (CPIAS). D. (1) CREFAS (2) core alterations and movement of irradiated fuel assemblies in the CTMT per Tech Spec 3.3.8 (CPIAS). Answer: D Reference Id: Q43922 Difficulty: 2.00 Time to complete: 2 10CFR Category: CFR 55.43 (5) 55.43 (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. Cognitive Level: Memory Question Source: New Comment: Proposed reference to be provided to applicant during examination: NONE Technical

Reference:

Technical Specifications, Technical Requirements Manual. REV 0

ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Senior Reactor Operator NRC Exam OPERATING EXPERIENCE QUESTION K&A: Ability to (a) predict the impacts of the following malfunctions or operations on the ARM system-and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Blown power-supply fuses. Learning Objective: 65049 Explain the operation of the CPIAS Module. Justification: A. Incorrect: CPIAS will cross trip to CREFAS when actuated. The TRM for FBEVAS will not apply and it directs only suspending fuel movements in the Fuel Building. B. Incorrect: A loss of power to the ARM will result in the BOP-ESFAS module sensing a trip and actuating the CPIAS module which will result in a cross trip signal being sent to the CREFAS module. The TS for CREFAS will not apply but it does apply to irradiated fuel assembly movements.. C. Incorrect: CPIAS will cross trip to CREFAS when actuated. TS is the correct procedure. D. Correct: A loss of power to the ARM will result in the BOP-ESFAS module sensing a trip and actuating the CPIAS module which will result in a cross trip signal being sent to the CREFAS module. TS 3.3.8 directs suspending core alterations and movement of irradiated fuel in the CTMT immediately. REV 0

ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Senior Reactor Operator NRC Exam 19. This Exam Level SRO Appears on: SRO EXAM 2010 SRO EXAM 2012 Tier 3 K/A # 2.1.14 Importance 3.1 Rating: Which ONE of the following describes when a plant-wide announcement is required to be made? A. Changing from Mode 3 to Mode 2. B. Energizing PNA-D25 after a permit has been cleared. C. Starting HCN-A01C (CTMT Normal ACU Fan) from the Control Room. D. AFB-P01 (Essential Motor Driven Aux Feed Pump) started automatically on AFAS-1. Answer: A Reference Id: Q43785 Difficulty: 2.00 Time to complete: 2 10CFR Category: CFR 55.43 (5) 55.43 (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. Cognitive Level: Memory Question Source: PV Bank Not Modified Comment: Proposed reference to be provided to applicant during examination: NONE Technical

Reference:

ODP-1, Operations Department Principles and Standards K&A: Knowledge of criteria or conditions that require plant-wide announcements, such as pump starts, reactor trips, mode changes etc. Learning Objective: 30265 ODP-1 Reactivity Management Justification: A. Correct - Plant-wide announcements shall be made when changing modes. B. Incorrect - 120 Vac distribution panels are not required to be announced. C. Incorrect - 480 Vac motor starts are not required to be announced. D. Incorrect - Equipment that starts automatically is not required to be announced. REV 0

ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Senior Reactor Operator NRC Exam 20. This Exam Level: SRO Appears on: SRO EXAM 2008 SRO EXAM 2012 Tier 3 K/A #: 2.1.25 Importance 4.2 Rating: Given the following conditions: x Unit-1 has been shutdown for five days and is currently in Mode 5 x The RCS is being maintained at 102 ft 6 inches in preparation for installing Steam Generator Nozzle Dams x The Steam Generator primary manways are off x RCS temperature is 135 ºF Per the tables found in the Unit-1 Safety Analyses Operational Data (SAOD) during a sustained Loss of Shutdown Cooling the RCS ... A. time to boil is 18.9 minutes B. time to boil is 23.3 minutes C. makeup flowrate to compensate for boil off is 76.9 gpm D. makeup flowrate to compensate for boil off is 98.5 gpm Answer: D Reference Id: Q5424 Difficulty: 4.00 Time to complete: 5 10CFR Category: CFR 55.43 (5) 55.43 (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. Cognitive Level: Comprehension / Anal Question Source: PV Bank Not Modified Comment: Proposed reference to be provided to applicant during examination: Unit-1 Safety Analysis Operational Data (SAOD) Technical

Reference:

Unit-1 Safety Analysis Operational Data (SAOD) K&A: Ability to interpret reference materials, such as graphs, curves, tables, etc. Learning Objective: L56598 Provided with Time to Boil curves, determine time to core boiling using the TTB curves in the back of the core data book and describe what this value is used for in accordance with 40EP-9EO11. REV 0

ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Senior Reactor Operator NRC Exam Justification: A. Incorrect: time to boil at midloop is 14.7 minutes (18.9 comes from flange level after core reload). B. Incorrect: time to boil at midloop is 14.7 minutes (23.3 comes from flange level prior to core reload). C. Incorrect: 76.9 gpm is the makeup requirement for midloop after core reload. D. Correct: this is the makeup rate for midloop prior to core offload. REV 0

By: B.S. Blackmore Safety Analysis Operational Data SAOD Unit 1 Manual 3990 MWt Rev 2 Reviewer: Ness Kilic Page 18 of 40 TABLE 2.4.1 Key Reactor Core Parameters Following a Loss of SDC With the RCS Drained to the Reactor Vessel Flange Reactor Vessel Head On Prior to Core Reload (3990 MW Core) Time after Decay Heatup Makeup Time after Decay Heatup Makeup Reactor Heat Rate Flowrate Reactor Heat Rate Flowrate Shutdown Load (F/Min.) (gpm)** Shutdown Load (F/Min.) (gpm) (days) (MW th) (days) (MW th) 1.0 24.44 5.67 173.5 10 10.42 2.42 74.0 2.0 20.02 4.64 142.1 11 10.05 2.33 71.4 3.0 17.25 4.00 122.5 12 9.72 2.26 69.0 3.5 16.19 3.76 114.9 13 9.43 2.19 67.0 4.0 15.30 3.55 108.6 14 9.16 2.13 65.0 4.5 14.54 3.37 103.2 15 8.92 2.07 63.3 5.0 13.88 3.22 98.5 16 8.70 2.02 61.8 5.5 13.31 3.09 94.5 17 8.48 1.97 60.2 6.0 12.83 2.98 91.1 18 8.29 1.92 58.9 6.5 12.39 2.87 88.0 19 8.10 1.88 57.5 7.0 12.01 2.79 85.3 20 7.93 1.84 56.3 7.5 11.67 2.71 82.9 25 7.15 1.66 50.8 8.0 11.37 2.64 80.7 30 6.53 1.51 46.4 8.5 11.10 2.58 78.8 40 5.59 1.30 39.7 9.0 10.85 2.52 77.0 50 4.92 1.14 34.9 9.5 10.62 2.46 75.4 80 3.76 0.87 26.7 Source of Data: SA-13-C00-1996-004

         ** The makeup flowrate listed is to compensate for boil off (not required flow to prevent boiling)

By : B.S. Blackmore Safety Analysis Operational Data SAOD Unit 1 Manual 3990 MWt Rev. 2 Reviewer Page 9 of 40 Ness Kilic TABLE 2.2.4 Time to Boil Following a Loss of SDC During Midloop Operation with A Large or Small Cold Leg Opening After Core Reload (3990 MW Core) Time after Time after Time to Boil (minutes) Time to Boil (minutes) Reactor Reactor Shutdown Shutdown Shutdown Cooling Heat Exchanger Inlet Temperature (F) Shutdown Cooling Heat Exchanger Inlet Temperature (F) (days) (days) 100 110 120 130 135 140 100 110 120 130 135 140 1.0 15.7 14.3 12.9 11.4 10.7 10.0 10 36.9 33.5 30.2 26.8 25.1 23.5 2.0 19.2 17.4 15.7 14.0 13.1 12.2 11 38.2 34.8 31.3 27.8 26.1 24.3 3.0 22.3 20.3 18.2 16.2 15.2 14.2 12 39.5 35.9 32.3 28.8 27.0 25.2 3.5 23.7 21.6 19.4 17.3 16.2 15.1 13 40.7 37.0 33.3 29.6 27.8 25.9 4.0 25.1 22.8 20.5 18.3 17.1 16.0 14 42.0 38.1 34.3 30.5 28.6 26.7 4.5 26.4 24.0 21.6 19.2 18.0 16.8 15 43.1 39.2 35.2 31.3 29.4 27.4 5.0 27.7 25.2 22.7 20.1 18.9 17.6 16 44.2 40.2 36.1 32.1 30.1 28.1 5.5 28.9 26.2 23.6 21.0 19.7 18.4 17 45.3 41.2 37.1 33.0 30.9 28.8 6.0 30.0 27.2 24.5 21.8 20.4 19.1 18 46.4 42.1 37.9 33.7 31.6 29.5 6.5 31.0 28.2 25.4 22.6 21.1 19.7 19 47.4 43.1 38.8 34.5 32.3 30.2 7.0 32.0 29.1 26.2 23.3 21.8 20.4 20 48.5 44.1 39.6 35.2 33.0 30.8 7.5 32.9 29.9 26.9 23.9 22.5 21.0 25 53.7 48.9 44.0 39.1 36.6 34.2 8.0 33.8 30.7 27.7 24.6 23.0 21.5 30 58.8 53.5 48.1 42.8 40.1 37.4 8.5 34.6 31.5 28.3 25.2 23.6 22.0 40 68.7 62.5 56.2 50.0 46.9 43.7 9.0 35.4 32.2 29.0 25.8 24.1 22.5 50 78.1 71.0 63.9 56.8 53.3 49.7 9.5 36.2 32.9 29.6 26.3 24.7 23.0 80 102.2 92.9 83.6 74.3 69.7 65.0 Current outage schedules do not support reloads in less than 10 days. Source of Data: SA-13-C00-1996-004

By: B.S. Blackmore Safety Analysis Operational Data SAOD Unit 1 Manual 3990 MWt Rev 2 Reviewer: Ness Kilic Page 19 of 40 TABLE 2.4.2 Key Reactor Core Parameters Following a Loss of SDC With the RCS Drained to the Reactor Vessel Flange Reactor Vessel Head On After Core Reload (3990 MW Core) Time after Decay Heatup Makeup Time after Decay Heatup Makeup Reactor Heat Rate Flowrate Reactor Heat Rate Flowrate Shutdown Load (F/Min.) (gpm)** Shutdown Load (F/Min.) (gpm) (days) (MWth) (days) (MWth) 1.0 19.06 4.42 135.3 10 8.13 1.89 57.7 2.0 15.62 3.62 110.9 11 7.84 1.82 55.7 3.0 13.46 3.12 95.5 12 7.58 1.76 53.8 3.5 12.63 2.93 89.7 13 7.36 1.71 52.2 4.0 11.93 2.77 84.7 14 7.14 1.66 50.7 4.5 11.34 2.63 80.5 15 6.96 1.61 49.4 5.0 10.83 2.51 76.9 16 6.79 1.57 48.2 5.5 10.38 2.41 73.7 17 6.61 1.53 47.0 6.0 10.01 2.32 71.1 18 6.47 1.50 45.9 6.5 9.66 2.24 68.6 19 6.32 1.47 44.9 7.0 9.37 2.17 66.5 20 6.19 1.44 43.9 7.5 9.10 2.11 64.6 25 5.58 1.29 39.6 8.0 8.87 2.06 63.0 30 5.09 1.18 36.2 8.5 8.66 2.01 61.5 40 4.36 1.01 31.0 9.0 8.46 1.96 60.1 50 3.84 0.89 27.2 9.5 8.28 1.92 58.8 80 2.93 0.68 20.8 Current outage schedules do not support reloads in less than 10 days. Source of Data: SA-13-C00-1996-004

         ** The makeup flowrate listed is to compensate for boil off (not required flow to prevent boiling)

By : B.S. Blackmore Safety Analysis Operational Data SAOD Unit 1 Manual 3990 MWt Rev. 2 Reviewer Page 20 of 40 Ness Kilic TABLE 2.4.3 Time to Boil Following a Loss of SDC with the RCS Drained to the Reactor Vessel Flange Reactor Vessel Head On Prior to Core Reload (3990 MW Core) Time after Time after Time to Boil (minutes) Time to Boil (minutes) Reactor Reactor Shutdown Shutdown Shutdown Cooling Heat Exchanger Inlet Temperature (F) Shutdown Cooling Heat Exchanger Inlet Temperature (F) (days) (days) 100 110 120 130 135 140 100 110 120 130 135 140 1.0 19.4 17.6 15.9 14.1 13.2 12.3 10 45.5 41.4 37.2 33.1 31.0 29.0 2.0 23.7 21.5 19.4 17.2 16.1 15.1 11 47.2 42.9 38.6 34.3 32.2 30.0 3.0 27.5 25.0 22.5 20.0 18.7 17.5 12 48.8 44.3 39.9 35.5 33.3 31.0 3.5 29.3 26.6 24.0 21.3 20.0 18.6 13 50.3 45.7 41.1 36.6 34.3 32.0 4.0 31.0 28.2 25.4 22.5 21.1 19.7 14 51.8 47.1 42.4 37.6 35.3 32.9 4.5 32.6 29.6 26.7 23.7 22.2 20.8 15 53.2 48.3 43.5 38.7 36.2 33.8 5.0 34.2 31.1 27.9 24.8 23.3 21.7 16 54.5 49.5 44.6 39.6 37.2 34.7 5.5 35.6 32.4 29.1 25.9 24.3 22.7 17 55.9 50.8 45.7 40.7 38.1 35.6 6.0 37.0 33.6 30.2 26.9 25.2 23.5 18 57.2 52.0 46.8 41.6 39.0 36.4 6.5 38.3 34.8 31.3 27.8 26.1 24.4 19 58.5 53.2 47.9 42.6 39.9 37.2 7.0 39.5 35.9 32.3 28.7 26.9 25.1 20 59.8 54.4 48.9 43.5 40.8 38.0 7.5 40.6 36.9 33.2 29.5 27.7 25.9 25 66.3 60.3 54.3 48.2 45.2 42.2 8.0 41.7 37.9 34.1 30.3 28.4 26.5 30 72.6 66.0 59.4 52.8 49.5 46.2 8.5 42.7 38.8 34.9 31.1 29.1 27.2 40 84.8 77.1 69.4 61.7 57.8 54.0 9.0 43.7 39.7 35.8 31.8 29.8 27.8 50 96.4 87.6 78.8 70.1 65.7 61.3 9.5 44.6 40.6 36.5 32.5 30.4 28.4 80 126.1 114.6 103.2 91.7 86.0 80.2 Source of Data: SA-13-C00-1996-004

ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Senior Reactor Operator NRC Exam 21. This Exam Level: SRO Appears on: SRO EXAM 2012 Tier 3 K/A #: 2.2.11 Importance 3.3 Rating: Which ONE of the following installations require a Temporary Modification? A. Alternate power supplied to NHN-M04 during a refueling outage. B. Domestic service flush line aligned to NCN-P01A while it is under clearance. C. Discharge pressure gauge on a LPSI pump while performing a surveillance test. D. Jumpers installed in an PPS channel while performing a troubleshooting work order. Answer: A Reference Id: Q1363 Difficulty: 3.00 Time to complete: 2 10CFR Category: CFR 55.43 (5) 55.43 (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. Cognitive Level: Memory Question Source: PV Bank Not Modified Comment: Proposed reference to be provided to applicant during examination: NONE Technical

Reference:

81DP-0DC17 (Temporary Modification Control) K&A: Equipment Control Knowledge of the process for controlling temporary changes. Learning Objective: L57327 Identify those plant changes that are NOT considered Temporary Modification. Justification: A. Correct: Per Appendix D of 81DP-0DC17, Temporary power installations connecting permanent plant equipment either bus, motor or valve, if the temporary power comes from one in-plant bus to another in-plant bus. B. Incorrect: Flushing a system while under clearance is similar to air assisted draining and does not require a Tmod.. C. Incorrect: LPSI ST pressure gauge has a permanently installed plant adapter for the ST and does not require a Tmod. D. Incorrect: This is controlled by the work control process and a Tmod is not required. REV 0

ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Senior Reactor Operator NRC Exam 22. This Exam Level: SRO Appears on: SRO EXAM 2007 SRO EXAM 2012 Tier 3 K/A #: 2.2.18 Importance 3.9 Rating: Given the following conditions: x Unit 1 is in a Midloop condition x Maintenance requests permission to re-lug ESFAS jumper leads Prior to this Work Order being released to the field, who (by title) is responsible to verify the proper RCS perturbation code? A. Releasing Organization and Outage Coordinator B. Releasing Organization and Operations Shift Manager C. Outage Coordinator and Midloop Operations Coordinator D. Midloop Operations Coordinator and Operations Shift Manager Answer: D Reference Id: Q10380 Difficulty: 4.00 Time to complete: 3 10CFR Category: CFR 55.43 (4) 55.43 (4) Radiation hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions. Cognitive Level: Memory Question Source: PV Bank Not Modified Comment: Proposed reference to be provided to applicant during examination: NONE Technical

Reference:

40OP-9ZZ16 (RCS Drain Ops) & 40OP-9ZZ20 (Reduced Inventory Ops) K&A: Knowledge of the process for managing maintenance activities during shutdown operations. Learning Objective: 30222 process for managing maintenance activities while shutdown Justification: A. Incorrect: The releasing organization and outage coordinator control clearances and other activities (making them seem correct), but not work orders. B. Incorrect:The releasing organization and outage coordinator control clearances and other activities (making them seem correct), but not work orders. C. Incorrect:The releasing organization and outage coordinator control clearances and other activities (making them seem correct), but not work orders. D. Correct: By procedure 40DP-9ZZ30 Appendix A, only these 2 control this activity. REV 0

ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Senior Reactor Operator NRC Exam 23. This Exam Level: SRO Appears on: SRO EXAM 2012 Tier 3 K/A #: 2.3.11 Importance 4.3 Rating: Given the following conditions: x A radioactive gas release permit is being written. x The release will be a routine, continuous release and will be less than 10% of any dose / dose rate ODCM requirement. Using the provided copy of Appendix J of 74RM-9EF20 (Gaseous Radioactive Release Permits and Offsite Dose Assessment), whose AUTHORIZATION (if any) is required for this release? A. RMS Technician. B. No authorization required. C. Control Room Supervisor/Shift Manager. D. Radiological Services Department Leader. Answer: B Reference Id: Q43918 Difficulty: 3.00 Time to complete: 3 10CFR Category: CFR 55.43 (4) 55.43 (4) Radiation hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions. Cognitive Level: Memory Question Source: New Comment: Proposed reference to be provided to applicant during examination: Copy of Appendix J of 74RM-9EF20. Technical

Reference:

74RM-9EF20 (Gaseous Radioactive Release Permits and Offsite Dose Assessment) K&A: Ability to control Radiation Releases Learning Objective: L82028 Given that a radioactive gaseous release is in progress. Identify Operations Department responsibilities In Accordance With 74RM-9EF20. Justification: A. Incorrect: RMS Technician can review and approve continuous release permits. Not Authorize. B. Correct: Per Note C. Authorization of Permits for routine continuous releases are not required. C. Incorrect: CRS/SM will authorize other types of permits per this appendix. D. Incorrect: Radiological Services DL will authorize other types of permits per this appendix. REV 0

GASEOUS RADIOACTIVE RELEASE PERMITS AND 74RM-9EF20 Page 75 of 83 Rev. 15 Appendix J Page 1 of 2 OFFSITE DOSE ASSESSMENT (Sample) Appendix J - Release Permit Review And Approval Matrix Release Description Level as% Radiological Vice Radiation Operations Radiation of Release of any Services CRS/Shift President Protection Department Protection Action Dose/Dose Department Manager Nuclear Supervision Leader Director Levels Rate ODCM Leader Production Requirement Less than or Equal to 50% of the Review and Dose/Dose Authorize Admin. Approval N/A N/A N/A N/A Rate < 40% (c) Dose/Dose (e) Rate Limit (a) Greater than 50% of but less than Dose/Dose Review and Acknowledge Acknowledge Authorize the Admin. Rate >40% Review N/A Approval (b) (b) (c) Dose/Dose and <80% Rate Limit (a) Greater than or equal to the Admin. Dose/Dose Review and Acknowledge Authorize Acknowledge Review Review Dose/Dose Rate > 80% Approval (b) (d) (b) Rate Limit (a)(f) NOTES

a. Applies to the quarterly and annual air and organ dose limits and instantaneous dose rate limits and not to the 31 day dose projection limits.

b. Acknowledgment requires that the appropriate individual be informed that the applicable dose/dose rate limit is being approached and that actions should be taken to reduce future releases. Acknowledgment should be obtained prior to release but can be obtained as soon as practical after the release.

c. Authorization of Permits for routine continuous releases are not required.

d. Under abnormal (emergency) conditions verbal approval for exceeding ODCM Requirement limits may be given by the CRS/Shift Manager when performing the release if it will bring the plant in to a safer condition. A notification to the NRC within one hour in accordance with 10CFR50.72 will be required after approval. If ODCM Requirement limits for dose are exceeded (ODCM sections 4.4a, 4.4b, 4.1a, 4.1b, 4.2a or 4.2b) comply with ODCM Requirement 5.1.

e. Continuous release permits meeting this requirement may be reviewed and approved by the RMS Technician.

f. The Plant Review Board shall review all Release Permits when an ODCM Requirement has actually been exceeded.

ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Senior Reactor Operator NRC Exam 24. This Exam Level: SRO Appears on: SRO EXAM 2012 Tier 3 K/A #: 2.4.38 Importance 4.4 Rating: Which ONE of the following is the lowest (least severe) Emergency Action Level that REQUIRES the EC to direct accountability, per the Emergency Plan? A. Unusual Event. B. Alert. C. Site Area Emergency. D. General Emergency. Answer: C Reference Id: Q8347 Difficulty: 2.00 Time to complete: 2 10CFR Category: CFR 55.43 (5) 55.43 (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. Cognitive Level: Memory Question Source: PV Bank Not Modified Comment: Proposed reference to be provided to applicant during examination: NONE Technical

Reference:

EP-0901 (ERO Position Checklists) K&A: Ability to take actions called for in the facility emergency plan, including supporting or acting as emergency coordinator if required. Learning Objective: L59732 Given an emergency event in progress, Determine if assembly and/or accountability are required. Justification: A. Incorrect: NUE requires the use of the ERO position checklist and may be chosen since it is the lowest of the EAL Classifications. B. Incorrect: Per step 12 App L of EP-0900,can be performed at Alert if DESIRED. C. Correct: Per step 6 of App L of EP-0900, Assembly/Accountability is only REQUIRED at SAE or higher. D. Incorrect: Assembly/Accountability is REQUIRED at GE, but it is not the lowest EAL Classification. REV 0

ES-401 Sample Written Examination Form ES 401 - 5 Question Worksheet PVNGS 2012 Senior Reactor Operator NRC Exam 25. This Exam Level: SRO Appears on: SRO EXAM 2012 Tier 3 K/A #: 2.4.40 Importance 4.5 Rating: Given the following conditions: x Unit 2 has declared a SITE AREA EMERGENCY. x The Unit 2 Shift Manager has been relieved as Emergency Coordinator (EC). Which ONE of the following positions must approve a PVNGS worker receiving Potassium Iodide (KI)? A. Unit 2 Shift Manager. B. Emergency Coordinator. C. Radiological Protection Monitor. D. Emergency Operations Director. Answer: B Reference Id: Q43919 Difficulty: 3.00 Time to complete: 2 10CFR Category: CFR 55.43 (4) 55.43 (4) Radiation hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions. Cognitive Level: Memory Question Source: PV Bank Not Modified Comment: Proposed reference to be provided to applicant during examination: NONE Technical

Reference:

EP-0905 (Protective Actions) K&A: Knowledge of SRO responsibilities in emergency plan implementation. Learning Objective: L92080 Identify the Emergency Coordinator's responsibilities associated with Emergency Exposure. Justification: A. Incorrect - If the Unit 2 SM was the EC this would be correct. SM also will direct plant operations during the event. EC controls AO movements. B. Correct - Per step 2.5 of EP-0905, the EC-STSC and EC-TSC are responsible for approving KI use by onsite emergency workers. C. Incorrect - the RPM is used to consult on such matters, but does not approve the dose. D. Incorrect - EOD will make many decisions during the event. Candidate may confuse EC with the EOD. REV0}}

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