ML091110225

Initial Examination 2009-301 Draft RO Written Exam

ML091110225
Person / Time
Site:	Oconee
Issue date:	03/13/2009
From:	NRC/RGN-II
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ER-09-301
Download: ML091110225 (150)

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2009 NRC REACTOR OPERATOR EXAM 1 POINT Question 1 Unit 1 initial conditions:

• Reactor power = 100%

Current conditions:

• Both Main FDWPs tripped

• Reactor has NOT tripped

• All Main Turbine Stop Valves are open

• The generator breakers are closed (PCB-20 & PCB-21)

Based on the current conditions, which ONE of the following would be the required operator actions?

A. Manually insert control rods and ensure both channels of AFIS have actuated B. Manually insert control rods and start ALL three HPI pumps and initiate Emergency Boration C. Manually trip the turbine and verify EFW has actuated and control EFW flow within limits D. Manually trip the turbine and start ALL three HPI pumps and initiate HPI Forced Cooling

2009 NRC REACTOR OPERATOR EXAM Question 1 T1/G1 - jmb 007EA2.02, Reactor Trip Stabilization - Recovery Ability to determine and interpret the following as they apply to a Reactor Trip:

Proper actions to be taken if the automatic safety functions have not taken place (4.3/4.6)

KIA MATCH ANALYSIS Requires knowledge of the actions directed in the EOP if the reactor fails to trip following a loss of Main Feedwater Pumps.

ANSWER CHOICE ANALYSIS Answer: C A. Incorrect: 1st part is incorrect. Plausible as manual rod insertion is directed by Rule 1 would be correct if Main FDW were available, however since Main FDW is not available the T/G should be tripped to allow Mod Temp & Doppler to aid in Rx shutdown to get within ECCS & EFDW capabilities. 2nd part is plausible as the turbine is not tripped and would result in an AFIS actuation if the reactor were tripped ..

B. Incorrect: Manual rod insertion is directed by Rule 1. Per Rule 1 only 2 HPI pumps are started (A or B HPI and C HPI pumps)

C. Correct: Per UNPP Tab if Main FDW is not available the turbine-generator is tripped and EFW is initiated. AMSAC should start EFW, operator would manually control EFW flow to control flow within pump and header limits.

D. Incorrect: 1st part is correct. Second part is incorrect. Per Rule 1 only 2 HPI pumps are started (A or B HPI and C HPI pumps)

Technical Reference(s): EP/1/A/1800/001 Rev 36 EOP IMAs, UNPP Tab and Rule 1 Proposed references to be provided to applicants during examination: None Learning Objective: EAP-IMA Obj. R4, EAP-UNPP Obj. R10 Question Source: M (EAP111 001)

Question History: Last NRC Exam _ _ _ __

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis

2009 NRC REACTOR OPERATOR EXAM 1 POINT Question 2 Unit 2 initial conditions:

• Reactor Power = 100% stable

• =

RCS Pressure 2045 psig decreasing slowly

• Pressurizer Temperature = 640°F decreasing slowly

• Quench Tank Level = 83 inches slowly rising

• =

Quench Tank Temperature 125°F slowly rising

• Quench Tank Pressure = 30 psig rising slowly Which ONE of the following describes the cause of the condition described above and the impact on departure from nucleate boiling ratio (DNBR)?

A. Pressurizer PORV leaking by / DNBR is moving further from limits B. Pressurizer PORV leaking by / DNBR is approaching limits C. Pressurizer Spray Valve leaking by / DNBR is moving further from limits D. Pressurizer Spray Valve leaking by / DNBR is approaching limits

2009 NRC REACTOR OPERATOR EXAM Question 2 T1/G1

' 00BAG2.2.22, Pressurizer Vapor Space Accident Knowledge of limiting conditions for operations and safety limits.

(4.0/4.7)

KIA MATCH ANALYSIS Requires knowledge of difference between indications of a PZR vapor space leak and spray flow and assess the impact of the reduced RCS pressure on DNBR (Reactor Core Safety Limit parameter)

ANSWER CHOICE ANALYSIS Answer: B A. Incorrect: Indications given are consistent with a leak into the Quench Tank from either the PORV or a leaky safety valve. DNBR is the ratio of CHF to Actual HF.

Lowering RCS pressure reduces CHF value and decreases DNBR. Plausible if definition of DNBR is inverted.

B. Correct: Indications given are consistent with a leak into the Quench Tank from either the PORV or a leaky safety valve. DNBR is the ratio of CHF to Actual HF. Lowering RCS pressure reduces CHF value and decreases DNBR.

C. Incorrect: Indications given are consistent with a leak into the Quench Tank from either the PORV or a leaky safety valve. Plausible as the pressure and PZR temperature reduction is also an indication of excess spray flow through the spray valve. However QT indication would not support it as a cause. Second part is plausible if the definition DNBR is inverted.

D. Incorrect: Indications given are consistent with a leak into the Quench Tank from either the PORV or a leaky safety valve. Plausible as the pressure and PZR temperature reduction is also an indication of excess spray flow through the spray valve. However QT indication would not support it as a cause. Second part is correct.

Technical Reference(s): PNS*PZR Rev 16a Proposed references to be provided to applicants during examination: None Learning Objective: PNS*PZR Obj R19, ADM*TSS Obj R1 Question Source: N Question History: Last NRC Exam _ _ _ __

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis

2009 NRC REACTOR OPERATOR EXAM 1 POINT Question 3 Which ONE of the following combinations of parameters describes the indications of establishing boiler-condenser mode heat transfer following a SBLOCA?

ASSUME SG LEVELs ARE AT THE LOSCM SETPOINT AND TeVs CLOSED RCS Hot Leg water level is _ _ _ / SG Pressures are _ __

A. below the SG water level/increasing B. below the SG water level/decreasing C. above the SG upper tube sheet / increasing D. above the SG upper tube sheet / decreasing

2009 NRC REACTOR OPERATOR EXAM Question 3 T1/G1 - okm 009EK1.01, Small Break LOCA Knowledge of the operational implications of the following concepts as they apply to the small break LOCA: Natural circulation and cooling, including reflux boiling (4.2/4.7)

KIA MATCH ANALYSIS Requires knowledge of the plant conditions required for boiler-condenser cooling (reflux boiling) during a SBLOCA ANSWER CHOICE ANALYSIS Answer: A A. Correct: Hot Leg level must be established at some level below the SG secondary side level to allow condensation of primary side steam. When BCM is established SG pressure will increase due the temperature increase in the SG as heat is transferred into the SG.

B. Incorrect: Hot Leg level is correct. Hot Leg level is plausible in that level being above the upper tube sheet level exists when transitioning from sustained Two-Phase natural circulation flow towards BCM flow. BCM requires Hot Leg level to be voided much lower.

C. Incorrect: Hot Leg level is plausible in that level being above the upper tube sheet level exists when transitioning from sustained Two-Phase natural circulation flow towards BCM flow. BCM requires Hot Leg level to be voided much lower. SG pressure response is correct.

D. Incorrect: Hot Leg level is plausible in that level being above the upper tube sheet level exists when transitioning from sustained Two-Phase natural circulation flow towards BCM flow. BCM requires Hot Leg level to be voided much lower. Hot Leg level is plausible in that level being above the upper tube sheet level exists when transitioning from sustained Two-Phase natural circulation flow towards BCM flow.

BCM requires Hot Leg level to be voided much lower.

Technical Reference(s): TA-AM1 Proposed references to be provided to applicants during examination: None Learning Objective: TA-AM1 Obj. R16 Question Source: N Question History: Last NRC Exam _ _ _ __

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis

2009 NRC REACTOR OPERATOR EXAM 1 POINT Question 4 Unit 1 conditions:

• ReS Pressure = 200 psig deceasing

• HPI Train A Flow =995 gpm

• HPI Train B Flow =490 gpm Based on the conditions above, which ONE of the following describes the required operator actions to protect the HPI pumps?

A. Throttle BOTH Train A & B HPI flows to <475 gpm per pump B. Throttle ONLY Train A HPI flow to <750 gpm C. Throttle BOTH Train A & B HPI flow to <950 gpm combined D. Throttle ONLY Train B HPI flow to <475 gpm

2009 NRC REACTOR OPERATOR EXAM Question 4 T1/G1 - jmb 011 EK2.02, Large Break LOCA Knowledge of the interrelations between the Large Break LOCA and the following: Pumps (2.6/2.7)

KIA MATCH ANALYSIS Requires knowledge of relationship between HPI pump status and flow to determine required HPI pump throttling criteria to ensure pump operation within limits and core cooling is maintained.

ANSWER CHOICE ANALYSIS Answer: D A. Incorrect. Flow is acceptable in the A header due to 2 pumps operating aligned to the header. B Header flow requires throttling to <475 gpm per Rule 6. Plausible as the direction is correct if only one HPI pump is supplying the A header.

B. Incorrect. Plausible as this is the value of total flow in Rule 6 when operating HPI in piggyback mode with either only one LPI pump running or only one piggyback valve open.

C. Incorrect: Plausible as this is the value in Rule 6 if only HPI A & B operating with HP-409 open.

D. Correct: Flow is above 475 flow limit and throttling is required per Rule 6 Technical Reference(s): EP/1/A/1800/001 Rev 36 (Rule 6), OMP 1-18 Proposed references to be provided to applicants during examination: None Learning Objective: ADM-OMP Obj. R10, 52, EAP-LOSCM Obj R16 Question Source: N Question History: Last NRC Exam _ _ _ __

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis

2009 NRC REACTOR OPERATOR EXAM 1 POINT Question 5 Unit 1 initial conditions:

• Reactor power = 65%

• 1 LPSW-6 (UNIT 1 RCP COOLERS SUPPLY) fails closed Current conditions:

• AP/24 (Loss of LPSW) in progress

• RCP Temperatures: (All temperatures slowly increasing) 1A1 1A2 1B2 Upper Guide 182°F 197°F 185° Bearing Temp Radial Bearing 219°F Temp Based on the above conditions, which ONE of the following is the action directed per AP/24?

A. Manually trip the Reactor and stop ALL RCPs B. Manually trip the Reactor and stop RCPs 1A2 & 1B1 ONLY

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\ C. Stop RCP 1A2 ONLY and verify FDW re-ratios properly D. Stop RCP 1 B1 ONLY and verify FDW re-ratios properly

2009 NRC REACTOR OPERATOR EXAM Question 5 T1/G1 -1mb 015AA2.02, Reactor Coolant Pump Malfunctions

(

Ability to determine and interpret the following as they apply to the Reactor Coolant Pump Malfunctions (Loss of RC Flow): Abnormalities in RCP air vent flow paths and/or oil cooling system (2.8/3.0)

KIA MATCH ANALYSIS Requires the ability to determine impact of loss of LPSW on RCP bearing oil cooling temperatures and action required as a result of the loss ANSWER CHOICE ANALYSIS Answer: A A. CORRECT: Both RCP 1A2 & 1B1 are above immediate trip criteria, however AP/24 recognizes the impending heating of all the RCPs so it directs tripping the reactor and then tripping all the RCPs.

B. Incorrect - Plausible because AP/16 directs that if any RCP meets immediate trip criteria and less than 3 RCPs will be remain, then manually trip the Rx and immediately stop the affected RCPs only C. Incorrect: Plausible in that failure to recognize that 1B1 is also above trip criteria would result in this selection which is directed by AP/16. If only one RCP is tripped below 70% power Rx trip is not required and FDW reratio is verified.

D. Incorrect: Plausible in that failure to recognize that 1A2 is also above trip criteria would result in this selection which is directed by AP/16. If only one RCP is tripped below 70% power Rx trip is not required and FDW reratio is verified.

Technical Reference(s): AP/16 Rev 26, AP/24 Rev 23 Proposed references to be provided to applicants during examination: None Learning Objective: EAP-APG Obj. R9 Question Source: M (EAP21 0983)

Question History: Last NRC Exam _ _ _ __

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis

2009 NRC REACTOR OPERATOR EXAM 1 POINT Question 6 Unit 1 plant conditions:

Time =0400

• Reactor power =100%

• 1A HPI Pump operating

• 1B HPI Pump in AUTO Time =0401

• 1A HPI Pump amps cycling

• 1SA2/C-2 (HP Injection Pump Disch Header Pressure Low) actuated Time =0405

• AP/14 (Loss Of Normal HPI Makeup and/or RCP Seal Injection) in progress

• All HPI pumps are OFF Based on the conditions above, which ONE of the following describes the pump used to restore RCS makeup and the suction source used as directed by AP/14?

A. 1B HPI pump / BWST B. 1B HPI pump / LOST C. 1C HPI Pump / BWST D. 1C HPI Pump / LOST

2009 NRC REACTOR OPERATOR EXAM Question 6 T1/G1 - jmb 022AA2.02, Loss of Reactor Coolant Makeup Ability to determine and interpret the following as they apply to the Loss of Reactor Coolant Makeup: Charging pump problems (3.2/3.7)

KIA MATCH ANALYSIS Requires ability to determine the preferred HPI pump used to restore RCS makeup following indication of a loss of suction to the operating HPI pump and and the suction source used to restore the pump.

ANSWER CHOICE ANALYSIS Answer: C A. Incorrect: The 1 B HPI pump is plausible as it is the standby pump and would be the pump started if the running pump had not lost suction. The water source is correct.

B. Incorrect: The 1B HPI pump is plausible as it is the standby pump and would be the pump started if the running pump had not lost suction. The water source is plausible as it is the normal source and would be selected if step 4.30 is misread or misinterpreted.

C. Correct: 1C HPI Pump is the preferred pump per step 4.27. The BWST is directed to be used as a suction source based on indication that the running HPI pump lost suction. (Step 4.30).

D. Incorrect: The pump selected is correct. The water source is plausible as it is the normal source and would be selected if step 4.30 is misread or misinterpreted.

Technical Reference(s): AP/14 Rev. 16 Proposed references to be provided to applicants during examination: None Learning Objective: EAP-APG Obj R9 Question Source: M (EAP210922)

Question History: Last NRC Exam _ _ _ __

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis

2009 NRC REACTOR OPERATOR EXAM 1 POINT Question 7 Unit 1 plant conditions:

• Load Shed has occurred concurrent with a LBLOCA

• 1A and 1B LPI pump failed to start Based on the above conditions, which ONE of the following describes actions to restore LPI flow in accordance with EOP Enclosure 5.1 (ES Actuation)?

Manual reset of Load Shed is ...

A. NOT required / Align flow path and start 1C LPI Pump B. required / Align flow path and start 1C LPI Pump C. NOT required / Start 1C LPI Pump immediately D. required / Start 1C LPI Pump immediately after Load Shed is reset

2009 NRC REACTOR OPERATOR EXAM Question 7 T1/G1 - jmb 025AK1.01, Loss of Residual Heat Removal System Knowledge of the operational implications of the following concepts as they apply to Loss of Residual Heat Removal System: Loss of RHRS during all modes of operation (3.9/4.3)

KIA MATCH ANALYSIS Requires knowledge of actions required to restore core decay heat removal following a failure of the LPI/DHR Pumps ANSWER CHOICE ANALYSIS Answer: A A. Correct: Pushing the Control Room MFB monitor RESET push buttons is not required because the signal for the 1C LPI Pump is removed 5 seconds after the Load Shed actuated. Ene!. 5.1 directs manually aligning and starting 1C LPI Pump.

B. Incorrect: First part is incorrect. Plausible because reset is required for other components load shed. Second part is correct.

C. Incorrect: First part is correct. Second part is incorrect. Starting the 1C pump immediately without aligning the flowpath will not result in flow to the core.

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D. Incorrect: First part is incorrect. Plausible because reset is required for other components load shed. Second part is incorrect. Starting the 1C pump immediately without aligning the flowpath will not result in flow to the core.

Technical Reference(s): EP/1/1800/001 Rev 36 Ene!. 5.1 Proposed references to be provided to applicants during examination: None Learning Objective: EAP*ESA Obj. R17, EL*PSL Obj. R6 Question Source: N Question History: Last NRC Exam _ _ _ __

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis

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2009 NRC REACTOR OPERATOR EXAM 1 POINT Question 8 Unit 1 initial conditions:

• Reactor power = 100%

Current conditions:

• RCS pressure (actual) =2255 psig increasing

• 1SA-2/D-3 (RC PRESS HIGH/LOW) in alarm

• 1RC-1 (pzr Spray) indicates CLOSED

• 1RC-66 (PORV) indicates CLOSED

• ALL Pressurizer Heaters indicate ON

• Reactor does not trip Which ONE of the following describes the cause of the above indications and the component(s) which would limit the RCS pressure increase?

ASSUME NO OPERATOR ACTION

_ _ _ has failed LOW and RCS pressure would be automatically limited by _ __

A. Selected RCS Wide Range Pressure / PZR Code Safety Valves B. Selected RCS Wide Range Pressure / the PZR PORV

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C. Controlling RCS Narrow Range Pressure / PZR Code Safety Valves D. Controlling RCS Narrow Range Pressure / the PZR PORV

2009 NRC REACTOR OPERATOR EXAM Question 8 T1/G1 - jmb 027 AK2.03, Pressurizer Pressure Control Malfunction Knowledge of the interrelations between the Pressurizer Pressure Control and the following: Controllers and Positioners (2.6/2.8)

KIA MATCH ANALYSIS Requires knowledge of impact of RCS NR Pressure failure on automatic operation of the controllers for the Pzr heaters, Spray Valve and PORV.

ANSWER CHOICE ANALYSIS Answer: C A. Incorrect: Plausible because WR Pressure is used for ES but not for RPS and Pressure control. Second part is correct.

B. Incorrect: Plausible because WR Pressure is used for ES but not for RPS and Pressure control. Second part is incorrect. Plausible because the PORV is actuated by a separate input when selected to LOW vs HIGH.

C. Correct: Failure of the controlling NR Pressure signal will energize the pressurizer heaters and prevent automatic operation of both the spray valve and PORV. Pressure will increase to the PZR safety valve setpoint if not stopped manually.

D. Incorrect: First part is correct. Second part is incorrect. Plausible because the PORV is actuated by a separate input when selected to LOW vs HIGH.

Technical Reference(s): AP/44 Rev. 00 (Draft), AP/28 Rev. 14 Proposed references to be provided to applicants during examination: None Learning Objective: IC-RCI Obj R6, EAP-APG Obj R9 Question Source: N Question History: Last NRC Exam _ _ _ __

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis

2009 NRC REACTOR OPERATOR EXAM 1 POINT Question 9 Unit 3 initial conditions:

• =

Reactor Power 95% decreasing

• =

RCS Pressure 2455 psig increasing

• =

Tavg 598°F increasing

• =

Pressurizer level 355 inches increasing Current conditions:

=

• Reactor Power 4% decreasing

=

• ReS Pressure 2735 psig (maximum observed) decreasing

=

• Tavg 596°F decreasing

• Pressurizer level = 365 inches decreasing Based on the conditions above, which ONE of the following describes whether of not the RCS Pressure safety limit has been exceeded and the bases for safety limit?

RCS Pressure Safety Limit has _ _ _ _ _ . This Safety Limit _ _ _ _ __

A. been exceeded / ensures RCS pressure is maintained below 110% of design pressure to prevent RCS pressure boundary failure B. been exceeded / ensures that RCS pressure remains in the assumed range used in the analysis for reactivity accidents including slow rod withdrawal C. NOT been exceeded / ensures RCS pressure is maintained below 110% of design pressure to prevent RCS pressure boundary failure D. NOT been exceeded / ensures that RCS pressure remains in the assumed range used in the analysis for reactivity accidents including slow rod withdrawal

2009 NRC REACTOR OPERATOR EXAM Question 9 T1/G1 - jmb 029EG2.2.25, Anticipated Transient Without Scram (ATWS)

Knowledge of the bases in Technical Specifications for limiting conditions for operations and safety limits.

(3.2/4.2)

KIA MATCH ANALYSIS Requires knowledge of RCS pressure status in relation to the RCS Pressure Safety Limit during an ATWS event and bases for safety limit ANSWER CHOICE ANALYSIS Answer: C A. Incorrect: RCS pressure has not exceeded 2750 psig safety limit value. Second part is correct.

B. Incorrect: RCS pressure has not exceeded 2750 psig safety limit value. Second part is incorrect. Plausible as the basis for the RPS high pressure trip set point.

C. Correct: RCS pressure given as maximum observed is below the safety limit value. The limit is based on ensuring that RCS integrity is maintained by assuring that RCS pressure is less than 110% of design pressure value of 2500 psig.

D. Incorrect: First part is correct. Second part is incorrect. Second part is incorrect.

Plausible as the basis for the RPS high pressure trip set point.

Technical Reference(s): Tech Spec 2.1.1.2 and Bases Proposed references to be provided to applicants during examination: None Learning Objective: ADM-TSS Obj 4 & 5 Question Source: N Question History: Last NRC Exam _ _ _ _ __

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis

2009 NRC REACTOR OPERATOR EXAM 1 POINT Question 10 Unit 1 plant conditions:

• Primary to secondary leakage in 1A SG

• =

Reactor Power 29% decreasing

• Pressurizer level = 160 inches decreasing

• ALL HPI Pumps running

• 1HP-5 closed Based on the conditions above, which ONE of the following describes the method used to determine the SG Tube Leak Rate and what method will be used to shutdown the reactor per the EOP SGTR Tab?

RCS Leak rate is determined by ...

A. monitoring 1 RIA-59 & 1 RIA-60 / Continue controlled shutdown B. monitoring 1 RIA-59 & 1RIA-60 / Trip the Reactor C. performing an RCS inventory balance / Continue controlled shutdown D. performing an RCS inventory balance / Trip the Reactor

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2009 NRC REACTOR OPERATOR EXAM Question 10 T1/G1 - jmb 038EK3.06, Steam Generator Tube Rupture Knowledge of the reasons for the following responses as they apply to the SGTR: Actions contained in EOP for RCS water inventory balance, S/G tube rupture, and plant shutdown procedures (4.2/4.5)

KIA MATCH ANALYSIS Requires knowledge of the method used to determine RCS leak rate in the SGTR EOP and the method of shutdown used based on power level and leak rate, ANSWER CHOICE ANALYSIS Answer: D A. Incorrect: Plausible in that 1RIA-59 &-60 (MS Line N-16 gamma detectors) are accurate above 40% power. Below 40% they provide a trend but can not be used to determine leakrate. Also the EOP SGTR Tab is the correct procedural guidance due to the size of the rupture.

B. Incorrect: Plausible in that 1RIA-59 &-60 (MS Line N-16 gamma detectors) are accurate above 40% power. Below 40% they provide a trend but can not be used to determine leakrate. Also, AP/31 would be correct if the tube leakage were <25 gpm.

C. Incorrect: Plausible in that inventory balance is used to estimate leakrate; however AP/31 is wrong but it would be correct if the tube leakage were <25 gpm.

D. Correct: Inventory balance is used to estimate leakrate and the EOP SGTR Tab is the correct procedural guidance the when the leak is this large.

Technical Reference(s): EAP*SGTR, EP/1/1800/01 SGTR Tab Proposed references to be provided to applicants during examination: None Learning Objective: EAP*SGTR Obj R1, R2 Question Source: N Question History: Last NRC Exam _ _ _ __

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis

2009 NRC REACTOR OPERATOR EXAM 1 POINT Question 11 Unit 1 conditions:

• Reactor Tripped

• Main Turbine Control Valves are ALL closed

• Main Turbine Stop Valves 1 & 3 (1 MS-1 05 & 103) are both open Based on the conditions above, which ONE of the following describes the action required (if any) and the reason action is (or is not) required?

A. No action is required / ALL Control Valves being closed isolates the Main Steam supply to the Main Turbine and prevents a post trip overcooling due to excessive steam flow.

B. No action is required / ALL Control Valves being closed isolates the Main Steam supply to the Main Turbine to prevent turbine damage from overspeeding following a generator trip from 100% power.

C. Place both EHC Pumps in PULL TO LOCK / ALL Stop Valves must be closed to ensure the Main Steam Lines are isolated from each other to prevent a main steam line break from effecting both SGs.

D. Place both EHC Pumps in PULL TO LOCK / ALL Stop Valves must be closed to ensure the Main Steam supply to the Main Turbine steam chest is isolated to prevent turbine damage from overspeeding following a generator trip from 100%

power.

2009 NRC REACTOR OPERATOR EXAM Question 11 T1/G1 040AK3.01, Steam Line Rupture Knowledge of the reasons for the following responses as they apply to the Steam Line Rupture: Operation of steam line isolation valves (4.2/4.5)

KIA MATCH ANALYSIS Requires knowledge of the reason the Turbine Stop Valves are required to be closed following a Reactor/Turbine Trip Per discussion with NRC Lead Examiner - substituted Turbine Stop Valves (TSV's) for MSIV's and kept KIA. TSVs perform similar function of MSIVs for Steam Line Rupture.

ANSWER CHOICE ANALYSIS Answer: C A. Incorrect: Plausible as the control valves isolate steam flow to the turbine and would prevent a post trip overcooling from steam flow to the turbine.

B. Incorrect: Plausible as isolating the steam supply following a loss of load would minimize the possibility of exceeding turbine speed ratings that could lead to damage.

C. Correct: EOP IMA require the Turbine Stop Valves (TSVs) to be closed. The required action is to place the EHC Pumps in PTL. The TSVs serve as the isolation valves between the OTSGs in the RB and the Main Turbine. All four TSVs are needed. If for example a MSLB occurred, both S/Gs would discharge through the break by way of the steam chest. Once the TSVs closed, then the steam loss would be isolated to just the single affected steam generator I MS line.

D. Incorrect: the action is correct however the reason is incorrect. The reason is plausible as isolating the steam supply following a loss of load would minimize the possibility of exceeding turbine speed ratings that could lead to damage.

Technical Reference(s): EP/1/1800/001 EOP IMAs, TS Bases 3.3.15 & 3.7.2 Proposed references to be provided to applicants during examination: None Learning Objective: STG-MT Obj. R3, EAP-IMA Obj. R5 Question Source: N Question History: Last NRC Exam _ _ _ __

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis

2009 NRC REACTOR OPERATOR EXAM 1 POINT Question 12

( Unit 1 initial conditions:

• Reactor power 100% =

Current conditions:

• Station Blackout

• RCS Temperature 2 minutes after trip

- Tc = 550°F

- Th = 556°F

- CETCs 558°F =

=

• SG Pressures 1010 psig stable Based on the conditions above, which ONE of the following describes the response of the RCS heat removal parameters over the next 5 minutes during the establishing of natural circulation?

ASSUME POWER HAS NOT BEEN RESTORED RCS Tcold RCS Thot CETCs A. Stable Stable Increasing B. Decreasing Stable Stable C. Stable Increasing Increasing D. Decreasing Increasing Stable

2009 NRC REACTOR OPERATOR EXAM Question 12 T1/G1 jmb 055EA1.01, Station Blackout Ability to operate and monitor the following as they apply to a Station Blackout:

In-core thermocouple temperatures (3.7/3.9)

KIA MATCH ANALYSIS Requires the ability to determine the status of RCS heat removal based on the relationship between RCS Loop Temperatures and CETC temperature ANSWER CHOICE ANALYSIS Answer: C A. Incorrect: Plausible because initially Thot would be relatively stable but would be follow the trend of CETC temp. Tcold and CETC response are correct.

B. Incorrect: Thot and CETC response would be correct if at low decay heat levels or natural circ had developed. Tcold response is incorrect as it would be held constant by the SG Pressure/Temp. Tcold response is plausible as it would be the response to HPI flow in the cold legs.

C. Correct: In the 5 minute period following the trip natural circulation conditions will be developing. Thot and CETC will be increasing with Tcold being held constant by SG pressures in order to build in an adequate thermal driving head to establish flow. After flow is established CETC & Thot will stabilize and eventually decrease as either decay heat level drops off or SG pressures are reduced.

D. Incorrect: This combination is a variation of the indications as natural circ is interrupted by cold HPI flow into the cold legs.

Technical Reference(s): TA-AM1 Rev OBc Proposed references to be provided to applicants during examination: None Learning Objective: TA-AM1 Obj R3 Question Source: N Question History: Last NRC Exam _ _ _ __

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis

2009 NRC REACTOR OPERATOR EXAM 1 POINT Question 13 Unit 1 initial conditions:

• LBLOCA occurred 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> ago

=

• RCS Pressure 30 psig

• 1A & 1B LPI Pumps are running Current conditions:

• 1KVIA is deenergized

• 1SA-18/A-3 RVLlS/ICCM/RG1.97 Train A Trouble actuated Based on the conditions above, which ONE of the following describes the impact on the LPI system instrumentation and what alternate indication can be used to determine the status of the LPI pumps?

A. LPI HDR 1A INJ FLOW (gpm) is Blank I 1A LPI Pump amps and breaker indicating lights B. LPI HDR 1A INJ FLOW (gpm) is Blank I 1A LPI HDR flow computer point (OAC)

C. LPI HDR 1B INJ FLOW (gpm) is Blank I 1B LPI Pump amps and breaker indicating lights D. LPI HDR 1B INJ FLOW (gpm) is Blank I 1B LPI HDR flow computer point (OAC)

2009 NRC REACTOR OPERATOR EXAM Question 13 T1/G1 jmb 057AA1.05, Loss of Vital AC Instrument Bus Ability to operate and 1 or monitor the following as they apply to the Loss of Vital AC Instrument Bus: Backup instrument indications (3.2/3.4)

KIA MATCH ANALYSIS Requires ability to assess of LPI pump operating status and correlate pump flow to operating current (amps) and Cooler L1 T following a loss of an instrument bus (KVIA).

ANSWER CHOICE ANALYSIS Answer: A A. Correct: The loss of 1KVIA de-energizes Train A ICCM and results in LPI Train A Injection Flow dixon indicator failing. Pump Amps are not powered from ICCM and remain normal (approx 53-54 amps) and cooler.aT can be calculated using pump inlet temp and cooler outlet temp which are not powered by ICCM.

B. Incorrect: First part is correct. The loss of 1KVIA de-energizes Train A ICCM and results in LPI Train A Injection Flow dixon indicator failing. Second part is plausible if assumption is made that the OAC is not dependent on ICCM.

C. Incorrect: First part is incorrect. Plausible as it is the indication lost if KVIB were de-energized. Failure to apply the correct train power supply would result in this selection. Second part is correct. This indication is available and is also not impacted by ICCM.

D. Incorrect: First part is incorrect. Plausible as it is the indication lost if KVIB were de-energized. Failure to apply the correct train power supply would result in this selection. Second part would be correct and is not impacted by train A. Plausible if assumption is made that the OAC is not dependent on ICCM.for the train selected.

Technical Reference(s): OP/1/A/1105/012, OP/1/A16101/005, IC-RCI Proposed references to be provided to applicants during examination: None Learning Objective: IC-RCI Obj. R 43 & 59 Question Source: N Question History: Last NRC Exam _ _ _ __

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis

2009 NRC REACTOR OPERATOR EXAM 1 POINT Question 14 Unit 1 initial conditions:

• Reactor power = 100%

• 1SA-5/A-11 (EL DC System 1CA Trouble) actuates Current conditions:

• NEO standing in front of 1CA Charger observes the charger voltage increases to

-150 volts THEN drops to 0 volts Based on the above conditions, which ONE of the following describes the status of the 1CA Charger and the 1DCA Bus?

1CA Charger _ _ _ breaker automatically tripped and 1DCA Bus is being powered from - -

A. AC Input / 1CA Battery B. AC Input / the alternate unit's CA Battery C. DC Output / 1CA Battery D. DC Output / the alternate unit's CA Battery

2009 NRC REACTOR OPERATOR EXAM Question 14 T1/G1okm 058AK1.01, Loss of DC Power Knowledge of the operational implications of the following concepts as they apply to Loss of DC Power: Battery charger equipment and instrumentation (2.8/3.1)

KIA MATCH ANALYSIS Requires knowledge of the indications of a failing battery charger and the operational implications of the loss of a Vital DC Battery Charger to the Vital DC system ANSWER CHOICE ANALYSIS Answer: A A. Correct: Requires knowledge of the battery charger output over-voltage interlock at ~ 147 volts for 5 secs that trips the charger AC Input breaker. Loss of the charger (the normal source) will not result in loss of power to the bus due to the control battery floating on the bus.

B. Incorrect: Knowledge of the battery charger output over-voltage interlock at ~ 147 volts for 5 secs that trips the charger AC Input breaker is correct. Second part is incorrect but plausible if the DCA Bus is confused with the DC Panelboards immediately down-stream of the DCA Bus.

C. Incorrect: Plausible in that the over-voltage interlock (~ 147 volts for 5 secs) trip of the charger AC Input breaker could be applied to the DC output breaker. Second part is correct in that 1CA Battery will supply 1DCA Bus without a loss of power D. Incorrect: Plausible in that the over-voltage interlock (~ 147 volts for 5 secs) trip of the charger AC Input breaker could be applied to the DC output breaker. Second part is plausible if the DCA Bus is confused with the DC Panel boards immediately down-stream of the DCA Bus Technical Reference(s): 1SA-12 C1 Rev 02 Proposed references to be provided to applicants during examination: None Learning Objective: EL-DCD R2, R4 Question Source: N Question History: Last NRC Exam _ _ _ __

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis

2009 NRC REACTOR OPERATOR EXAM 1 POINT Question 15 Unit 3 initial conditions:

• Reactor power = 100%

Current conditions:

• LPSW header pressure =20 psig increasing Based on the above conditions, which ONE of the following describes the status of statalarm 3SA-9/C3 (LPSW Low Press RB Aux Cooler Isolation) and if the RBACs isolate, how is LPSW flow restored?

A. Actuated / Automatically when LPSW pressure returns above setpoint B. Actuated / Manually after depressing LPSW LOW PRESS DIG CH 1 AND 2 push buttons C. Not Actuated / Automatically when LPSW pressure returns above setpoint D. Not Actuated / Manually after depressing LPSW LOW PRESS DIG CH 1 AND 2 push buttons

(

2009 NRC REACTOR OPERATOR EXAM Question 15 T1/G1 jmb 062AG2.4.46, Loss of Nuclear Service Water

( Ability to verify that the alarms are consistent with the plant conditions.

(4.2/4.2)

K/A MATCH ANALYSIS Requires ability to evaluate alarm status based on plant conditions and determine status of LPSW to the RB Aux Coolers.

ANSWER CHOICE ANALYSIS Answer: B A. Incorrect: Plausible in that the alarm status is correct; alarm setpoint was reached at 23 psig decreasing and pressure must rise to > 23 psig in order for the S/A to automatically clear. RBACs did isolate; however rising LPSW header pressure above 23 psig will not automatically restore flow to the RBACs.

B. CORRECT: Alarm status is correct; alarm setpoint was reached at 23 psig decreasing and pressure must rise> 23 psig in order for the S/A to automatically clear. RBACs did isolate and they must be manually restored to their normal OPEN/AUTO status after depressing both LPSW LOW PRESS DIG CH 1 AND 2 pushbuttons.

C. Incorrect: Plausible except that the Alarm status is incorrect; alarm setpoint was reached at 23 psig decreasing and pressure must rise to > 23 psig in order for the S/A to automatically clear. RBACs did isolate; however rising LPSW header pressure above 23 psig will not automatically restore flow to the RBACs.

D. Incorrect: Plausible except that the Alarm status is incorrect; alarm setpoint was reached at 23 psig decreasing and pressure must rise to > 23 psig in order for the S/A to automatically clear. RBACs did isolate and they must be manually restored to their normal OPEN/AUTO status after depressing both LPSW LOW PRESS DIG CH 1 AND 2 push buttons.

Technical Reference(s): Statalarm 3SA-9/C3 (LPSW Low Press RB Aux. Cooler Isolation)

Proposed references to be provided to applicants during examination: None Learning Objective: SSS-LPW Obj R13, PNS-RBC Obj. R14 Question Source: N Question History: Last NRC Exam _ _ _ __

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis

2009 NRC REACTOR OPERATOR EXAM 1 POINT Question 16

( Unit 1 initial conditions:

• Reactor power 100% =

• Instrument Air Pressure decreasing

• AP/22 (Loss of Instrument Air) initiated Current conditions:

=

• Instrument Air pressure 61 psig decreasing

• FDW Pump L\P OAC alarms actuate

• 1A & 1B Main FDW Pump speeds are both increasing Based on the above conditions, which ONE of the following describes the actions required by AP/22?

A. Commence a plant shutdown. If at any time two or more CRD temperatures are

>180 oF, then trip the reactor.

B. Commence a plant shutdown. If at any time SG level approaches main FDW pump trip criteria, then trip the reactor.

C. Manually trip the reactor. Manually trip both main FDW pumps.

D. Manually trip the reactor. Take both FDW Masters to Hand and decrease demand to zero.

2009 NRC REACTOR OPERATOR EXAM Question 16 T1/G1- OKM 065AA1.05, Loss of Instrument Air Ability to operate and I or monitor the following as they apply to the Loss of Instrument Air: RPS (3.3/3.3)

KIA MATCH ANALYSIS Question tests knowledge of when to trip the reactor during a loss of IA event.

Therefore, the question tests knowledge of the ability to operate the RPS, via tripping reactor, when a loss of instrument air occurs.

ANSWER CHOICE ANALYSIS Answer: C A. Incorrect: Plausible in that a rapid unit shutdown is required; however the method directed is a manual Rx trip. Second part is plausible as it is the immediate trip criteria for loss of CC flow to the CRDMs during the loss of IA is valid.

B. Incorrect: Plausible in that a rapid unit shutdown is required; however the method directed is a manual Rx trip. Second part is plausible in that OMP 1-18 dictates a Manual Rx Trip and tripping of both MFWPS if any SG reaches >96% on the OR level.

C. Correct. AP/22 requires the reactor to be tripped when FDW is not controllable. The OAC alarm actuates at about 30 psig, well below the -65 psig where FDW valves can stop responding to control signals. Applicants need to know when the OAC alarm actuates. (See referenced SAEL and Step 4.3 of the AP) Therefore, the AP requires that the reactor be tripped and the MFDW pumps to be tripped.

D. Incorrect: AP/22 requires MFDW pumps to be tripped immediately after the reactor is manually due to loss of FDW controllability. Plausible in that the candidate could erroneously think that Feedwater control valves (and FDW demand) would still be controllable if taken to Hand on the ICS stations.

Technical Reference(s): AP/22 Rev 25 (Loss of Instrument Air), Lesson Plan SSS-lA, Instrument Air System, Rev. 18, Simulator Guide, SAE-L 035, Loss of Instrument Air, Rev. 9.

Proposed references to be provided to applicants during examination: None Learning Objective: SSS-IA R44, R45, R53 Question Source: B Question History: Last NRC Exam ONS 2006 RO Exam Q#62 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis

2009 NRC REACTOR OPERATOR EXAM 1 POINT Question 17

( Unit 3 initial conditions:

• =

Reactor power 100%

• AP/34 (Degraded Grid) in progress

• =

Generator Output (+) 900 MWe

(-) 390 MVAR Based on the above conditions, which ONE of the following describes the status of the Main Generator and the action directed by AP/34?

The Main Generator is and the operator will _ _ _ _ Generator output voltage to correct this condition.

A. over-excited I increase B. over-excited I decrease C. under-excited I increase D. under-excited I decrease

(

2009 NRC REACTOR OPERATOR EXAM Question 17 T1/G1 - jmb 077 AK1.03, Generator Voltage and Electric Grid Disturbances Knowledge of the operational implications of the following concepts as they apply to Generator Voltage and Electric Grid Disturbances: Under-excitation (3.3/3.4)

KIA MATCH ANALYSIS Requires knowledge of operating characterisitics associated with generator MWe and MVAR and the operational concern with operating in an underexcited condition. Also requires knowledge of the operation of the excitation system to correct the condition.

ANSWER CHOICE ANALYSIS Answer: C A. Incorrect: Generator condition is incorrect. Plausible if MVAR condition is misinterpreted as lagging vs leading. Action to reduce MVAR loading is required and is accomplished by raising generator field voltage to increase generator terminal voltage.

B. Incorrect: Generator condition is incorrect. Plausible if MVAR condition is misinterpreted. Action directed is incorrect but consistent with the action if an over-excited condition did exist.

C. Correct: Negative MVAR indicates that generator terminal voltage is below system voltage and the generator is operating under-excited with a Leading power factor. Action to reduce MVAR loading is required and is accomplished by raising generator field voltage to increase generator terminal voltage.

D. Incorrect: Generator condition is correct. Action directed is incorrect and consistent with the action if an over-excited condition did exist.

Technical Reference(s): 3/AP/34, STG-01S Proposed references to be provided to applicants during examination: None Learning Objective: STG-01S Obj R26, EAP-APG Obj R9 Question Source: N Question History: Last NRC Exam _ _ _ _ __

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis

2009 NRC REACTOR OPERATOR EXAM 1 POINT Question 18 Unit 1 initial conditions:

• Loss of Heat Transfer Conditions exist due to the loss of ALL FDW sources

• RCS pressure = 2320 psig increasing

• Core SCM = 42°F increasing Current conditions:

=

• RCS pressure 2210 psig stable

• Core SCM = 56°F increasing Based on the conditions above, which ONE of the following correctly completes the statement below regarding HPI throttling and the parameter/trend used to determine if the criteria for HPI throttling is met?

HPI flow be throttled because A. may NOT / RCS pressure is stable B. may NOT / CETCs are increasing C. may / RCS pressure is stable D. may / CETCs are decreasing

(

2009 NRC REACTOR OPERATOR EXAM Question 18 T1/G1 - jmb BE04EK2.2, Inadequate Heat Transfer Knowledge of the interrelations between the (Inadequate Heat Transfer) 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.

(4.2/4.2)

KIA MATCH ANALYSIS Requires knowledge of criteria for throttling HPI during HPI Forced Cooling based on the status of core cooling provided by HPI (CETC trend and Core SCM status)

ANSWER CHOICE ANALYSIS Answer: D A. Incorrect: First part is incorrect. Criteria for throttling HPI during HPI cooling is based on Core SCM >0 and CETC decreasing. Plausible if pressure approaching RV-P/T criteria is applied.

B. Incorrect: First part is incorrect. Criteria for throttling HPI during HPI cooling is based on Core SCM >0 and CETC decreasing. Plausible if correlation between increasing Core SCM and stable pressure is not recognized as indication that CETC temps are decreasing.

C. Incorrect: First part is correct. Second part is incorrect as the criteria is based on CETC trend and Core SCM not RCS pressure.

D. Correct: Criteria for throttling HPI during HPI cooling is based on Core SCM >0 and CETC decreasing. Core SCM increasing with RCS pressure stable indicates that CETC temp is decreasing.

Technical Reference(s): EP/1/1800/01 Rule 6 Proposed references to be provided to applicants during examination: None Learning Objective: EAP-HPI CD Obj. R3 Question Source: B (EAP140310)

Question History: Last NRC Exam _ _ _ _ __

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis

2009 NRC REACTOR OPERATOR EXAM 1 POINT Question 19 Unit 3 initial conditions:

• Reactor power = 100%

• SASS is in MANUAL

• ALL PZR level Channels = 220 inches

• Pzr Level Channel 3 is selected Current conditions:

• PZR temperature channel "B" fails LOW Which ONE of the following describes how Standby Shutdown Facility (SSF) Pzr Level indication will be initially affected by the above temperature failure and how will 3HP-120 respond to the failure?

SSF Pzr Level indication will initially ____and 3HP-120 will _ _ __

A. decrease / be unaffected B. stay the same / be unaffected C. decrease / open D. stay the same / open

2009 NRC REACTOR OPERATOR EXAM Question 19 T1/G2 - okm 028AK2.02, Pressurizer Level Control Malfunction Knowledge of the interrelations between the Pressurizer Level Control Malfunctions and the following: Sensors and detectors (2.6/2.7)

KIA MATCH ANALYSIS Requires the student to know the response of the plant to a malfunction in the Pzr Level Control circuit specifically Pzr level temperature compensation.

ANSWER CHOICE ANALYSIS Answer: D A. Incorrect: First part is incorrect. Plausible if SSF Pzr level is assumed to Temperature compensated from the affected channel. Second part incorrect and plausible if relationship between Pzr Temperature and ICCM Train is misapplied and the affect is assumed to be on level 1 and 2 which are associated with ICCM train "A".

B. Incorrect: First part is correct. Second part is incorrect as noted above.

C. Incorrect: First part is plausible as noted in distracter A. Second part is correct as noted below.

D. Correct: PZR temperature channel "8" feeds ICCM Train "8". PZR level 3 comes from ICCM Train "8". With a loss of temperature compensation the associated level indication will decrease. The SSF PZR level indication is not temperature compensated and will not change.

Technical Reference(s): IC-RCI Rev 19, PNS-PZR Rev 16a Proposed references to be provided to applicants during examination: None Learning Objective: PNS-PZR Obj R14/15 & 31, IC-RCI Obj R13 Question Source: N Question History: Last NRC Exam _ _ _ __

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis

2009 NRC REACTOR OPERATOR EXAM 1 POINT Question 20 Plant initial conditions:

• Unit 1 in MODE 6

- De-fuel in progress

• Unit 2 in MODE 5

• Unit 1 & 2 RB Purge Fans running

• 1 RIA-3 (Fuel Transfer Canal Wall) = 1.4 mr/hr

• 1RIA-6 (Spent Fuel Pool) = 1.72 mr/hr Current conditions:

• 1 RIA-3 (Fuel Transfer Canal Wall) = 1.5 mr/hr

=

• 1 RIA-6 (Spent Fuel Pool) 15.2 mr/hr Based on the above conditions, which ONE of the following describes REQUIRED operator actions?

A. Stop Unit 1 RB Purge Fan and start a SFP Filtered exhaust fan.

B. Start ALL Outside Air Booster Fans and stop Unit 1 RB Purge Fan.

C. Stop Unit 2 RB Purge Fan and start a SFP Filtered exhaust fan.

( D. Start ALL Outside Air Booster Fans and close ALL containment penetrations

2009 NRC REACTOR OPERATOR EXAM Question 20 T1/G2 - jmb 036AA2.02, Fuel Handling Incidents Ability to determine and interpret the following as they apply to the Fuel Handling Incidents: Occurrence of a fuel handling incident (3.4/4.1 )

KIA MATCH ANALYSIS Requires the ability to determine the location of a fuel handling accident based on changes in radiation levels and what actions are required to mitigate it.

ANSWER CHOICE ANALYSIS Answer: C A. Incorrect: First part is incorrect. Plausible if RIA indication is misinterpreted. Starting the SFP Filtered Exhaust Fan is directed if the FH accident occurs in the U 1& 2 SFP area.

B. Incorrect: Procedure directs starting the Outside Air Booster fans for both control rooms and stopping U1 RB Purge Fan if the event occurs in the RB.

C. Correct: RIAs indicate spent fuel damage in the SFP. Action to stop Unit 2 purge fan is directed to allow placing the SFP Filtered Exhaust Fan in service.

D. Incorrect: First part is incorrect as noted above. Second part is correct.

Technical Reference(s): AP/9, Spent Fuel Damage Proposed references to be provided to applicants during examination: None Learning Objective: FH-FHS R47 Question Source: 8 (FH014702)

Question History: Last NRC Exam ONS 2007 NRC RO Exam Q#22 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis

2009 NRC REACTOR OPERATOR EXAM 1 POINT Question 21 Unit 1 initial conditions:

• 1B SG tube leakage =20 gpm

• RCS temperature = 525°F

=

• RCS pressure 1200 psig Current conditions:

• RCS temperature =425°F

• RCS pressure = 450 psig Based on the above conditions, which ONE of the following describes the change in SCM and SG tube leakage rate?

From initial to current conditions, subcooling margin has _ _ _ _ _ and SG tube leakage rate has _ _ _ __

Assume SG Tube flaw size remains constant A. increased / increased B. increased / decreased

( C. decreased / increased D. decreased / decreased

2009 NRC REACTOR OPERATOR EXAM Question 21 T1/G2 -OKM 037AK1.01, Steam Generator (S/G) Tube Leak Knowledge of the operational implications of the following concepts as they apply to Steam Generator Tube Leak: Use of steam tables (2.9/3.3)

K/A MATCH ANALYSIS ,

Requires the student to use Steam Tables to calculate sub-cooling margin for initial and current conditions to determine changes in steam generator tube leak rate ANSWER CHOICE ANALYSIS Answer: D A. Incorrect: First part is incorrect. SCM must be determined/calculated; and calculation shows that it has decreased from -44 F to -35 F; Second part is incorrect and requires determining SG pressure from steam tables to determine the leak rate will decrease since the delta P between primary and secondary has decreased.

B. Incorrect: First part is incorrect: Plausible if SCM must incorrectly determined/calculated; Second part is correct.

C. Incorrect: First part is correct: SCM must be determined/calculated; and calculation shows that it has decreased from -44 F to -35 F. Second part is incorrect as noted above.

D. Correct: SCM must be determined/calculated; and calculation shows that it has decreased from -44 F to -35 F. SGTL rate will decrease due to the decrease in primary to secondary DP (RCS to Secondary DP initial conditions

=-365 psig / current conditions =-115 psig.)

Technical Reference(s): Steam Tables Proposed references to be provided to applicants during examination: Steam Tables Learning Objective: EAP-SGTR, R6 Question Source: M (EAP090602)

Question History: Last NRC Exam _ _ _ __

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis

2009 NRC REACTOR OPERATOR EXAM 1 POINT Question 22

( Unit 3 plant conditions:

• Reactor power 100%=

• Main condenser vacuum =28" Hg decreasing slowly Based on the above conditions, which ONE of the following describes the highest vacuum (inches Hg) at which a MANUAL Reactor Trip would be required per 3AP/27 (Loss of Condenser Vacuum)?

A. 25 B. 22 C. 20 D. 19

2009 NRC REACTOR OPERATOR EXAM Question 22 T1/G2 - OKM 051AG2.4.2, Loss of Condenser Vacuum Knowledge of system set points, interlocks and automatic actions associated with EOP entry conditions.

(4.5/4.6)

KIA MATCH ANALYSIS Requires the student to know set points for low vacuum trips and the set point for immediate manual reactor trip from the Loss of Condenser Vacuum AP ANSWER CHOICE ANALYSIS Answer: B A. Incorrect: Plausible in that 25"Hg decreasing is the SA "COND VACUUM LOW" setpoint B. CORRECT: per AP/27 IAAT Step 4.1; manually trip Reactor if ~22" Hg is reached C. Incorrect: Plausible in that 20"Hg provides margin for the operator to respond to trip prior to the MFWPT trip.

D. Incorrect: Plausible in that 19"Hg is the FWP automatic trip on low vacuum Technical Reference(s): 3/AP/27 Rev 03 Proposed references to be provided to applicants during examination: None Learning Objective: EAP-APG R7; STG-MT R13; CF-FPT R8 Question Source: NEW Question History: Last NRC Exam _ _ _ __

Question Cognitive Level: I Memory or Fundamental Knowledge Comprehension or Analysis

2009 NRC REACTOR OPERATOR EXAM 1 POINT Question 23 Unit 1 initial conditions:

• Mode 6

• Core alteration in progress

• All four SR Nls in service

• SR 1NI-1 and SR 1NI-3 are the designated Nls for Fuel Handling Current conditions:

• Power supply to SR 1NI-1 fails (0 vdc)

Based on the above conditions, which ONE of the following describes the impact on fuel movement?

Fuel movement. ..

A. may continue because two SR Nls remain in service.

B. may continue because one of the designated SR Nls is still in service.

C. must be stopped until SR 1NI-1 is returned to service.

D. must be stopped until two designated SR Nls are in service.

2009 NRC REACTOR OPERATOR EXAM Question 23 T1/G2 - okm 032AK1.01, Loss of Source Range Nuclear Instrumentation Knowledge of the operational implications of the following concepts as they apply to Loss of Source Range Nuclear Instrumentation: Effects of voltage changes on performance (2.5/3.1)

KIA MATCH ANALYSIS Requires knowledge of the effects of changing power supply voltage on the shared SR and WR Nls and the operational implications of this change ANSWER CHOICE ANALYSIS Answer: D A. Incorrect: Plausible because the two SR Nls remain in service however the procedure requires the Nls used to monitor core reactivity be designated in advance.

B. Incorrect: Plausible if it is not understood that the procedure requires 2 designated Nls to be operable.

C. Incorrect: Plausible as restoring SR NI-1 would restore 2 designated Nls however the procedure allows an alternate NI to be designated and then continue fuel movement. It is not required to restore NI-1 if another NI is can be designated to replace it.

(

D. Correct: Procedure requires movement to be stopped until 2 Nls used to monitor core reactivity can be designated.

Technical Reference(s): OP/1/A/1502/007 Rev 81 Proposed references to be provided to applicants during examination: None Learning Objective: FH-FHS Obj R20 Question Source: M (FH041401)

Question History: Last NRC Exam _ _ _ _ __

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis

2009 NRC REACTOR OPERATOR EXAM 1 POINT Question 24 Unit 3 initial conditions:

• SBLOCA

• EOP ICC tab is in progress

• No ECCS injection sources are available Current conditions:

• 3A & B HPI pumps operating

• Conditions require the opening of 3RC-4 and PORV Based on the above conditions, which ONE of the following describes actions required to operate 3RC-4 and the reason for aligning this flow path?

To open 3RC-4 place the control switch to to allow RCS pressure to decrease and when the flow path is established ..

A. OPEN AND depress OPEN PERMIT / increase HPI injection flow B. OPEN AND depress OPEN PERMIT / reduce stresses on the SG tubes C. OPEN / increase HPI injection flow D. OPEN / reduce stresses on the SG tubes

2009 NRC REACTOR OPERATOR EXAM Question 24 T1/G2 -okm/jmb 074EA1.23, Inadequate Core Cooling Ability to operate and monitor the following as they apply to an Inadequate Core Cooling: PORV block valve indicators, switches, controls (for both RCS and S/G)

(3.9/4.0)

KIA MATCH ANALYSIS Requires the student to know the controls and to understand the reasons/results on the plant of opening the PORV and RC-4 during an ICC condition ANSWER CHOICE ANALYSIS Answer: C A. Incorrect: First part is incorrect. Operation of RC-4 (PORV Block) requires operating the Control Switch only and verifying light indication changes as the PORV requires the use of the OPEN PERMIT button. RCS pressure will decrease causing HPI flow to increase per the design of the procedure.

B. Incorrect: First part is incorrect. Operation of RC-4 (PORV Block) requires operating the Control Switch only and verifying light indication changes as the PORV requires the use of the OPEN PERMIT button. Second part is plausible because RCS pressure will decrease and will reduce the LlP across the SG tubes. However this effect is not the purpose for reducing RCS pressure.

( C. Correct: Operation of RC-4 (PORV Block) requires operating the Control Switch only and verifying light indication changes. RCS pressure will decrease causing HPI flow to increase per the design of the procedure.

D. Incorrect: First part is correct. Second part is plausible because RCS pressure will decrease and will reduce the LlP across the SG tubes. However this effect is not the purpose for reducing RCS pressure.

Technical Reference(s): EOP ICC Tab, EOP Reference Document and EAP-ICC Lesson Plan Proposed references to be provided to applicants during examination: NONE Learning Objective: EAP-ICC Obj. R5, PNS-Pzr R37 Question Source: N Question History: Last NRC Exam _ _ _ __

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis

2009 NRC REACTOR OPERATOR EXAM 1 POINT Question 25 Unit 3 plant conditions:

• Control Room Evacuation complete due to a non-fire event

• ASDP has been "manned"

• Following indications are observed at the ASDP:

- Turbine Header Pressure =1011 psig and slowly decreasing

- TBVs demand is 12% in Automatic and decreasing

- RCS T Hot =560°F and slowly decreasing

- Pzr Level = 140 inches increasing

- SG SU levels =48 inches and increasing

- ALL RCPs are operating Based on the conditions above, which ONE of the following correctly describes the action(s) required (if any) per AP/8 (Loss of Control Room)?

The operator must take ...

A. NO actions the plant is responding as expected.

B. manual control of TBVs to stabilize SG Pressure.

C. manual control and cycle 3B HPI pump to stabilize Pzr level.

D. manual control of FDW Startup Control Valves and lower SG levels.

2009 NRC REACTOR OPERATOR EXAM Question 25 T1/G2 *jmb BA06AK3.4, Shutdown Outside Control Room Knowledge of the reasons for the following responses as they apply to the (Shutdown Outside Control Room): RO or SRO function within the control room team as appropriate to the assigned position, in such a way that procedures are adhered to and the limitations in the facilities license and amendments are not violated.

(3.8/3.8)

KIA MATCH ANALYSIS Requires knowledge of control parameters and reason for actions required to establish control from the ASDP to establish stable Hot Standby conditions ANSWER CHOICE ANALYSIS Answer: D A. Incorrect: Operator action is required, FDW SU control valves are not performing correctly SG levels are above 25 inches and rising. Plausible as the SG response would be normal if RCPs were not running and SG levels would control at 50% OR Level (240 inches SUR)

B. Incorrect: TBVs are responding normally to post trip conditions and to the overfeed causing SG pressure to decrease C. Incorrect: Per AP/8 Pzr level should be maintained 100-220 inches. Plausible if Pzr level were to be controlled at the normal level of 220 inches D. Correct: Manual control is needed; FDW SU control valves should be closing and controlling at 25 inches on SUR Level post trip (LLL)

Technical Reference(s): APIa Rev 11 Proposed references to be provided to applicants during examination: NONE Learning Objective: IC*ASP Obj. T1, EAP*APG Obj R9 Question Source: N Question History: Last NRC Exam _ _ _ __

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis

2009 NRC REACTOR OPERATOR EXAM 1 POINT Question 26 Boron precipitation is primarily a concern for a leg rupture. Failure to initiate the Boron Dilution flow path when required would challenge the ability to _ _ __

A. hot / prevent boron stratification in the cold leg to ensure a restart event does not occur when the cold leg volume enters the core.

B. hot / ensure boron does not concentrate in the core and block the long term core cooling flow path C. cold / prevent boron stratification in the cold leg to ensure a restart event does not occur when the cold leg volume enters the core.

D. cold / ensure boron does not concentrate in the core and block the long term core cooling flow path

2009 NRC REACTOR OPERATOR EXAM Question 26 T1/G2 -jmb BE08EG2.2.38, LOCA Cooldown Knowledge of conditions and limitations in the facility license (3.6/4.5)

KIA MATCH ANALYSIS Requires knowledge of the requirements for establishing the Long Term Boron dilution flowpath and the impact on the ability to meet the 10CFR50.46 requirements for coolable geometry/flowpath.

ANSWER CHOICE ANALYSIS Answer: D A. Incorrect. First part is plausible if assumed the flowpath is off of the hot leg. Second part is incorrect. Plausible as boron stratification/dilution is a concern for RCP restart criteria following a LOSCM.

B. Incorrect. First part is plausible if assumed the flowpath is off of the hot leg. Second part is correct.

C. Incorrect. First part is correct. During a Cold Leg Rupture, the coolant will enter the vessel through the Core Flood nozzles into the inlet plenum and flow out the Cold Leg break. The core will continue to steam through the Internal Vent Valves and out the Cold Leg Break. This steaming will cause the boric acid to precipitate in the core due to the evaporation of the water. Second part is incorrect. Plausible as boron stratification/dilution is a concern for RCP restart criteria following a LOSCM.

D. Correct. During a Cold Leg Rupture, the coolant will enter the vessel through the Core Flood nozzles into the inlet plenum and flow out the Cold Leg break. The core will continue to steam through the Internal Vent Valves and out the Cold Leg Break. This steaming will cause the boric acid to precipitate in the core due to the evaporation of the water. The precipitated Boron could then block the flowpath for LPI through the core.

Technical Reference(s): EOP LCD tab; EOP Reference Document Proposed references to be provided to applicants during examination: None Learning Objective: PNS-LPI Obj R28 Question Source: M (PNS122707)

Question History: Last NRC Exam _ _ _ __

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis

2009 NRC REACTOR OPERATOR EXAM 1 POINT Question 27 Unit 1 initial conditions:

• Reactor power =100%

• ACB-3 closed Current conditions:

• Switchyard Isolation

• RCS pressure = 1146 psig stable

• All SCM =21°F increasing

• Keowee Unit 2 emergency locked out Based on the above conditions, which ONE of the following describes the first procedure that will be performed by an RO after EOP IMAs and Symptom Check is complete?

A. Enclosure 5.1 (ES Actuation)

B. Enclosure 5.2 (Placing RB Hydrogen Analyzers In Service)

C. Enclosure 5.5 (Pressurizer and LOST Level Control)

O. Enclosure 5.38 (Restoration of Power)

2009 NRC REACTOR OPERATOR EXAM Question 27 T1/G2 - jmb BE14EA2.1, EOP Enclosures Ability to determine and interpret the following as they apply to the (EOP Enclosures): Facility conditions and selection of appropriate procedures during abnormal and emergency operations.

(3.4/4.0)

KIA MATCH ANALYSIS Requires the student to recognize plant conditions and select the appropriate Rule/Enclosure to initiate.

ANSWER CHOICE ANALYSIS Answer: A A. Correct: Conditions indicate ES has actuated. Encl. 5.1 is the appropriate Enclosure to use first.

B. Incorrect: Plausible as this Enclosure is directed to be performed as part of Encl 5.1.

C. Incorrect: Plausible as this enclosure would be appropriate to use first if the ES conditions were not present.

D. Incorrect: Plausible as this Enclosure would be appropriate if it is not recognized that power is still available. The power conditions given could lead to the assumption that power is not available as this Enc!. is used to restore power.

Technical Reference(s): EP/1/1800/001 Parallel Action Page, OMP 1-18 Proposed references to be provided to applicants during examination: None Learning Objective: EAP-EOP, R26 Question Source: N Question History: Last NRC Exam _ _ _ __

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis

2009 NRC REACTOR OPERATOR EXAM 1 POINT Question 28 Unit 1 plant conditions:

• Unit 1 startup in progress

• 1A 1 and 1 B 1 RCPs are operating Based on the above conditions, which ONE of the following describes a condition that would prevent 1A2 RCP from starting?

A. RCS Tc =340°F B. RCP Oil Lift Pressure = 700 psig C. HPI Seal Injection flow rate = 28 gpm D. Total Component Cooling flow = 554 gpm

2009 NRC REACTOR OPERATOR EXAM Question 28 T2/G1 -okm 003K4.04, Reactor Coolant Pump Knowledge of RCPS design feature{s) and/or interlock{s) which provide for the following: Adequate cooling of RCP motor and seals (2.8/3.1)

KIA MATCH ANALYSIS Requires the student to know RCP motor starting interlocks associated with motor cooling and seals ANSWER CHOICE ANALYSIS Answer: D A. Incorrect: Plausible; RC temperature must be ~350°F to start the fourth RCP. This is the 3rd RCP start.

B. Incorrect: Plausible; must know that oil lift pressure interlock is satisfied if ~ 600 psig C. Incorrect: Plausible; must know difference between Unit 1 and Units 2/3 SI requirements; for Unit 1 is ~ 22 gpm (Unit 2/3 is ~ 30 gpm)

D. Correct: CC flow should be > 575 gpm Technical Reference(s): PNS-CPM Proposed references to be provided to applicants during examination: None Learning Objective: PNS-CPM, R19 Question Source: 8 (PNS061401)

Question History: Last NRC Exam 2008 ONS Retest RO Exam Q#28 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis

2009 NRC REACTOR OPERATOR EXAM 1 POINT Question 29 Unit 1 initial conditions:

• Reactor power 100%=

Current conditions:

• 1TE de-energized Based on the above conditions, which ONE of the following describes the status of the listed HPI components?

A. 1B HPI pump de-energized 1HP-26 energized B. 1B HPI pump de-energized 1HP-27 de-energized C. 1C HPI pump de-energized 1HP-26 de-energized D. 1C HPI pump de-energized 1HP-27 energized

2009 NRC REACTOR OPERATOR EXAM Question 29 T2/G1 - gcw 004K2.05, Chemical and Volume Control System (CVCS)

( Knowledge of bus power supplies to the following: MOVs (2.7/2.9)

KIA MATCH ANALYSIS Requires knowledge to recognize the potential problems associated with a loss of power to HPI pumps and MOVs.

ANSWER CHOICE ANALYSIS Answer: A A. Correct: 1TE supplies power to the 1B HPI pump and 1HP-26 is supplied by 1TC B. Incorrect: Fist part is correct. Second part is incorrect. Plausible as could be correct if the pumps power supplies followed the standard convention. 1HP-27 is powered from 1TO.

C. Incorrect: both parts are incorrect. Plausible if you do not know the power supply convention for the HPI system. 1HP-26 is supplied by 1TC O. Incorrect: First parts is incorrect. Plausible if you do not know the power supply convention for the HPI system. Second part is correct.

Technical Reference(s): PNS-HPI Proposed references to be provided to applicants during examination: None Learning Objective: PNS-HPI T2 Question Source: N Question History: Last NRC Exam _ _ _ __

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis

2009 NRC REACTOR OPERATOR EXAM 1 POINT Question 30 Unit 3 plant conditions:

• Reactor power = 100%

• Aligning of 3B LPI Train for BWST recirc is in progress Based on the above conditions, which ONE of the following describes a condition that must be met prior to operating 3LP-42 (LPI RETURN TO BWST) in accordance with OP/3/A/1104/004 (Low Pressure Injection System)?

An individual must be designated to ...

A. close 3LP-42 in the event of an Engineered Safeguards actuation.

B. close 3LP-42 in the event of an LPI piping overpressure condition.

C. throttle 3LP-42 to ensure adequate LPI pump NPSH is maintained.

D. throttle 3LP-42 to ensure LPI pump minimum flow requirements are met.

(

2009 NRC REACTOR OPERATOR EXAM Question 30 T2/G1 - okm/jmb 005A4.05, Residual Heat Removal System (RHRS)

( Ability to manually operate and/or monitor in the control room: Position of RWST recirculation valve (locked when not in use, continuously monitored when in use).

(2.8/2.8)

KIA MATCH ANALYSIS Requires the student to know the procedural requirement and the reason for the requirement associated with operation of the BWST recirculation valve ANSWER CHOICE ANALYSIS Answer: A A. Correct: Per LPI operating procedure limit and precaution an individual must be designated at LP-42 (and LP-40 and -41) to manually close these valves should an ES actuation occur; required for Modes 1-4 B. Incorrect: Plausible as the current RCS Pressure is above the pressure limits associated with the LPI system piping and leakage past an isolation valve could lead to an overpressure condition. (OE: PIP 0-97-4645)

C. Incorrect: Plausible because flowrates are established to ensure pump flow is within the operating limits to prevent pump from running with inadequate NPSH, however LP-42 is fully opened by the procedure and the LPI cooler outlet valve is used to throttle flow.

D. Incorrect: Plausible because flowrates are established to ensure pump flow is within the operating limits to prevent pump from running with inadequate flow, however LP-42 is fully opened by the procedure and the LPI cooler outlet valve is used to throttle flow.

Technical Reference(s): Rev 130 OP/3/Al11 04/004 L&P 2.22 & Enclosure 4.4 Proposed references to be provided to applicants during examination: NONE Learning Objective: PNS-LPI Obj. R34 Question Source: N Question History: Last NRC Exam _ _ _ __

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis

2009 NRC REACTOR OPERATOR EXAM 1 POINT Question 31 Unit 1 initial conditions:

• Reactor power 100% =

Current conditions:

• SBLOCA

• ES Channels 1 and 2 fail to actuate Based on the above conditions, which ONE of the following describes the affect on the fuel and the required action per EOP Enclosure 5.1 (ES Actuation)?

A. Peak cladding temperature could exceed the ECCS criteria of 1200°F Manually actuate ES Digital Channels 1 and 2 B. Peak cladding temperature could exceed the ECCS criteria of 1200°F Immediately place all ES Channels 1 and 2 components in their ES position C. Peak cladding temperature could exceed the ECCS criteria of 2200°F Manually actuate ES Digital Channels 1 and 2 D. Peak cladding temperature could exceed the ECCS criteria of 2200°F

( Immediately place all ES Channels 1 and 2 components in their ES position

2009 NRC REACTOR OPERATOR EXAM Question 31 T2/G1 - gcw 006K3.02, Emergency Core Cooling System (ECCS)

( Knowledge of the effect that a loss or malfunction of the ECCS will have on the following: Fuel (4.2/4.4)

KIA MATCH ANALYSIS Requires the student to know that a loss or malfunction of the ECCS will affect fuel clad temperature ANSWER CHOICE ANALYSIS Answer: C A. Incorrect: First part is incorrect. Plausible as 12000F is the threshold temp used to apply Severe Accident Management Guidelines in the EOP. Second part is correct.

B. Incorrect: First part is incorrect as noted above. Second part is incorrect. Plausible as the actions would be appropriate during individual component verification when directed by Enc!. 5.1 C. Correct: Failure of ES 1 & 2 would result in failure to start the HPI pumps which would lead elevated clad temperatures and a challenge to the ECCS criteria. Enc/. 5.1 directs manually actuating ES channels that fail to actuate.

D. Incorrect: First part is correct. Second part is incorrect as noted above.

Technical Reference(s): EP/1/A11800/001 Rev 36 Enc/. 5.1 Proposed references to be provided to applicants during examination: NONE Learning Objective: EAP-ESA Obj. R9, IC-ES Obj. R1 Question Source: NEW Question History: Last NRC Exam _ _ _ __

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis

2009 NRC REACTOR OPERATOR EXAM 1 POINT Question 32

( Unit 1 initial conditions:

• Loss of all Feedwater

• HPI forced cooling initiated

• Quench Tank pressure = 50 psig increasing Current conditions:

• Quench Tank pressure =3 psig stable Which ONE of the following describes the containment response to the above conditions?

A. RB Normal sump level rises. 1RIA-47 radiation level increases B. RB Normal sump level rises. 1RIA-47 radiation level remains constant C. RB Normal sump level remains constant. 1RIA-47 radiation level increases D. RB Normal sump level remains constant. 1RIA-47 radiation level remains constant

2009 NRC REACTOR OPERATOR EXAM Question 32 T2/G1 - gcw 007K3.01 Pressurizer Relief/Quench Tank Knowledge of the effect that a loss or malfunction of the PRTS will have on the following: Containment (3.3/3.6)

KIA MATCH ANALYSIS Requires knowledge of impact of discharge from PORV to the Quench Tank and indications of failed/blown rupture disk and the impact of the failure on containment parameters.

Plausibility based around whether applicant recognizes status of QT rupture disk. If disk is assumed to have blown, then containment pressure would rise. With normal levels of RCS activity an applicant would have to determine what the effects on containment radiation would be and where the leakage is directed (Misc. Waste vs RBNS)

ANSWER CHOICE ANALYSIS Answer: A A. Correct. Decrease in Quench Tank pressure indicates the Rupture Disk has blown. Inventory from the Quench Tank will go to the RBNS causing a level increase. RCS activity in the inventory will result in 1RIA-47 reading increase.

( B. Incorrect. RBNS response is correct. 1RIA-47 response is incorrect but plausible if RCS activity is assumed to be negligible or the source of QT pressure rise is due to DW/B Bleed inleakage. (OE)

C. Incorrect. RBNS response is incorrect but plausible if the pressure reduction is assumed to be caused by draining to the Misc Waste System via the Component Drain flowpath. 1RIA-47 response is correct.

D. Incorrect. RBNS response is incorrect as noted above. 1RIA-47 response is consistent with inventory going to Misc Waste or assuming activity is negligible (OE with Demin Water & B Bleed Holdup Tank water leak into the Quench Tank).

Technical Reference(s): PNS-CS Rev 16, PNS-PZR Rev 16a Proposed references to be provided to applicants during examination: None Learning Objective: PNS-CS (R7)

Question Source: New Question History: Last NRC Exam: _ _ _ _ _ __

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis

2009 NRC REACTOR OPERATOR EXAM 1 POINT Question 33

( Which ONE of the following describes the interlock associated with the High Pressure Injection (HPI) valves (HP-1 and HP-2) and Component Cooling (CC) valves (CC-1 and CC-2) for the Letdown Coolers?

The interlock ...

A. isolates the Letdown Cooler if letdown temperature reaches interlock setpoint.

B. ensures CC flow is established before letdown flow.

C. prevents placing the two Letdown Coolers in service at the same time.

D. ensures letdown flow,is secured after CC flow.

(

2009 NRC REACTOR OPERATOR EXAM Question 33 T2/G1 - okm/jmb 008K1.02, Component Cooling Water System (CCWS)

Knowledge of the physical connections and/or cause-effect relationships between the CCWS and the following systems: Loads cooled by CCWS (3.3/3.4)

KIA MATCH ANALYSIS Requires the student to know that the impact of reduced system flow on the components cooled by the Component Cooling Water System.

ANSWER CHOICE ANALYSIS Answer: B A. Incorrect: Plausible if the High Temp interlock to close HP-5 is misapplied to HP-1 &

HP-2.

B. Correct: Interlock opens CC-1 & CC2 prior to opening HP-1 & HP-2 to ensure CC flow is established prior to establishing Letdown Flow.

C. Incorrect: Plausible if interlock is misapplied to prevent simultaneous operation of the cooler valves.

D. Incorrect: Plausible as this is the reverse order of the interlock.

(

l, Technical Reference(s): PNS-CC, OP/1/Al1104/008 Rev. 59 Proposed references to be provided to applicants during examination: NONE Learning Objective: PNS-CC R15 Question Source: B (PNS112202)

Question History: Last NRC Exam _ _ _ __

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis

(

2009 NRC REACTOR OPERATOR EXAM 1 POINT Question 34 Which ONE of the following states the automatic OPEN setpoints (psig) for 1 RC-1 (pzr Spray) and 1 RC-66 (PORV) in Mode 1?

A. 2205 / 2450 B. 2205 / 2500 C. 2255 / 2450 D. 2255 / 2500

2009 NRC REACTOR OPERATOR EXAM Question 34 T2/G1 - okm/jmb 010K4.03, Pressurizer Pressure Control System (PZR PCS)

Knowledge of PZR PCS design feature(s) and/or interlock(s) which provide for the following: Over pressure control (3.8/4.1 )

KIA MATCH ANALYSIS Requires knowledge of Pzr PCS setpoints for automatic pressure control.

ANSWER CHOICE ANALYSIS Answer: A A. Correct: 1RC-1 (pzr Spray) setpoint is 2205 psig and 1RC-66 (PORV) is 2450 psig when in HIGH (Mode 1)

B. Incorrect: 1RC-1 (pzr Spray) setpoint is correct. 1RC-66 (PORV) setpoint is incorrect. Plausible as 2500 psig is the Pzr Safety Valve setpoint.

C. Incorrect: 1RC-1 (pzr Spray) setpoint is incorrect. Plausible as 2255 psig is the Pzr High pressure alarm setpoint. 1RC-66 (PORV) setpoint is correct.

D. Incorrect: 1RC-1 (pzr Spray) setpoint is incorrect. Plausible as 2255 psig is the Pzr High pressure alarm setpoint. 1RC-66 (PORV) setpoint is incorrect as noted above.

Technical Reference(s): PNS-PZR, Proposed references to be provided to applicants during examination: NONE Learning Objective: PNS-PZR R5 Question Source: NEW Question History: Last NRC Exam _ _ _ __

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis

2009 NRC REACTOR OPERATOR EXAM 1 POINT Question 35 Unit 2 initial conditions:

• Reactor power 100% =

• Feedwater transient occurs Current conditions:

=

• Reactor power 65% stable

=

• Pzr level 250" slowly decreasing

=

• RCS pressure 2195 psig (highest reached) slowly decreasing

• Pzr temperature 627°F =

Which ONE of the following describes the status of the Pzr heaters and the yellow dot on the OAC PIT display within thirty seconds of the above transient?

A. All Pzr heaters are "OFF" I yellow dot is "ON" the blue saturation line B. All Pzr heaters are "OFF" I yellow dot is "LEFT" of the blue saturation line C. Some Pzr heaters are "ON" I yellow dot is "ON" the blue saturation line D. Some Pzr heaters are "ON" I yellow dot is "LEFT" of the blue saturation line

2009 NRC REACTOR OPERATOR EXAM Question 35 T2/G1 - okm 01 OK5.01, Pressurizer Pressure Control System (PZR PCS)

Knowledge of the operational implications of the following concepts as they apply to the PZR PCS: Determination of condition of fluid in PZR, using steam tables (3.5/4.0)

KIA MATCH ANALYSIS Requires knowledge of PZR parameters, controls/interlocks, and use of Steam Tables to determine the condition of the PZR fluid and the response of the PZR PCS.

ANSWER CHOICE ANALYSIS Answer: D A. Incorrect: Both parts are incorrect. First part is plausible because normally all heaters will be off at this pressure. Second part is plausible if steam tables are not used correctly.

B. Incorrect: First part is plausible because normally all heaters will be off at this pressure. The second part is correct because the PIT display will show the yellow Pzr "dot" has moved to the left of the saturation curve whenever subcooled conditions exist in the Pzr (627°F) .

C. Incorrect: The first part is correct because the transient resulted in an insurge which subcooled the pzr; this causes the Pzr Water Space Saturation Recovery circuit to turn ON Bank 2 Heaters. The second part is incorrect and plausible if steam tables are not used correctly.

D. Correct: The transient resulted in an insurge which subcooled the pzr; this causes the Pzr Water Space Saturation Recovery circuit to turn ON Bank 2 Heaters; the PIT display will show the yellow Pzr "dot" has moved to the left of the saturation curve whenever subcooled conditions exist in the Pzr (627°F)

Technical Reference(s): PNS-PZR Proposed references to be provided to applicants during examination: NONE Learning Objective: PNS-PZR R22, R27, R29 Question Source: Bank (PNS142902)

Question History: Last NRC Exam _ _ _ __

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis

2009 NRC REACTOR OPERATOR EXAM 1 POINT Question 36 Unit 2 plant conditions:

• Reactor power =100%

• I&E technicians incorrectly remove the HI TEMPERATURE TRIP bistable from the 2B RPS Channel Based on the above conditions, which ONE of the following describes the affect on the Reactor Protection System?

2B RPS Channel is ____ and the associated CRD Breaker is _ _ __

A. Tripped / Closed B. Tripped / Open C. Not Tripped / Closed D. Not Tripped / Open

(

2009 NRC REACTOR OPERATOR EXAM Question 36 T2/G1 - okm 012K6.01, Reactor Protection System

( Knowledge of the effect of a loss or malfunction of the following will have on the RPS: Bistables and bistable test equipment (2.8/3.3)

KIA MATCH ANALYSIS Requires knowledge of RPS Channel internal relationships between RPS bistables, relays, and associated CRD Breakers due to loss of a bistable ANSWER CHOICE ANALYSIS Answer: A A. Correct: Loss of the bistable results in the 2B K-Relay becoming de-energized.

This trips the 2B RPS Channel, however power is still available to the 2B CRD Breaker from the other 3 RPS Channels.

B. Incorrect: Plausible in that the 2B RPS Channel is tripped; however power is still available to the 2B CRD Breaker from the other 3 RPS Channels C. Incorrect: Plausible in that the student must know that loss of anyone bistable will trip an (2B) RPS Channel. Power is still available to the 2B CRD Breaker from the other 3 RPS Channels so the breaker is closed.

( D. Incorrect: Plausible in that the student must know that loss of anyone bistable will trip an (2B) RPS Channel and that power is still available to the 2B CRD Breaker from the other 3 RPS Channels Technical Reference(s): IC-RPS Proposed references to be provided to applicants during examination: NONE Learning Objective: IC-RPS Obj. R7 R22 Question Source: N Question History: Last NRC Exam _ _ _ __

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis

2009 NRC REACTOR OPERATOR EXAM 1 POINT Question 37 Unit 3 plant conditions:

• Reactor Tripped due to SBLOCA

• ES Digital Channels 1, 3, and 5 failed to automatically actuate Based on the above conditions, which ONE of the following lists the safety related components that will be in their ES condition?

ASSUME NO OPERATOR ACTIONS A. 3A HPI Pump / 3B LPI Pump / 3A RBCU B. 3B HPI Pump / 3B LPI Pump / 3C RBCU C. 3A HPI Pump / 3A LPI Pump / 3A RBCU D. 3B HPI Pump / 3A LPI Pump / 3C RBCU

2009 NRC REACTOR OPERATOR EXAM Question 37 T2/G1 - okm/jmb 013K5.01, Engineered Safety Features Actuation System (ESFAS)

Knowledge of the operational implications of the following concepts as they apply to the ESFAS: Definitions of safety train and ESF channel (2.8/3.2)

KIA MATCH ANALYSIS Requires knowledge of the ES Channels to their associated safety train components ANSWER CHOICE ANALYSIS Answer: B A. Incorrect: 3A HPI Pump would not have gone to its ES position B. Correct: All components are on ES Digital Channels 2, 4, and 6 C. Incorrect: 3A LPI Pump would not have gone to its ES position D. Incorrect: 3A LPI Pump would not have gone to its ES position Technical Reference(s): EOP Enc. 5.1 Rev 36a Proposed references to be provided to applicants during examination: NONE Learning Objective: IC-ES Obj. R14, EAP-ESA Obj R2 Question Source: NEW Question History: Last NRC Exam _ _ _ __

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis

2009 NRC REACTOR OPERATOR EXAM 1 POINT Question 38 Unit 1 initial conditions:

• Reactor power = 50%

Current conditions:

• LBLOCA occurs

• 1TC de-energized Based on the above conditions, which ONE of the following describes the status of Reactor Building Cooling Units five (5) minutes after ES actuates?

ASSUME NO OPERATOR ACTIONS 1A RBCU 1B RBCU 1C RBCU A. LOW LOW OFF B. LOW OFF LOW C. OFF LOW LOW D. OFF OFF OFF

2009 NRC REACTOR OPERATOR EXAM Question 38 T2/G1 - okm/jmb 022K2.01, Containment Cooling System (CCS)

Knowledge of power supplies to the following: Containment cooling fans (3.0/3.1)

KIA MATCH ANALYSIS Requires knowledge of power supplies to Reactor Building Cooling Units (RBCUs)

ANSWER CHOICE ANALYSIS Answer: C A. Incorrect: Plausible as the RBCU power supplies are not sequenced such that the letter designator follows the power supply arrangement. If 1C RBCU fan is applied to TC bus this choice would be plausible.

B. Incorrect: Plausible if candidate confuses the typical power supply arrangement where TC supplies "B" safety train components and TE supplies "C" safety train components.

C. Correct: 1TD suppies 1X9 which supplies 1C RBCU and 1TE supplies 1XS3 which supplies 1B RBCU and 1TC supplies 1XS8 which supplies 1A RBCU. ES will only start the 1B & 1C RBCUs.

D. Incorrect: Plausible as there is a time delay on the restart of the RBCUs. Incorrect application of the time delay could result in selecting this distracter.

Technical Reference(s): PNS-RBC, EL-EPD Proposed references to be provided to applicants during examination: NONE Learning Objective: PNS-RBC R1, R14, R15 Question Source: NEW Question History: Last NRC Exam _ _ _ __

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis

2009 NRC REACTOR OPERATOR EXAM 1 POINT Question 39 Unit 2 initial conditions:

• Reactor Building Spray actuated Current conditions:

• RB Pressure = 2 psig slowly decreasing Based on the above conditions, which ONE of the following describes the operation and any control room actions associated with resetting the RB Pressure Contact Buffers and ES Digital Channels 7 & 8?

RB Pressure Contact Buffers ____ and ES Digital Channels 7 & 8_ _ __

A. automatically reset / must be reset manually B. automatically reset / automatically reset C. must be reset manually / must be reset manually D. must be reset manually / automatically reset

2009 NRC REACTOR OPERATOR EXAM Question 39 T2/G1 *okm 026A4.05, Containment Spray System (CSS)

( Ability to manually operate and/or monitor in the control room: Containment spray reset switches (3.5/3.5)

KIA MATCH ANALYSIS Requires knowledge of ES RBS system operation concerning the automatic features and manual actions needed to reset containment spray channels. (Oconee does not have "Containment spray reset switches")

ANSWER CHOICE ANALYSIS Answer: A A. Correct: RBS pressure switches supply Analog Channel Contact Buffers that automatically reset themselves when RB pressure decreases below 10 psig; this is unique as compared to the other Analog channels which must be manually reset once actuated. ES Digital Channels 7 & 8 are like the other Digital channels in that they must be manually reset.

B. Incorrect: Plausible in that the first part is correct. It is possible for the candidate to confuse the unique aspect of the RBS Contact Buffers automatically resetting to the Digital channels. Also, ES permissive signals automatically reset.

C. Incorrect: Plausible in that the second part is correct. It is possible for the student to confuse the standard manual resetting of the Digital channels to the RBS Analog channels D. Incorrect: Plausible in that this is the exact opposite of actual operation; candidate could confuse the two parts knowing that ES RBS channel operation is different.

Also, ES permissive signals automatically reset.

Technical Reference(s): IC*ES lesson Proposed references to be provided to applicants during examination: NONE Learning Objective: IC*ES R3, R13 Question Source: NEW Question History: Last NRC Exam _ _ _ __

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis

2009 NRC REACTOR OPERATOR EXAM 1 POINT Question 40 Unit 1 plant conditions:

• SB LOCA has occurred

• EOP Enclosure 5.12 (ECCS Suction Swap to RBES) is in progress

• 1LP-19 & 1LP-20 (1A &1 B RX BLDG SUCTION) are OPEN

• 1LP-21 & 1LP-22 (1A & 1B LPI BWST SUCTION) are OPEN

=

• RB pressure 12 psig stable Based on the above conditions, which ONE of the following describes the suction source that is supplying the RB Spray pumps and the reason for the addition of caustic to the RBES?

A. BWST / to enhance iodine entrainment in RB Spray water B. BWST / to minimize hydrogen production from zirc-water reaction C. RBES / to enhance iodine entrainment in RB Spray water D. RBES / to minimize hydrogen production from zirc-water reaction

2009 NRC REACTOR OPERATOR EXAM Question 40 T2/G1 - okm 026G2.4.21, Containment Spray System (CSS)

Knowledge of the parameters and logic used to assess the status of safety functions, such as reactivity control, core cooling and heat removal, reactor coolant system integrity, containment conditions, radioactivity release control, etc.

(4.0/4.6)

KIA MATCH ANALYSIS Requires knowledge of RB Spray system status based on containment parameters to assess safety functions such as radioactivity release control (iodine)

ANSWER CHOICE ANALYSIS Answer: C A. Incorrect: First part is incorrect. Plausible as it would be the source if RB pressure is less than the press supplied to the pump from the BWST. The current pressure is higher than the head supplied from the BWST. The second part is correct.

B. Incorrect: First part is incorrect as noted above. Second part is incorrect. Plausible as pH is controlled to minimize H2 production from interaction with aluminum and zinc in the RB not the zirc-water reaction which takes place in the Reactor Vessel.

C. Correct: The current pressure is higher than the head supplied from the BWST so the suction source to the RBS pumps is the RBES. One of the purposes of the causic addition is to enhance the absorption of iodine into solution in the RBES.

D. Incorrect: First part is correct. Second part is incorrect as noted above.

Technical Reference(s): EP/1/A11800/001 Rev 36, LOSCM Tab, Enc/. 5.12 Proposed references to be provided to applicants during examination: NONE Learning Objective: PNS-BS R1, EAP-LOSCM R23 Question Source: M (EAP062301)

Question History: Last NRC Exam _ _ _ __

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis

2009 NRC REACTOR OPERATOR EXAM 1 POINT Question 41

( Unit 1 initial conditions:

• Reactor Power = 100% stable

• 1RIA-16 = 14 mr/hr increasing

• 1RIA-17 = 0.01 mr/hr stable

=

• 1RIA-59 12 gpm slowly increasing

=

• 1RIA-60 5.8 E -3 gpd stable

• 1RIA-40 in alarm HIGH Current conditions:

=

• Rx power 45% decreasing

• 1SA-8/E-1 0 "RM N-16 Primary to Secondary Tube Leak" actuated

=

• 1RIA-16 21 mr/hr increasing

=

• 1RIA-17 3.7 mr/hr increasing

=

• 1RIA-59 27.1 gpm increasing

=

• 1RIA-60 1.9 gpm increasing Based on the above conditions, which ONE of the following describes the SG(s) with indications of a tube leak and the procedure used to mitigate this event?

A. 1A SG ONLY I AP-31 (Primary to Secondary Leakage)

B. 1A SG ONLY I EOP SGTR Tab C. 1A & 1B SG I AP-31 (Primary to Secondary Leakage)

D. 1A & 1B SG I EOP SGTR Tab

2009 NRC REACTOR OPERATOR EXAM Question 41 T2/G1 -jmb/okm

( 039A2.03, Main and Reheat Steam System (MRSS)

, Ability to (a) predict the impacts of the following malfunctions or operations on the MRSS; and (b) based on predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Indications and alarms for main steam and area radiation monitors (during SGTR)

(3.4/3.7)

KIA MATCH ANALYSIS Requires the ability to calculate the leakrate of a SGTR using the change in area rad monitor reading and the presence of Rad monitor alarms and determine the appropriate procedure to use for the event.

ANSWER CHOICE ANALYSIS Answer: D A. Incorrect: Selected SG is incorrect. Plausible as the increase in RIA readings is significantly higher in the 1A SG. Procedure selection is incorrect. Plausible if threshold for transition to SGTR is not recognized.

B. Incorrect: Selected SG is incorrect. Plausible as the increase in RIA readings is significantly higher in the 1A SG. Procedure selection is correct. Transition to SGTR if leakrate is >25 gpm.

C. Incorrect: SG selection is correct. Procedure selection is incorrect as noted above D. Correct: Both RIA-59 & RIA-60 indicate an increase. 1RIA-59 and -60 both reading elevated indicate leaks in both SGs. Calculated leakrate is >25 gpm transistion criteria between AP/31 and SGTR Tab using either the RCS inventory balance or 1RIA-59 & 60.

Technical Reference(s): EOP Reference Document - SGTR pg 1 Proposed references to be provided to applicants during examination: NONE Learning Objective: EAP-SGTR R2, RAD-RIA Obj. R2 Question Source: N Question History: Last NRC Exam _ _ _ __

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis

2009 NRC REACTOR OPERATOR EXAM 1 POINT Question 42

( Unit 2 initial conditions:

• Reactor power = 100%

Current conditions:

• Feedwater Valve L.\P controlling signal fails LOW Assume no operator actions have been taken Based on the above conditions, which ONE of the following describes the Main Feedwater Pump Turbines (FWPTs) response and the required actions per AP/28 (ICS Instrument Failures)?

BOTH FWPTs speed -------

/ Place ----------

A. increases to HSS (high speed stop) / BOTH 2A & 2B MAIN FDW PUMPs to Hand and lower speed to prevent tripping FWPs on high discharge pressure B. decreases to LSS (low speed stop) / BOTH 2A & 2B MAIN FDW PUMPs to Hand and raise speed to prevent tripping the Reactor on high RCS pressure C. increases to HSS (high speed stop) / ALL Feedwater Control Valves (Main &

Startup) to Hand and adjust as necessary to restore stable plant conditions D. decreases to LSS (low speed stop) / ALL Feedwater Control Valves (Main &

Startup) to Hand and adjust as necessary to restore stable plant conditions

2009 NRC REACTOR OPERATOR EXAM Question 42 T2/G1 - okm 039A4.03, Main and Reheat Steam System (MRSS)

Ability to manually operate and/or monitor in the control room: MFW pump turbines (2.8/2.8)

K/A MATCH ANALYSIS Requires the ability to diagnose a failure that specifically affects MFW pumps; to understand the affect and to know the required actions.

ANSWER CHOICE ANALYSIS Answer: A A. Correct: A low valve dp failure causes both MFPs to speed up to the HSS (high speed stop). The operator must take the MFP bailey stations to Hand and decrease speed to restore/stabilize feed flow. Without this action the MFPs may trip on high discharge pressure B. Incorrect: Plausible if mis-diagnoses of low dp means low feed flow and FWPT speed will need to be increased. Correspondingly low feed flow means that the RCS could trip on high RCS pressure C. Incorrect: Plausible in that MFP speed will increase but there is no speed correction back to original flow. MCVs operation as stated is acceptable per the PTR guidelines D. Incorrect: Plausible if misdiagnoses FWPT response and there is no speed correction back to original flow. MCVs operation as stated is acceptable per PTR guidelines Technical Reference(s): AP/28 Rev. 14 - Valve DP Failure Proposed references to be provided to applicants during examination: NONE Learning Objective: STG-ICS Obj. R21 Question Source: NEW Question History: Last NRC Exam _ _ _ __

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis

2009 NRC REACTOR OPERATOR EXAM 1 POINT Question 43

( Unit 3 initial conditions:

• 04:00:00

• Reactor power = 70% stable

• 3A Main Feed Pump suction pressure =236 psig decreasing Current conditions:

• 04:01 :25

• 3A Main FDW Pump suction pressure =230 psig increasing Based on the current conditions, which ONE of the following describes the current status of the Main Feedwater pumps and the plant?

A. Both MFPs are operating I Plant runback in progress at 20%/min B. Both MFPs are operating I Reactor power = 70% stable C. 3A MFP has tripped I Plant runback in progress at 20%/min D. 3A MFP has tripped I Reactor power =65% stable

(

2009 NRC REACTOR OPERATOR EXAM Question 43 T2/G1 - okm 059A3.03, Main Feedwater (MFW) System Ability to monitor automatic operation of the MFW, including: Feedwater pump suction flow pressure (2.5/2.6)

KIA MATCH ANALYSIS Requires knowledge of setpoints for MFP low suction pressure runback / trip and the ability to determine the status of the plant and MFW system components ANSWER CHOICE ANALYSIS Answer: A A. Correct: Conditions are met for a CBP/Fdw Pump Low Suction Pressure runback <<235 psig; for >90 sec MFP trips; Runback rate = 20%/min.). Only 85 secs has elapsed therefore runback is still in progress and MFPT is not tripped.

B. Incorrect: Plausible in that both MFPs are still operating << 90 secs); second part is plausible as it is the current power level and is consistent with the FDWP operating status (i.e. no run back to 65% due to FDWPT Trip)

C. Incorrect: Plausible if the time delay is incorrectly assumed to be met and the MFP has tripped. Runback status is correct.

D. Incorrect: Plausible if the time delay is incorrectly assumed to be met and the MFP has tripped. Second part is consistent with MFP status and the runback is incorrectly assumed to have stopped at the loss of MFP run back setpoint Technical Reference(s): CF-FDW, STG-ICS LP pg 27 Proposed references to be provided to applicants during examination: NONE Learning Objective: CF-FDW R7 & R38; STG-ICS R3 & R4 Question Source: NEW Question History: Last NRC Exam _ _ _ __

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis

2009 NRC REACTOR OPERATOR EXAM 1 POINT Question 44 Unit 1 initial conditions:

• Reactor power 100% =

• 1TDEFWP is OOS Current conditions:

• Both Main Feed Pumps trip

• 1A MDEFWP fails to start

=

• 1A SG level 26 inches XSUR decreasing

• 1B SG level = 185 inches XSUR increasing

• Enclosure 5.9 (Extended EFDW Operation) is in progress

• 1FDW-313 and 1FDW-314 (1A & 1B EFDW LINE DISCH X-CONNs) open Based on the above conditions, which ONE of the following describes the maximum flowrate (gpm) allowed to each SG per EOP Rule 7?

A. 1A SG =300 / 1B SG =300 B. 1A SG = 500 / 1B SG = 450 C. 1A SG =600 / 1B SG =600 D. 1A SG = 100 / 1B SG = 600

2009 NRC REACTOR OPERATOR EXAM Question 44 T2/G1 - okm 061A2.04, Auxiliary / Emergency Feedwater (AFW) System Ability to (a) predict the impacts of the following malfunctions or operations on the AFW; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: pump failure or improper operation (3.4/3.8)

KIA MATCH ANALYSIS Requires an understanding of the impact of a loss of 2 EFDWPs and the direction contained in the EOP Rules to address the condition ANSWER CHOICE ANALYSIS Answer: A A. Correct: Total flowrate given is the pump flow limit and is balanced between SG's which would be desired.

B. Incorrect: Total flowrate given is incorrect. Plausible as it the limit associated with the TDEFDWP. 1A SG would require a higher flow rate to recover level as it is lower than the 1B SG.

C. Incorrect: Plausible as the values given would be the normal values if both the MDEFDW pumps are running.

(

D. Incorrect: Plausible as the flowrates are consistent with the Dry SG flow limitation for feeding the 1A SG vice with both MDEFDW pumps running.

Technical Reference(s): EOP Rule 3 & Rule 7 Rev 36 Proposed references to be provided to applicants during examination: NONE Learning Objective: EAP-LOHT Obj. R26, R27 Question Source: N Question History: Last NRC Exam _ _ _ __

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis

2009 NRC REACTOR OPERATOR EXAM 1 POINT Question 45 Unit 1 Plant Conditions:

• Reactor power = 100%

• Pressurizer level = 219 inches and stable

• 1B 1 Rep parameters:

- Lower seal cavity pressure = 900 psig decreasing

=

- Upper seal cavity pressure 100 psig decreasing

=

- Seal Return flow 1.1 gpm decreasing

- Seal Leakage flow = 0.9 gpm increasing Based on the above conditions, which ONE of the following 1 B1 RCP seal(s) are failing?

A. Lower ONLY B. Middle ONLY C. BOTH Lower and Upper D. BOTH Middle and Upper

2009 NRC REACTOR OPERATOR EXAM Question 45 T2/G1 *okm 003K6.02, Reactor Coolant Pump System Knowledge of the effect of a loss or malfunction of the following will have on the RCPS components: RCP Seals and Seal Water Supply (2.7/3.1)

KIA MATCH ANALYSIS Requires knowledge and comprehension of the indications of a loss of RCP seals ANSWER CHOICE ANALYSIS Answer: 0 A. Incorrect: If ONLY the Lower seal failed, then Upper cavity pressures would increase to - 1025 psig and lower cavity pressure would increase to - 2150 psig.

B. Incorrect: If ONLY the middle seal failed, then upper and lower cavity pressures would equalize to -1050 psig. Seal Return (SR) decreasing and Seal Leakage (SL) increasing is an indication that the upper seal is also failing. With only the middle seal failed, SR would increase. SL would hardly change at all.

C. Incorrect: If the lower and upper seals both failed, upper cavity pressure would decrease to -atmospheric (Same as given parameter change) and lower cavity pressure would increase up to -2150 psig (opposite from given parameter change).

Seal return (SR) flow would decrease as upper cavity pressure decreases due to greater back pressure from the LOST than atmospheric or the QT. SR flow would become Seal Leakage flow.

O. Correct: The middle and upper seals both failed. The dP across the lower seal increases to 2150 psig and both the lower and upper cavity pressures decrease toward atmospheric or QT. Seal return flow decreases as upper cavity pressure decreases due to greater back pressure from the LOST than to atmospheric or to the QT. Seal Leakage flow increases as Seal Return flow becomes leakage flow.

Technical Reference(s): PNS*CPS Proposed references to be provided to applicants during examination: NONE Learning Objective: PNS*CPS R20 Question Source: B (PNS072006)

Question History: Last NRC Exam _ _ _ __

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis

2009 NRC REACTOR OPERATOR EXAM 1 POINT Question 46 Unit 1 initial conditions:

• =

Reactor power 25% power

• 1T Amps = 0

• CT1 Amps =2000

• Keowee Unit 1 is Out of Service

• Central Switchyard is energizing the STBY Buses

• PCB 17 (OCONEE WHo STARTUP TRANS. CT1 TIE) is open for maintenance Current conditions:

• Yellow Bus lockout occurs Based on the conditions above, which ONE of the following alignments will supply power to Unit 1's Main Feeder Buses?

A. 1T (Unit 1 Auxiliary Transformer)

B. CT-1 (Unit 1 Startup Transformer)

C. CT-4 (Keowee Underground to STBY Buses Transformer)

D. CT-5 (1 OOKV line to STBY Buses)

2009 NRC REACTOR OPERATOR EXAM Question 46 T2/G1 062A3.05, AC Electrical Distribution System Ability to monitor automatic operation of the ac distribution system, including:

Safety-related indicators and controls (3.5/3.6)

KIA MATCH ANALYSIS Requires ability to determine auto transfer of offsite electrical power to the onsite 4160 VAC distribution system following a yellow bus lockout ANSWER CHOICE ANALYSIS Answer: D A. Incorrect: Plausible if fail to recognize that conditions given indicate that 1T is not currently supplying power to the MFBs B. Incorrect: Plausible as CT-1 could be energized via the RED bus if PCB-17 was closed/operable and this would be the first choice of power after a unit trip C. Incorrect: Plausible as CT-4 would be the preferred source of power to the standby buses normally, however since CT- 5 is already supplying the bus CT-4 will not.

D. Correct: If the Red bus path is not available and Central path is established this is the power path since the SL breakers are closed therefore the SK breakers can NOT close Technical Reference(s): SA*3/C3, OP/0/1108/01 Encl 3.37, EL*EPD, OP-OC*EL*EPD*2 Proposed references to be provided to applicants during examination: None Learning Objective: EL*EPD Obj R17 & 20 Question Source: N Question History: Last NRC Exam _ _ _ __

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis

2009 NRC REACTOR OPERATOR EXAM 1 POINT Question 47 Initial conditions:

• 80th Keowee Units generating to the grid at ~ 60 MWs

• AC8-4 closed Current conditions:

• A S8LOCA occurs on Unit 2

• When ES channels 1&2 actuate

- Keowee Unit #2 Emergency locks out

- Keowee Main Step-Up Transformer locks out Within thirty (30) seconds of these actions, which ONE of the following Keowee breaker combinations should exist?

A. AC8-1 open and AC8-2 closed

8. AC8-1 closed and AC8-3 closed C. AC8-3 open and AC8-4 closed D. AC8-3 closed and AC8-4 open

2009 NRC REACTOR OPERATOR EXAM Question 47 T2/G1 062K1.02, AC Electrical Distribution System Knowledge of the physical connections and/or cause effect relationships between the ac distribution system and the following systems: ED/G (4.1/4.4)

KIA MATCH ANALYSIS Requires knowledge of KHU Step-up Transformer lockout and Emergency lockout of KHU and determination of resulting lineup to electrical distribution system ANSWER CHOICE ANALYSIS Answer: D A. Incorrect - ACB-2 can't close KHU-2 locked out. Plausible if interface between ES start and emergency lockout is misinterpreted B. Incorrect - ACB-1 won't close because Keowee Main Xfmer LO signal C. Incorrect - ACB-3 closes 8.5 sec after the Emerg start signal (Zone overlap Protection). KHU-2 Emergency Lock out opens ACB-4 and prevents auto closure D. Correct - ACB-3 closes 8.5 sec after the Emerg start signal if ACB-4 opens due to Emerg LO and the Keowee Main Xfmer is Locked out thus completing the requirements for Zone Overlap Protection Technical Reference(s): OP/0/A/2000/100 KHS Alarm Response Guide SA-1/A-1 OP/2/A/2000/102 KHU-2 Alarm Response Guide 2SA-2/A-3, EL-KHG Proposed references to be provided to applicants during examination: None Learning Objective: EL-KHG Obj. R10, R11 Question Source: B (EL0411 07)

Question History: Last NRC Exam _ _ _ __

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis

2009 NRC REACTOR OPERATOR EXAM 1 POINT Question 48 c ONS conditions:

• Station Blackout exists

• DC inverters (all) input voltages decreasing Based on the above conditions; which ONE of the following describes actions directed per the Blackout tab of the EOP and why?

Each unit's ____ inverters are de-energized to _ _ __

A. DIA, DIB, DIC, and DID / prevent inverter damage due to high current B. DIA, DIB, DIC, and DID / extend available battery life C. KI, KX, KU, and KOAC / prevent inverter damage due to high current D. KI, KX, KU, and KOAC / extend available battery life

2009 NRC REACTOR OPERATOR EXAM Question 48 T2/G1 063A 1.01, DC Electrical Distribution System

( Ability to predict and/or monitor changes in parameters associated with operating the DC electrical system controls including: Battery capacity as it is affected by discharge rate (2.5/3.3)

KIA MATCH ANALYSIS Requires knowledge of the correlation of the impact of battery load on available capacity as the bases for actions directed in the EOP ANSWER CHOICE ANALYSIS Answer: D A. Incorrect: Plausible in that only Essential Inverters (KI,KU, KX, KOAC) are de-energized; wrong also in that the reason is to extend battery life; plausible in that inverters could be damaged due to high current as input voltages start to decrease.

B. Incorrect: The reason is correct; plausible in that only Essential Inverters are de-energized. Vital Inverters (such as DID) are kept in service to supply such things as Nls and safety related instrumentation.

C. Incorrect: Plausible in that Essential Inverters (KI,KU, KX, KOAC) are de-energized; wrong in that the reason is to extend battery life; plausible in that inverters could be damaged due to high current as input voltages start to decrease D. Correct: KI, KU, & KX DC input breakers are opened to extend battery life.

Technical Reference(s): Rev 36 of EP/1/Al1800/01, EAP-SBO Proposed references to be provided to applicants during examination: None Learning Objective: EAP-SBO Obj. R8, EL-DCD Obj R1 Question Source: B (EAP220801)

Question History: Last NRC Exam _ _ _ __

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis

(

2009 NRC REACTOR OPERATOR EXAM 1 POINT Question 49 Operators are preparing to synchronize KHU-1 to the grid per OP/O/A/11 06/019, Keowee Hydro At Oconee The operator notes the following indications:

• Keowee 1 Line Volts = 13.8 kV

• Keowee 1 Output Volts = 13.4 kV

=

• Grid Frequency 60 cycles

• Keowee Frequency = 60.5 cycles Based on the above conditions, which one of the following correctly describes the synchroscope indication prior to closing ACB-1 and the control that will be used to correct this condition?

The synchroscope will be rotating "fast" in the direction and the _ _ __

control switch that will be used to adjust the synchroscope indication.

A. clockwise 1 UNIT 1 AUTO VOLTAGE ADJUSTER B. clockwise 1 UNIT 1 SPEED CHANGER MOTOR C. counterclockwise 1 UNIT 1 AUTO VOLTAGE ADJUSTER D. counterclockwise 1 UNIT 1 SPEED CHANGER MOTOR

2009 NRC REACTOR OPERATOR EXAM Question 49 T2/G1

( 064A2.03, Emergency Diesel Generator (ED/G) System

" Ability to (a) predict the impacts of the following malfunctions or operations on the ED/G system; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Parallel operation of ED/Gs (3.1/3.1)

KIA MATCH ANALYSIS Requires knowledge of paralleling operations and actions/controls needed to correct and control paralleling operations ANSWER CHOICE ANALYSIS Answer: B A. Incorrect. Synchroscope would rotate in clockwise direction as incoming frequency is KHU Frequency which is higher than line frequency. KHU Voltage being below Line voltage will result in negative MVARs when the ACB is closed; however to correct the difference in frequencies the MOTOR SPEED CHANGER is adjusted. The AUTO VOLTAGE ADJUSTER would be used to correct the (-) MVARS.

B. Correct. Synchroscope would rotate in clockwise direction as (the higher) incoming frequency is KHU Frequency; use the MOTOR SPEED CHANGER to adjust. KHU Voltage being below Line voltage will result in negative MVARs when the ACB is closed.

C. Incorrect. Synchroscope would rotate in clockwise direction as incoming frequency is KHU Frequency which is higher than line frequency. KHU Voltage being below Line voltage will result in negative MVARs when the ACB is closed. Plausible if voltage and frequency indication relationship to incoming and running status is misapplied.

The AUTO VOLTAGE ADJUSTER would be used to correct the (-) MVARS not the synchroscope/frequency concern.

D. Incorrect. Synchroscope direction is incorrect as noted above. Correct: The MOTOR SPEED CHANGER is used to adjust synchroscope rotation.

Technical Reference(s): EL-KHG, OP/0/Al1106/019 Rev 83 Proposed references to be provided to applicants during examination: None Learning Objective: EL-KHG R11, R4, R20, R19, R7 Question Source: M (EL04111 0)

Question History: Last NRC Exam 2006 ONS RO NRC Exam Q#58 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis

2009 NRC REACTOR OPERATOR EXAM 1 POINT Question 50 Which ONE of the following describes a function provided by Process Radiation Monitors?

A. Monitor gamma radiation levels from radioactive materials for Radiation Protection use in dose planning (ALARA).

B. Monitor and control releases of radioactive materials to the environment during actual or potential releases.

C. Alert onsite personnel to unplanned increases in radiation levels in the plant to prevent exceeding exposure limits.

D. Provide post accident indication of localized radiation levels to the Control Room for use in assessment of plant conditions.

2009 NRC REACTOR OPERATOR EXAM Question 50 T2/G1 073G2.1.27, Process Radiation Monitoring (PRM) System Knowledge of system purpose and/or function (3.9/4.0)

KIA MATCH ANALYSIS Requires knowledge of the functions provided by the Process Radiation monitoring system ANSWER CHOICE ANALYSIS Answer: B A. Incorrect: Plausible as this is a function provided by Area Radiation Monitors.

B. Correct. Per lesson RAD-RIA one of the functions provided by Process Radiation Monitors is to "Assure releases do not exceed 10CFR20 limits". Per SLC 16.11.3 Bases "The radioactive ... effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in ...

effluents during actual or potential releases."

C. Incorrect: Plausible as this is a function provided by Area Radiation Monitors.

D. Incorrect: Plausible as this is a function provided by Area Radiation Monitors.

Technical Reference(s): SLC 16.11.3 & RAD-RIA Proposed references to be provided to applicants during examination: None Learning Objective: RAD-RIA Obj R1 Question Source: N Question History: Last NRC Exam _ _ _ __

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis

2009 NRC REACTOR OPERATOR EXAM 1 POINT Question 51 Unit 1 initial conditions:

• Reactor power 100% =

Current conditions:

• A, B, and C LPSW pumps tripped

• AP/24, Loss of LPSW in progress Based on the above conditions, which ONE of the following conditions will result in EOP entry?

A. Pressurizer level =278 inches B. Any CRD stator temperature =192 of C. Any RCP Motor Stator temperature =265 of D. Main Turbine journal bearing #9 vibration = 17 mils

(

2009 NRC REACTOR OPERATOR EXAM Question 51 T2/G1 076G2.4.2, Service Water System Knowledge of system set points, interlocks and automatic actions associated with EOP entry conditions.

(4.5/4.6)

KIA MATCH ANALYSIS Requires knowledge of criteria/setpoints in abnormal procedures which require EOP entry ANSWER CHOICE ANALYSIS Answer: D A. Incorrect - Plausible in that loss of CC will occur which will cause a loss of Letdown and Pzr level to increase. Pzr Level >375 inches is Rx trip criteria B. Incorrect - Plausible in that loss of CC will occur but wrong in that two CRD Stators (not any; one) must be ~ 180 OF per AP/20 which can be initiated from AP/24 C. Incorrect - Plausible per AP/24, is correct but the setpoint for RCP Motor Stator (T) is> 295 OF.

D. Correct - Per AP/24, if any Main Turbine journal bearing vibration (bearings 1-

10) is > 12 mils, THEN trip the Rx and trip the Turbine Generator

(

Technical Reference(s): AP/24 Rev 23, AP/20 Rev 9 Proposed references to be provided to applicants during examination: None Learning Objective: EAP-APG Obj R9 Question Source: B (EAP210966)

Question History: Last NRC Exam _ _ _ __

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis

2009 NRC REACTOR OPERATOR EXAM 1 POINT Question 52 Unit 3 initial conditions:

• Reactor Power =100%

• 3A LPSW Pump =running

• LPSW Line leak occurs Current conditions:

• Unit 3 LPSW Pressure =60 psig decreasing slowly Based on the conditions above, which ONE of the following describes the status of the Unit 3 LPSW Pumps and the appropriate action per AP/24 (Loss of LPSW)?

A. ONLY 3A LPSW pump is running / start LPSW pump 3B B. ONLY 3A LPSW pump is running / reduce LPSW loads as needed C. BOTH 3A and 3B LPSW pumps are running / reduce LPSW loads as needed D. BOTH 3A and 3B LPSW pumps are running / align Station ASW to the HPI pumps

2009 NRC REACTOR OPERATOR EXAM Question 52 T2/G1 076K4.02, Service Water System Knowledge of SWS design feature(s) and/or interlock(s) which provide for the following: Automatic start features associated with SWS pump controls (2.9/3.2)

KIA MATCH ANALYSIS Requires knowledge of the LPSW pump auto start setpoint and status with the Standby LPSW Pump Auto Start Circuit in DISABLE and the action required if pressure is below the setpoint.

ANSWER CHOICE ANALYSIS Answer: C A. Incorrect: Pump status is incorrect in that both LPSW pumps are running because current LPSW pressure is below the Auto Start signal setpoint of 70 psig. Action is consistent with a failure of the pump to start and is consistent with the direction in AP/24.

B. Incorrect: Pump status is incorrect as described above but plausible if Auto start interlock is misapplied or setpoint is assumed incorrectly. Action in second part is correct.

C. Correct: Both LPSW pumps are running because current LPSW pressure is

( below the Auto Start signal setpoint of 70 psig. Action directed if LPSW pressure remains below normal is to reduce LPSW loads.

D. Incorrect: Pump status is correct as noted above. Current LPSW pressure is below the Auto Start signal setpoint of 70 psig. Action is incorrect but plausible as HPSW is the backup source for cooling the HPI pumps via an automatic transfer when LPSW pressure is below HPSW pressure. Station ASW is the second backup source and it is manually aligned.

Technical Reference(s): AP/1/A/1700/024 Rv 23, AP/3/A11700/024 Rv 22, 3SA-9/A-9 Proposed references to be provided to applicants during examination: None Learning Objective: SSS-LPW Obj R23 Question Source: N Question History: Last NRC Exam _ _ _ __

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis

2009 NRC REACTOR OPERATOR EXAM 1 POINT Question 53

( Initial conditions:

• IA Pressure = 105 psig stable

• IA-2718 (Air Supply to Radwaste Facility) Open

• Radwaste Air pressure 78 psig (and stable)

• Instrument Air Compressors are aligned as follows:

- Primary IA compressor is Operating

- Backup IA compressors "A" and "B" in Standby 1

- Backup IA compressor "C" in Standby 2

- Auxiliary IA compressors in Auto Current conditions:

• IA pressure =91 psig stable Based on the conditions above, which ONE of the following describes the response of the IA system?

A. Only Auxiliary IA Compressors start B. Only Backup IA compressors "A" & "B" start C. ALL Backup IA compressors start; Auxiliary IA Compressors start D. ALL Backup IA compressors start; IA-2718 (Air Supply to Radwaste Facility)

CLOSES

2009 NRC REACTOR OPERATOR EXAM Question 53 T2/G1 078A3.01, Instrument Air System (lAS)

( Ability to monitor automatic operation of the lAS, including: Air pressure (3.1/3.2)

KIA MATCH ANALYSIS Requires ability to predict IA system automatic response to a lowering IA system pressure ANSWER CHOICE ANALYSIS Answer: B A. Incorrect: The AlA compressors do not start until AlA receiver pressure reaches 88 psig. the stem identifies pressure at 91 psig.

B. Correct: ONLY the 'A' and 'B' BtU IA compressors will start and they started at 93 psig.

C. Incorrect: Same as 'A' above for B/U instrument air compressors. The AlA compressors do not start until AlA receiver pressure reaches 88 psig.

D. Incorrect: See 'B' above, IA-2718 (Air Supply to Radwaste Facility) closes at IA pressure below 85 psig.

Technical Reference(s): APt22 Rv 35, SSS-IA Proposed references to be provided to applicants during examination: None Learning Objective: SSS-IA Obj. R8 & R39 Question Source: B (555040801)

Question History: Last NRC Exam _ _ _ __

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis

2009 NRC REACTOR OPERATOR EXAM 1 POINT Question 54

( Unit 2 initial conditions:

Time = 1011 at 0700

• =

Reactor power 100%

• RB pressure = -0.25 psig

• PT/2/A/06001001 (Periodic Instrument Surveillance) directs increasing RB pressure to 0 psig Current conditions:

Time = 1011 at 0800

• IA-91 (Inside RB IA isolation) opened Based on the above conditions, opening IA-90 (Outside RB IA isolation) requires an operator to be ...

A. in constant communication with control room and stationed locally; IA-90 may remain open indefinitely.

B. stationed locally with contingency actions identified to ensure containment operability; IA-90 may remain open indefinitely.

C. stationed locally with contingency actions identified to ensure containment operability; IA-90 must be closed by 10/1 at 1200.

D. in constant communication with control room and stationed locally; IA-90 must be closed by 1011 at 1200.

2009 NRC REACTOR OPERATOR EXAM Question 54 T2/G1 078K1.03, Instrument Air System (lAS)

( Knowledge of the physical connections and/or cause-effect relationships between the lAS and the following systems: Containment air (3.3/3.4)

KIA MATCH ANALYSIS Requires knowledge of physical relationship between IA system and containment (RB) and the requirements associated with aligning IA to the RB during plant operation ANSWER CHOICE ANALYSIS Answer: D A. Incorrect: First part is correct. Valve only allowed to be opened for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> while in Mode 1-4 per TS 3.6.3 B. Incorrect, TS 3.6.3 requires constant communications, contingency action is to be stationed locally. Valve only allowed to be opened for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> while in Mode 1-4.

C. Incorrect: TS 3.6.3 requires constant communications, contingency action is to be stationed locally, not to ensure containment operability. Time limit is correct.

D. Correct. TS 3.6.3 requires constant communication. Contingency action is to be stationed locally and valve only allowed to be opened for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> while in Mode 1-4.

Technical Reference(s): OP/O/A/1106/027 Rev 94, Enc14.39 Proposed references to be provided to applicants during examination: None Learning Objective: SSS-IA Obj R14 Question Source: B (SSS041411)

Question History: Last NRC Exam _ _ _ __

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis

2009 NRC REACTOR OPERATOR EXAM 1 POINT Question 55

{

\ Unit 1 Initial conditions:

• Reactor Power = 100%

• Reactor Building pressure = (-)0.3 psig

• Reactor Building avg temperature = 125°F increasing slowly

• RBCU Testing in progress

• RBCU Status:

- 1A - Low Speed

- 1 B - Off

- 1C - Off Current conditions:

• RBCU Status:

- 1A - High Speed

- 1B - Off

- 1C - High Speed Based on the conditions above, which ONE of the following describes the response of RB Pressure and the operational concern associated with the pressure response?

RB Pressure will _ _ and the concern is _ _ _ __

A. increase / exceeding the maximum allowable pressure limit assumed to support operation of the 1A and 1 B LPI Pumps on the Emergency Sump.

B. increase / exceeding the maximum allowable pressure limit assumed to support integrity of containment following a LOCA.

C. decrease / going below the minimum allowable pressure limit assumed to support operation of the 1A and 1 B LPI Pumps on the Emergency Sump.

D. decrease / going below the minimum allowable pressure limit assumed to support operation of instrumentation in the Reactor Building

2009 NRC REACTOR OPERATOR EXAM Question 55 T2/G1 f 103A 1.01, Containment System

\ Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the containment system controls including: Containment pressure, temperature, and humidity (3.7/4.1)

K/A MATCH ANALYSIS Requires the ability to predict the impact of changing RBCU configuration on RB pressure due to the resulting change in temperature and the operating limit that is affected by the change ANSWER CHOICE ANALYSIS Answer: C A. Incorrect: Pressure response is incorrect. Plausible as raising fan speed would increase discharge pressure of the cooler units however the overall RB pressure response would be a reduction in pressure due to the reduction in temperature in the building. The impact of the response is plausible as it is an inverse application of the limit approached as pressure is reduced.

B. Incorrect: Plausible as raising fan speed would increase discharge pressure of the cooler units however the overall RB pressure response would be a reduction in pressure due to the reduction in temperature in the building. Impact is plausible as it is the limit that would be approached if RB pressure increases.

C. Correct: Per OP/1/A/1104/015 "When starting/stopping RBCUs or changing LPSW flows, RB pressure will change as RB temperature changes. For example: It only takes:::: 12°F reduction in average air temperature to go from -

0.2 psig to -0.5 psig." The operation limit is correct per SLC 16.6.13 D. Incorrect: Pressure impact is correct. Per OP/1/A/11 04/015 "When starting/stopping RBCUs or changing LPSW flows, RB pressure will change as RB temperature changes. For example: It only takes::::: 12°F reduction in average air temperature to go from -0.2 psig to -0.5 psig." Operation impact is incorrect. Plausible as RB pressure is used to apply a correction to instrumentation based on RB Pressure during elevated pressure conditions.

Technical Reference(s):OP/1/A/1104/015 Rv39 , PT/1/A/0600/001 Rv304, SLC 16.6.13 Proposed references to be provided to applicants during examination: None Learning Objective: PNS-RBC Obj R20, ADM-TSS Obj R4 Question Source: N Question History: Last NRC Exam _ _ _ _ __

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis

2009 NRC REACTOR OPERATOR EXAM 1 POINT Question 56 Unit 1 initial conditions:

• RCS Cooldown is in progress

• RCS temperature =240°F

• RCS pressure = 260 psig

• PORV Setpoint Selector is in LOW Current conditions:

• RC LR PRESS ENABLE Switch is placed in OFF Based on the conditions given, which ONE of the following describes the operation of the PORV?

A. PORV will operate normally B. PORV will immediately open C. PORV will NOT open automatically OR manually D. PORV will NOT open automatically but can be opened manually

2009 NRC REACTOR OPERATOR EXAM Question 56 T2/G2 002K4.10, Reactor Coolant System (RCS)

( Knowledge of RCS design feature(s) and/or interlock(s) which provide for the following: Overpressure protection (4.2/4.4 )

K/A MATCH ANALYSIS Requires knowledge of operation of the PORV during Low Temperature Overpressure Protection conditions ANSWER CHOICE ANALYSIS Answer: D A. Incorrect: Plausible if the candidate doesn't know that turning off the indicator also removes the pressure input to the PORV Circuit.

B. Incorrect: Immediate opening of the PORV is plausible if the removal of the signal is assumed to fail the PORV open.

C. Incorrect: First part is correct. PORV will not operate in AUTO as the pressure input is removed from the circuit with the LR Selector in OFF and the PORV selected to LOW. Second part is incorrect; Manual operation of the PORV is not affected.

Plausible since the selector switch is a part of the operating circuit.

D. Correct: PORV will not operate in AUTO as the pressure input is removed from the circuit if the LR Selector is in OFF with the PORV selected to LOW. Manual operation of the PORV is not affected.

Technical Reference(s): PNS-PZR Proposed references to be provided to applicants during examination: None Learning Objective: PNS-PZR R37 Question Source: N Question History: Last NRC Exam _ _ _ __

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis

2009 NRC REACTOR OPERATOR EXAM 1 POINT Question 57 Unit 1 Power Range NI indications are as follows:

• NI-5 = 98.8%

• NI-6 = 99.2%

• NI-7 = 99.3%

• NI-8 = 98.8%

• NI-9=99.1%

Based on the conditions above, which ONE of the following NI signals will be supplying the input to the Unit 1 Chessell NI Chart Recorder?

A. NI-5 B. NI-6 C. NI-7 D. NI-9

2009 NRC REACTOR OPERATOR EXAM Question 57 T2/G2 015A3.01, Nuclear Instrumentation System (NIS)

( Ability to monitor automatic operation of the NIS, including: Console and cabinet indications (3.8/3.8)

KIA MATCH ANALYSIS Require the ability to determine which NI channel will be auto selected to provide the signal to the control board chart recorder to indicate reactor power.

ANSWER CHOICE ANALYSIS Answer: B A. Incorrect: Recorder uses second highest of NI-5 through 8, NI-5 is not the second highest B. Correct: NI-6 is the second highest of NI-5 through 8 C. Incorrect: NI-7 is not the second highest of NI-5 through 8. Plausible as it is the highest (most conservative) NI indication.

D. Incorrect: NI-9 is no longer input to Chessell. Plausible as prior to the recent modification NI-5, 6 & 9 were median selected and NI-9 is the median of the 5,6 &

9. (Unit 2 used this method as of 10/22/08)

Technical Reference(s): IC-NI Proposed references to be provided to applicants during examination: None Learning Objective: IC-NI Obj. R24 Question Source: B (IC072402)

Question History: Last NRC Exam _ _ _ __

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis

2009 NRC REACTOR OPERATOR EXAM 1 POINT Question 58 Unit 1 initial conditions:

• Reactor power = 40%

Current conditions:

• Final Feedwater temperature controlling signal fails HIGH Based on the conditions above, which ONE of the following describes the initial plant response and the appropriate ICS station(s) used to stabilize the plant?

Actual Feedwater Flow will ----

. Place the Diamond Panel in Manual and -------

in Hand.

A. decrease / 1A & 1 B FDW Masters B. decrease / Steam Generator Master C. increase / 1A & 1 B FDW Masters D. increase / Steam Generator Master

(

(

2009 NRC REACTOR OPERATOR EXAM Question 58 T2/G2 016K3.04, Non-Nuclear Instrumentation System (NNIS)

Knowledge of the effect that a loss or malfunction of the NNIS will have on the following: MFW system (2.6/2.7)

K/A MATCH ANALYSIS Requires knowledge of the effect of a failed FW Temp instrument on the ICS control of the FDW System and the appropriate FDW controls to operate to stabilize the plant ANSWER CHOICE ANALYSIS Answer: C A. Incorrect: Plant response is incorrect. Plausible if the impact of the temperature compensation on the flow instrument is misapplied and assumption is that indicated flow would go down resulting in indicated flow being < FDW demand. Hand/Auto Stations identified in AP/28 are the FDW Masters only.

B. Incorrect: Plant response is incorrect. Plausible if the impact of the temperature compensation on the flow instrument is misapplied and assumption is that indicated flow would go down resulting in indicated flow being < FDW demand. SG Master Hand/Auto Station is plausible if the failed signal is incorrectly assumed to be applied upstream of the SG Master instead of upstream of the FDW Loop Masters.

C. Correct: FW Temp failing high will increase FDW Demand signal and raise actual FDW Flow. Hand/Auto Stations identified in AP/28 are the FDW Masters only.

D. FW Temp failing high will increase FDW Demand signal and raise actual FDW Flow.

SG Master Hand/Auto Station is plausible if the failed signal is incorrectly assumed to be applied upstream of the SG Master instead of upstream of the FDW Loop Masters.

Technical Reference(s): AP/28 Encl 4J, Rev 14 Proposed references to be provided to applicants during examination: None Learning Objective: STG*ICS Obj R14, SAE*L 087 Obj R17 & R4 Question Source: N Question History: Last NRC Exam _ _ _ __

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis

2009 NRC REACTOR OPERATOR EXAM 1 POINT Question 59 Which ONE of the following describes the Train A (or B) ICCM/RVLlS plasma display indication of CORE SCM for a superheated core and the Core Exit Thermocouples used to calculate this indication.

A. Reverse video with negative numbers I Average of the 5 highest of the 12 for that train of ICCM B. Red flashing negative numbers I Avg of the 5 highest of the 12 for that train of ICCM C. Reverse video with negative numbers I Average of the 5 highest of the 24 qualified D. Red flashing negative numbers I Average of the 5 highest of the 24 qualified

2009 NRC REACTOR OPERATOR EXAM Question 59 T2/G2-okm 017K5.02, In-Core Temperature Monitor System (ITM)

Knowledge of the operational implications of the following concepts as they apply to the ITM system: Saturation and subcooling of water (3.7/4.0)

K/A MATCH ANALYSIS Requires knowledge of ITM signal input to Core SCM indications and determining the state of the core coolant in relation to saturation conditions Knowledge of saturated conditions are being tested by having the candidate differentiate between saturated and superheated ITM indications ANSWER CHOICE ANALYSIS Answer: A A. CORRECT: Superheated conditions are correctly displayed by Red/Flashing/Negative Numbers on a reverse video background; ITM SCM indication comes from the Avg of the 5 Highest reading CETCs for each ICCM train B. Incorrect: 2 nd part is correct; 1st part is incorrect but plausible in that red flashing negative numbers are displayed for superheated core conditions on the OAC Standout digital SCM monitors which do not always provide qualified indications.

C. Incorrect: 1st part is correct; 2 nd part is incorrect but plausible in that the avg of the five highest of the 24 qualified CETCs is used for the OAC SCM display when >2%

power.

D. Incorrect: Both parts incorrect. 1st part is incorrect but plausible in that red flashing negative numbers are displayed for superheated core conditions on the OAC Standout digital SCM monitors which do not always provide qualified indications. 2nd part is incorrect but plausible in that the avg of the five highest of the 24 qualified CETCs is used for the OAC SCM display when >2% power.

Technical Reference(s): IC-RCI Proposed references to be provided to applicants during examination: None Learning Objective: IC-RCI R44 Question Source: B (IC084202)

Question History: Last NRC Exam _ _ _ __

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis

2009 NRC REACTOR OPERATOR EXAM 1 POINT Question 60 Unit 1 initial conditions:

• Reactor power = 100%

• 1KI is de-energized Current conditions:

• Rx trip Based on the conditions given, which ONE of the following describes the operation of the Turbine Bypass valves (TBVs)?

The TBVs ...

A. fail closed.

B. fail to 50% open.

C. will continue to function normally in Auto.

D. will be controlled manually from the control room.

2009 NRC REACTOR OPERATOR EXAM Question 60 T2/G2 041 K2.01, Steam Dump System (SDS)/Turbine Bypass Control Knowledge of bus power supplies to the following: ICS, normal and alternate power supply (2.8/2.9)

KIA MATCH ANALYSIS Requires knowledge that KI supplies ICS Auto power and the effect of a loss of ICS Auto power on the operation of the Turbine Bypass Valves (TBVs)

ANSWER CHOICE ANALYSIS Answer: 0 A. Incorrect: Plausible as this is the response to a loss of ICS Auto & Hand power.

B. Incorrect: Plausible as this is a common hand auto station response to a loss of Hand power while the station is in hand per the alarm response. The TBV controllers have Auto & Hand power supplied to the controller in Hand.

C. Incorrect: Plausible as this is the response to a loss of ICS Hand power.

D. Correct: Per 1SA-2/B-11 the TBV's will swap to hand and are operable in Hand only.

Technical Reference(s): 1SA-2/B-11, AP/23 Rev 17 Proposed references to be provided to applicants during examination: None Learning Objective: STG-ICS Obj R33 Question Source: N Question History: Last NRC Exam _ _ _ __

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis

2009 NRC REACTOR OPERATOR EXAM 1 POINT Question 61

( Unit 1 initial conditions:

• Mode 6

• Oefueling in progress

• 1 RIA-6 (Spent Fuel Pool Area Monitor) = 4 mr/hr stable Current conditions:

• 1 RIA-6 monitor power supply fuse blows

=

• 1 RIA-6 local reading 0 mr/hr

• 1 RIA-6 View Node indication is magenta Based on the conditions above, which ONE of the following describes the impact on fuel handling activities per OP/1/A/1502/007 (Operations Oefueling/Refueling Responsibilities )?

Fuel Handling activities in the SFP may ...

A. NOT continue until a replacement monitor is in place that is equivalent to 1 RIA-6.

B. continue because only the SFP Portable Bridge monitor is required.

C. NOT continue until continuous RP coverage is present on the SFP Bridge.

O. continue after a portable air sampling instrument is placed in the SFP area.

2009 NRC REACTOR OPERATOR EXAM Question 61 T2/G2 - jmb 072A2.03, Area Radiation Monitoring System

( 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 (2.7/2.9)

KIA MATCH ANALYSIS Requires ability to predict the impact of a blown power supply fuse on SFP Area Rad monitor on the alarm function and Fuel Handling activities ANSWER CHOICE ANALYSIS Answer: A A. Correct: FH activity must stop. Both monitors are required per OP/1/A11502/007 and SLC 16.12.2. The replacement monitor must be a portable instrument equivalent in range, sensitivity, and able to alarm locally B. Incorrect: Plausible if the incorrect assumption is made that only one the two monitors is required C. Incorrect: Plausible in that FH activity must stop. Incorrect in that continuous RP coverage would be advantageous but this does not procedural compliance for loss

( of local monitors D. Incorrect: Plausible if the requirements of the replacement monitor are not known Technical Reference(s): OP/1/A11502/007 Rev 81 and SLC 16.12.2.

Proposed references to be provided to applicants during examination: None Learning Objective: RAD-RIA Obj R9, ADM-TSS Obj R4 Question Source: N Question History: Last NRC Exam _ _ _ __

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis

2009 NRC REACTOR OPERATOR EXAM 1 POINT Question 62 Unit 1 initial conditions:

• Mode 6 Current conditions:

• FTC Level = 20.9 ft (Placard on FTC Wall) decreasing

• East fuel carriage is in the RB

• West fuel carriage is in the SFP

• Main Fuel Bridge in transit to the upender with a spent fuel assembly in the mast Based on the conditions above, which ONE of the following describes actions that are directed per AP/26 (Loss of Decay Heat Removal)?

Place the fuel assembly into _ _ _ _ and position the _ _ _ __

A. the East Upender / West Fuel Carriage to the RB B. the East Upender / East Fuel Carriage to the SFP C. ONLY the original core location / West Fuel Carriage to the RB D. ONLY the original core location / East Fuel Carriage to the SFP

2009 NRC REACTOR OPERATOR EXAM Question 62 T2/G2-jmb 034A1.02, Fuel Handling Equipment System (FHES)

Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the Fuel Handling System controls including: Water level in the refueling canal (2.9/3.7)

KIA MATCH ANALYSIS Requires ability to interpret changes in FTC & SFP Levels and determine the correct location to place a fuel assembly and the upenders to prevent exceeding radiation release and off-site dose limits per 10 CFR 100 ANSWER CHOICE ANALYSIS Answer: B A. Incorrect: East upender is the correct location. Procedure directs placing the carriages in the SFP to allow FTT Isolation valves to be closed.

B. Correct: Procedure directs placing the fuel assembly in transit into a safe location and specifies the upender or original/intended location and positioning the carriages in the SFP in preparation for closing the FTT Isolation valves.

C. Incorrect: Assembly location is correct but it is not the only location allowed by the procedure. Plausible as it places the assembly at an elevation that reduces the chance of uncovery and below the core. Procedure directs placing the carriages in the SFP to allow FTT Isolation valves to be closed.

D. Incorrect: Assembly location is correct but it is not the only location allowed by the procedure. Plausible as it places the assembly at an elevation that reduces the chance of uncovery and below the core. Carriage location is correct.

Technical Reference(s): AP/26 Rev 20, TS 3.9.6 Proposed references to be provided to applicants during examination: None Learning Objective: EAP-APG Obj R9, FH-FHS Obj R7 Question Source: N Question History: Last NRC Exam _ _ _ __

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis

2009 NRC REACTOR OPERATOR EXAM 1 POINT Question 63 Unit 2 initial conditions:

• Reactor power = 100% stable

• =

Generator MWs 890 Mwe

• =

Condenser vacuum 28.5 inches Hg stable Current conditions:

• Condenser vacuum = 24.5 inches Hg slowly decreasing

• AP-27 (Loss of Condenser Vacuum) in progress Based on the conditions above, which ONE of the following describes the impact on Reactor power and a required action in accordance with AP/27?

Reactor power will ...

A. increase / reduce power to decrease turbine exhaust steam load on condenser B. increase / start and align the Main Vacuum Pumps to increase vacuum in the condenser C. remain the same / reduce power to decrease turbine exhaust steam load on condenser D. remain the same / start and align the Main Vacuum Pumps to increase vacuum in the condenser

2009 NRC REACTOR OPERATOR EXAM Question 63 T2/G2 - jmb 055G2.2.44, Condenser Air Removal System (CARS)

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. (4.2/4.4)

KIA MATCH ANALYSIS Requires knowledge of the impact of reduced vacuum on reactor power and the action taken by the operator to address the impact ANSWER CHOICE ANALYSIS Answer: D A. Incorrect: Reactor power impact is incorrect. Plausible if the assumption is made that the ICS system is not in integrated mode operating with CTPD as the controlling signal. Action is plausible because this action would have a beneficial effect on Vacuum and because AP's frequently reduce power.

B. Incorrect: Reactor power impact is incorrect as noted above. Action is correct. AP/27 directs the operator to dispatch an operator to align the vacuum pumps and start them from the control room.

C. Incorrect: Reactor power response is correct. CTPD will control ICS and maintain reactor power constant. The action is incorrect as noted above.

D. Correct: CTPD will controllCS and maintain reactor power constant. AP/27 directs the operator to dispatch an operator to align the vacuum pumps and start them from the control room.

Technical Reference(s): 2/AP/27 Rev 3 Proposed references to be provided to applicants during examination: None Learning Objective: EAP-APG Obj R9 Question Source: N Question History: Last NRC Exam _ _ _ __

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis

2009 NRC REACTOR OPERATOR EXAM 1 POINT Question 64 Which ONE of the following describes the RIA status that will automatically terminate a Gaseous waste release on Unit 3 and the status of the Waste Gas Exhauster if it is running prior to the alarm?

A. 3RIA-37 (NORM WD Gas) AND 3RIA-38 (HIGH WD Gas) must alarm / Waste Gas Exhauster stops automatically.

B. 3RIA-37 (NORM WD Gas) OR 3RIA-38 (HIGH WD Gas) must alarm / Waste Gas Exhauster stops automatically.

C. 3RIA-37 (NORM WD Gas) AND 3RIA-38 (HIGH WD Gas) must alarm / operator must manually stop the Waste Gas Exhauster.

D. 3RIA-37 (NORM WD Gas) OR 3RIA-38 (HIGH WD Gas) must alarm / operator must manually stop the Waste Gas Exhauster.

2009 NRC REACTOR OPERATOR EXAM Question 64 T2/G2 - jmb 071 K1.06, Waste Gas Disposal System (WGDS)

Knowledge of the physical connections and/or cause effect relationships between the Waste Gas Disposal System and the following systems: ARM and PRM systems (3.1/3.1 )

KIA MATCH ANALYSIS Requires knowledge of cause/effect relationship between the PRM System and the WGD system during releases ANSWER CHOICE ANALYSIS Answer: B A. Incorrect: Either 37 or 38 will isolate the GWD release and stop the exhauster.

B. Correct: Either 37 or 38 will isolate the GWD release and stop the exhauster.

C. Incorrect: Either 37 or 38 will isolate the GWD release and stop the exhauster.

D. Incorrect: Either 37 or 38 will isolate the GWD release and stop the exhauster when 37 or 38 alarms.

Technical Reference(s): 3SA-8/B-9, OP/3/Al1104/018 Rev73 Proposed references to be provided to applicants during examination: None Learning Objective: RAD-RIA Obj 14 Question Source: B Question History: Last NRC Exam 2005 NRC RO Exam Q#52 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis

2009 NRC REACTOR OPERATOR EXAM 1 POINT Question 65 Unit 2 initial conditions:

• Reactor power 100% =

• Switchyard Isolate occurs Current conditions:

• Unit 2 MFB 1 & 2 energized Based on the conditions above, which ONE of the following describes the suction supply to LPSW and when the LPSW pumps will restart?

Unit 1 & 2 LPSW is supplied via the ECCW ___ and LPSW pumps will restart

____ after power is restored?

A. first siphon / immediately B. second siphon / immediately C. first siphon / 10 seconds D. second siphon / 10 seconds

2009 NRC REACTOR OPERATOR EXAM Question 65 T2/G2 - jmb 075A4.01, Circulating Water System Ability to manually operate and/or monitor in the control room:

Emergency/essential SWS pumps (3.2/3.2)

KIA MATCH ANALYSIS Requires ability to determine the impact of the loss of Circulating Water (CCW) due to the loss of power and the response of the ECCW and LPSW systems to the loss ANSWER CHOICE ANALYSIS Answer: C A. Incorrect: First siphon flowpath from the intake to the CCW crossover header is the correct source to LPSW. LPSW start time is incorrect. Plausible if time delay is assumed to start when pumps are lost or if time delay is not applied.

B. Incorrect: The suction source is incorrect. Plausible as it is the source to supply cooling to the condenser in this condition. LPSW start time is incorrect as noted above.

C. Correct: First siphon flowpath from the intake to the CCW crossover header is the correct source to LPSW. LPSW pumps will receive a start signal 10 seconds after power is restored due to low system pressure.

(

D. Incorrect. Incorrect: The suction source is incorrect. Plausible as it is the source to supply cooling to the condenser in this condition. LPSW pump start time is correct.

Technical Reference(s): OP/1/A/11 04/01 0 Rev 127, OP/1/A/11 04/012 Rev 76 Proposed references to be provided to applicants during examination: None Learning Objective: STG-CCW Obj R11, SSS-LPW Obj R23 Question Source: N Question History: Last NRC Exam _ _ _ __

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis

2009 NRC REACTOR OPERATOR EXAM 1 POINT Question 66

( Which ONE of the following describes the requirements in OMP 1-02, Rules of Practice to ensure a Motor Operated Valve that acts as a throttle valve closes?

The switch must be placed in the CLOSED position and held ...

A. until the CLOSED indication light is lit.

B. for a minimum of five seconds after the CLOSED indication light is lit.

C. for a minimum of three seconds after the CLOSED indication light is lit.

D. until the OPENED indication light is off and the CLOSED indication light is lit.

2009 NRC REACTOR OPERATOR EXAM Question 66 T3 - jmb G2.1.1, Conduct of Operations Knowledge of conduct of operations requirements.

(3.8/4.2)

KIA MATCH ANALYSIS Requires knowledge of requirements to operate MOVs that act as throttling valves per OMP 1-02, Rules of Practice ANSWER CHOICE ANALYSIS Answer: B A. Incorrect: Plausible for normal valve operations but not per guidance given in OMP 1-2 for operating MOVs that act as throttling valves B. Correct: Per OMP 1-2, hold the switch for a minimum of 5 seconds after receiving the open or closed indication.

C. Incorrect: Plausible but incorrect in that the requirement is 5 seconds D. Incorrect: Plausible for normal valve operations but not per guidance given in OMP 1-2 for operating MOVs that act as throttling valves Technical Reference(s): OMP 1-02, Rules of Practice Rev 72 Proposed references to be provided to applicants during examination: None Learning Objective: ADM-OMP Obj R6 Question Source: B ADM043601 Question History: Last NRC Exam _ _ _ __

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis

2009 NRC REACTOR OPERATOR EXAM 1 POINT Question 67

( Unit 3 plant conditions:

Time: 0340

• Night Shift RO makes an AUTOLOG entry Time: 0905

• Day Shift RO reviewing the AUTO LOG identifies a mistake in the entry made at 0340 Based on the above conditions, which ONE of the following describes how the AUTOLOG entry is corrected per OMP 2-2 (Unit Logs)?

Unit AUTOLOG corrections may be made ...

A. ONLY by the CR SRO by editing the original entry.

B. by any member of the CR Team by editing the original entry.

C. ONLY by the CR SRO by making a late entry that references the original entry.

D. by any member of the CR Team by making a late entry that references the original entry.

(

2009 NRC REACTOR OPERATOR EXAM Question 67 T3 - jmb G2.1.18, Conduct of Operations Ability to make accurate, clear, and concise logs, records, status boards, and reports.

(3.6/3.8)

KIA MATCH ANALYSIS Requires ability to determine the method of correcting an archived AUTOLOG entry and who may make the correction.

ANSWER CHOICE ANALYSIS Answer: D A. Incorrect. The time of the correction is past the end of the shift that the original entry was made indicating that the log has been archived. Per OMP 2-2, correction can be made to archived entries by a late entry that references the original entry. Plausible as this method of correction is allowed if the log has not been archived. There is no requirement for the correction to be made only by the CRSRO.

B. Incorrect. The time of the correction is past the end of the shift that the original entry was made indicating that the log has been archived. Per OMP 2-2, correction can be made to archived entries by a late entry that references the original entry. Plausible as this method of correction is allowed if the log has not been archived.

C. Incorrect. The method of correction is correct. There is no requirement for the correction to be made only.by the CRSRO.

D. Correct. Per OMP 2-2, correction can be made to archived entries by a late entry that references the original entry. The time of the correction is past the end of the shift that the original entry was made indicating that the log has been archived. The correction can be made by any member of the control room team.

Technical Reference(s): OMP 2-2 Rev 18 Proposed references to be provided to applicants during examination: None Learning Objective: ADM-OMP Obj R24 Question Source: M (ADM042401)

Question History: Last NRC Exam _ _ _ _ __

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis

2009 NRC REACTOR OPERATOR EXAM 1 POINT Question 68 c Which ONE of the following activities is consistent with the conservative operating guidance contained in SOMP 1-2 (Reactivity Management)?

A. Manual rod withdrawal during a Feedwater transient to stop a temperature decrease caused by an instrument failure B. Manually increasing Feedwater flow to stop an RCS pressure increase caused by an RCS temperature increase C. Manually raising one Loop FDW demand while lowering the other Loop FDW demand to control LlTcold following an RCP trip D. Manually increasing turbine demand to reduce Turbine Header Pressure and RCS temperature

2009 NRC REACTOR OPERATOR EXAM Question 68 T3 G2.1.39, Conduct of Operations Knowledge of conservative decision making practices.

(3.6/4.3)

KIA MATCH ANALYSIS Requires knowledge of conservative reactivity management actions allowed during plant transients.

ANSWER CHOICE ANALYSIS Answer: C A. Incorrect: Manual rod withdrawal is not permitted.

B. Incorrect: Increasing FDW Flow is not permitted.

C. Correct: The sequence given is permitted as there is no intent to raise FDW Flow.

D. Incorrect: Increase in Turbine demand is only allowed if intent is to stabilize Turbine Header Pressure not to reduce pressure or RCS temperature.

Technical Reference(s): SOMP 1-2 Rev 6 Proposed references to be provided to applicants during examination: None Learning Objective: ADM-OMP Obj R23, TA-PTR Obj R1 Question Source: N Question History: Last NRC Exam _ _ _ __

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis

2009 NRC REACTOR OPERATOR EXAM 1 POINT Question 69

( In accordance with OMP 1-02 (Rules of Practice), which ONE of the following describes a condition which would allow Independent Verification of a single valve to be waived and the minimum level of approval required?

=

A. Dose received will be 14 mr for a single check Any Operations supervisor B. Valve located in a room where the area dose rate = 878 mr/hr Any Operations supervisor

=

C. Dose received will be 14 mr for a single check Only the Operations section manager D. Valve located in a room where the area dose rate = 878 mr/hr Only the Operations section manager

2009 NRC REACTOR OPERATOR EXAM Question 69 T3 - okm c G2.2.14, Equipment Control Knowledge of the process for controlling equipment configuration or status.

(3.9/4.3)

KIA MATCH ANALYSIS Requires knowledge of the independent verification process associated with equipment configuration control ANSWER CHOICE ANALYSIS Answer: A A. CORRECT: Per OMP 1-2, IV waiver allowed for personnel dose if a single valve IV will result in a dose of> 10 mrem. Any Ops supervision can make this determination.

B. Incorrect: Plausible in that the requirement for waiver due to dose rate is >1 Rem/hr.

2nd part is correct.

C. Incorrect: 1st part is correct. 2 nd part incorrect; plausible because the Operations section manager is responsible for approval of dose extension above admin limits.

D. Incorrect: Both parts incorrect. First part is plausible in that the requirement for waiver due to dose rate is >1 Rem/hr. 2 nd part incorrect; plausible because the Operations section manager is responsible for approval of dose extension above admin limits ..

Technical Reference(s): OMP 1-2 pg. 11 Rev. 72 Proposed references to be provided to applicants during examination: NONE Learning Objective: ADM-OMP R6,36,37,39,40 Question Source: M (ADM060902)

Question History: Last NRC Exam _ _ _ __

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis

2009 NRC REACTOR OPERATOR EXAM 1 POINT Question 70

( Which ONE of the following describes two (2) evolutions or tests that have pre-planned pre-job briefs per NSD 213 (Risk Management Process), Infrequently Performed Tests or Evolutions?

A. Unit 2 Mid-Loop Operations / Turbine Stop Valve Movement Test B. Unit 2 Mid-Loop Operations / Zero Power Physics Testing C. Control Rod Movement Test / Turbine Stop Valve Movement Test D. Control Rod Movement Test / Zero Power Physics Testing

(

2009 NRC REACTOR OPERATOR EXAM Question 70 T3 - okm G2.2.7, Equipment Control Knowledge of the process for conducting special or infrequent tests (2.9/3.6)

K/A MATCH ANALYSIS Requires knowledge of the definition of Infrequently Performed Tests or Evolutions (IPTEs)

ANSWER CHOICE ANALYSIS Answer: B A. Incorrect: Plausible in that knowledge of which tests/evolutions are listed/described in NSD 213, IPTEs. Of the two only Mid-Loop Ops is listed.

B. CORRECT: Per NSD-213 (Risk Management Process) and as mentioned in OMP 1-22 (Pre-job and Post-job Briefs) , Mid-loop Operations and Zero Power Physics Testing are listed/meet the criteria for an Infrequently Performed Test/Evol ution C. Incorrect: Plausible in that both evolutions affect reactivity but are specified as NSD 213 IPTEs. CRD Movement tests are performed quarterly. TSV tests are done every month.

D. Incorrect: Plausible in that knowledge of which tests/evolutions are listed/described in NSD 213, IPTEs. Of the two, only Zero Power Physics Testing is listed.

Technical Reference(s): NSD 213 Rev 7, OMP 1-22 Rev 11 Proposed references to be provided to applicants during examination: NONE Learning Objective: ADM-OMP R11, R28 Question Source: NEW Question History: Last NRC Exam _ _ _ __

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis

2009 NRC REACTOR OPERATOR EXAM 1 POINT Question 71

( ,

Unit 3 plant conditions:

• Spent Fuel Demineralizer Room dose rate =3000 MR/HR Based on the above condition, which ONE of the following describes the radiological posting requirements and the access controls for this area?

A Locked High Radiation I area MUST be posted with a Yellow Flashing Light B. Locked High Radiation I entrance MUST be Locked or Guarded C, Very High Radiation I area MUST be posted with a Yellow Flashing Light D, Very High Radiation I entrance MUST be Locked or Guarded c

2009 NRC REACTOR OPERATOR EXAM Question 71 T3 - okm G2.3.12, Radiation Control Knowledge of radiological safety principles pertaining to licensed operator duties, such as containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc.

(3.2/3.7)

KIA MATCH ANALYSIS Requires knowledge of radiological postings and required RP access controls ANSWER CHOICE ANALYSIS Answer: 8 A. Incorrect: Room posting should be LHRA; wrong in that this room is able to be locked; a flashing yellow light is only used when a room/area can not be reasonably locked B. CORRECT: Room posting should be LHRA (>1 rlhr - to - <500 R/hr). Since the SF Demineralizer Room is able to be locked in the plant it must be locked.

Student must apply the fact that this specific room ,can be locked otherwise a flashing yellow light would also be correct.

C. Incorrect: Plausible if posting requirements are not known (VHRAs = >500 r/hr) also wrong in that VHRAs must be locked at all times.

D. Incorrect: Plausible if posting requirements are not known (VHRAs =>500 r/hr).

Plausible in that VHRAs must be locked at all times.

Technical Reference(s): RAD RPP, NSD-507 Rev 13 Proposed references to be provided to applicants during examination: NONE Learning Objective: RAD-RPP R8, R24 Question Source: NEW Question History: Last NRC Exam _ _ _ __

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis

2009 NRC REACTOR OPERATOR EXAM 1 POINT Question 72 Unit 1 plant conditions:

• Reactor power = 100%

• 7 gpm primary to secondary leak in 1A SG Based on the above conditions, which ONE of the following describes the location that will have the greatest increase in general area dose rates and the operators response if he receives a dose alarm in the area?

A. Powdex / Remain in the area and monitor dose B. Powdex / Immediately stop and leave the area C. TB Sump / Remain in the area and monitor dose D. TB Sump / Immediately stop and leave the area

2009 NRC REACTOR OPERATOR EXAM Question 72 T3 - okm G2.3.14, Radiation Control

( Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or emergency conditions or activities.

(3.4/3.8)

KIA MATCH ANALYSIS Requires knowledge of increased contamination hazards due to a SGTL and radiological safety practices ANSWER CHOICE ANALYSIS Answer: B A. Incorrect: Plausible in that the Powdex will have the highest general dose rates due to concentrating the SGTL activity. Wrong - in that EDLs are not in effect (not in EOP) therefore he cannot work through a dose alarm B. CORRECT: The Powdex will gather and concentrate the RCS activity making it have the highest general dose rates; since emergency dose limits (EDLs) are not in affect (not in EOP) the NEO is expected to regard all dose alarms C. Incorrect: Plausible in that the TB Sump will be an area for secondary contamination to collect but it will not have dose rates as high as the Powdex; wrong - also in that EDLs are not in effect therefore he cannot work through dose alarms D. Incorrect: Plausible in that the TB Sump will be an area for secondary contamination to collect but it will not have dose rates as high as the Powdex; since emergency dose limits (EDLs) are not in affect (not in EOP) the NEO is expected to regard all dose alarms Technical Reference(s): AP-31 Rev 15, OMP 1-18 Rev 27 Proposed references to be provided to applicants during examination: NONE Learning Objective: EAP-EOP R20, ADM-OMP R10, R52 Question Source: NEW Question History: Last NRC Exam _ _ _ __

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis

2009 NRC REACTOR OPERATOR EXAM 1 POINT Question 73 Plant conditions:

• All Units Reactor power = 100%

• 1SA3/B-6 (Fire Alarm) actuated

• NEO reports flames and heavy smoke spreading to equipment and cable trays

• Fire location = Near the LPSW pumps, Column G30 SEE ATTACHMENT Based on the above information, which ONE of the following describes required actions?

Fire Location is an SSF Risk Area for _ _ / Perform a manual Rx Trip on _ _

A. Unit 1 ONLY / Unit 1 ONLY B. Unit 2 ONLY / Unit 2 ONLY C. Unit 1 ONLY / Unit 1 and 2 D. Unit 2 ONLY / Unit 1 and 2

2009 NRC REACTOR OPERATOR EXAM Question 73 T3 - okm G2.4.27, Emergency Procedures/Plans Knowledge of "fire in the plant" procedures.

(3.4/3.9)

KIA MATCH ANALYSIS Requires knowledge of the Fire Alarms, Fire Plan usage, and AP/25 SSF actions for an in-plant fire ANSWER CHOICE ANALYSIS Answer: B A. Incorrect: Plausible in that the fire is located partly on Unit 1 and 2 but the Attachment 1-fire plan dictates the fire as Unit 2 SSF Risk Area ONLY. Candidate must know manual trip criteria from AP/25 and trip only the affected unit (Unit 2 only)

B. CORRECT: Per ARG for 1SA3/B-6 the fire is a Challenging Active Fire; its location is between Unit 1 and Unit 2; Attachment 1 (provided) dictates the fire as Unit 2 SSF Risk Area ONLY. AP/25 requires only Unit 2 to be manually tripped.

C. Incorrect: Plausible if the fire plan is misread. Unit 3 has the largest amount of designated TBB SSF Risk Area designated; also candidate must know the units to be manually tripped D. Incorrect: Plausible if the fire plan is misread and the ALL 3 UNITS area is picked which would also lead to wrongly tripping all three units.

Technical Reference(s): 1SA3/B-6, AP/25 Rev 41 Proposed references to be provided to applicants during examination: 1SA3/B-6 Attachment 1 Body and the two enclosed fire plan maps Learning Objective: EAP-SSF R10 Question Source: NEW Question History: Last NRC Exam _ _ _ __

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis

(

2009 NRC REACTOR OPERATOR EXAM 1 POINT Question 74 Unit 1 plant conditions:

• Reactor tripped from 100% power

• The following Statalarms actuate:

- 1SA-1 /C-11 (ES Channel 7 Trip) 1SA-1/0-11 (ES Channel 8 Trip) 1SA-2/C-4 (RC Pressurizer Level Emerg High/Low) 1SA-2/C-8 (AFIS Header A Initiated) 1SA-2/0-5 (HP LOST Level Interlock Initiated) 1SA-8/A-3 (FOWPT A Trip) 1SA-8/A-6 (FOWPT B Trip)

Based on the above conditions, which ONE of the following emergency procedures has the highest priority?

A. Enclosure 5.1 (ES Actuation)

B. Rule 5 (Main Steam Line Break)

C. Enclosure 5.5 (pzr and LOST Level Control)

O. Rule 3 (Loss of Main or Emergency Feedwater)

2009 NRC REACTOR OPERATOR EXAM Question 74 T3 - okm G2.4.45, Emergency Procedures/Plans Ability to prioritize and interpret the significance of each annunciator or alarm (4.1/4.3)

KIA MATCH ANALYSIS Requires the ability to interpret the significance of statalarms, to diagnose the event, and to prioritize EOP rules and enclosures ANSWER CHOICE ANALYSIS Answer: B A. Incorrect: Plausible in that ES Actuation of all 8 Channels has most likely occurred B. CORRECT: AFIS actuation tells the operator that MSLB has occurred which is the highest priority due to the overcooling it can cause; in fact EHT has caused all the other SA conditions C. Incorrect: Plausible in that overcooling is causing the RCS inventory to contract to the point of possibly emptying the Pzr and LOST; however the overcooling would be stopped by Rule 5.

O. Incorrect: Plausible in that AFIS has caused the MFWPs to trip; a loss of both MFW and EFW would be a higher priority than Rule 5 but there is no reason to assume that EFW pumps are unavailable.

Technical Reference(s): EP/1/A/1800/001 Rev 36 Proposed references to be provided to applicants during examination: NONE Learning Objective: EAP-EOP R27 Question Source: N Question History: Last NRC Exam _ _ _ __

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis

2009 NRC REACTOR OPERATOR EXAM 1 POINT Question 75

( Unit 3 initial conditions

• =

Time 0700

• =

Reactor power 100%

• =

LOST level 82 inches stable

• LOST pressure = 29 psig stable Current conditions:

• Time = 0710

=

• LOST level 82 inches stable

• LOST pressure = 26 psig decreasing Based on the conditions above, which ONE of the following states the time the LOST will reach an operating limit and action required?

SEE ATTACHMENT A. 13 minutes / both trains of the HPI system must be declared inoperable B. 13 minutes / 3HP-24 & 25 must be opened if a transient requiring additional HPI flow occurs

( C. 20 minutes / both trains of the HPI system must be declared inoperable O. 20 minutes / 3HP-24 & 25 must be opened if a transient requiring additional HPI flow occurs

(

2009 NRC REACTOR OPERATOR EXAM Question 75 T3 - jmb G2.4.47, Emergency Procedures/Plans Ability to diagnose and recognize trends in an accurate and timely manner utilizing the appropriate control room reference material.

(4.2/4.2)

KIA MATCH ANALYSIS Requires the ability to diagnose trends, to apply the trend to reference material, and knowledge of corrective actions ANSWER CHOICE ANALYSIS Answer: D A. Incorrect: First part is incorrect. Plausible because the calculation is based on applying the divisions of the curve incorrectly and using the 2nd division instead of the first (84" vs 82") 3 psig/10 minutes =.3 psig/min limit =22 psig @84 inches current pressure 26 psig - limit value of 22 psig =4 psig. 4 + 0.3 =13.3 minutes.

Action given is incorrect and plausible as it is the action directed if pressure is outside the curve high vs low.

B. Incorrect: First part is incorrect as noted above. Second part is correct.

C. Incorrect: First part is correct. 3 psig/ 10 minutes = .3 psig/min limit = 20 psig @82 inches current pressure 26 psig - limit value of 20 psig =6 psig. 6 + 0.3 =20 minutes. Second part is incorrect as noted above.

=

D. Correct: First part is correct. 3 psig/1 0 minutes .3 psig/min limit 20 psig =

=

@82 inches current pressure 26 psig - limit value of 20 psig 6 psig. 6 + 0.3 =

20 minutes. Action given is correct per Encl. 3.39 (pg 2) compensatory measures are noted to immediately open 3HP-24 & 25 if a transient requiring additional HPI flow occurs.

Technical Reference(s): Rev 78 of OP/0/Al1108/001 Enclosure 4.39 Proposed references to be provided to applicants during examination: OP/0/Al11 08/001 .39 (page 1 only with action notes removed)

Learning Objective: PNS-HPI Obj. R35 Question Source: N Question History: Last NRC Exam _ _ _ __

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis