ML103500258

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Final Written Examination with Answer Key (401-5 Format)
ML103500258
Person / Time
Site: Nine Mile Point Constellation icon.png
Issue date: 10/27/2010
From: Todd Fish
Operations Branch I
To: Dean R
Constellation Energy Nuclear Group
Hansell S
Shared Package
ML101900588 List:
References
TAC U01797
Download: ML103500258 (224)
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Text

Nine Mile Point Unit 1 Level RO SRO Tier # 2 Group # 1 KIA # 262001 K1.02 Importance Rating Knowledge of the physical connections and/or cause- effect relationships between AC.

ELECTRICAL DISTRIBUTION and the following: D.C. electrical distribution Proposed Question: RO Question # 1 The plant is operating at 100% power when a breaker failure on Powerboard 17B causes a loss of power to 125 VDC Static Battery Chargers 171A and 171 B, only.

Which one of the following describes a result of this power loss?

A DC power is lost to RPS UPS 172; all other loads will continue to operate for a minimum of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> B. All 125 VDC loads continue to operate; AC power can be re-established to battery chargers 171Aand 171Bfrom PB 16B C. All 125 VDC loads continue to operate; AC power must be re-established to battery chargers 171A and 171B within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> D. DC load shedding automatically trips breakers to non-essential loads; DC power is available to essential loads for a minimum of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> Proposed Answer: C Explanation (Optional):

A Incorrect - The battery will continue to supply power to all DC loads B. Incorrect - The 171A and 171 B battery chargers cannot be supplied from PB 16B C. Correct - When both SR Static chargers for a battery are out of service or lose AC power, its associated battery will supply the 125 VDC loads. The time required to restore a battery charger to service is 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

D. Incorrect - There is no automatic DC load shedding on the trip of a battery charger Technical Reference(s): N1-0P-47A, Sect. B, Sect. D.9, (Attach if not previously provided)

and Att. 8 Proposed References to be provided to applicants during examination: No Learning Objective: N1-263000-RBO-1 0 (As available)

Question Source: Bank # ID: AUD 2008 RO 10 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: Not used on 2008 or 2009 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 b(07) Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features 55.43 Comments:

Nine Mile Point Unit 1 Level RO SRO Tier # 2 Group #

KIA # 212000 K1.02 Importance Rating Knowledge of the physical connections and/or cause- effect relationships between REACTOR PROTECTION SYSTEM and the following: Nuclear boiler instrumentation Proposed Question: RO Question # 2 The plant is operating at 100% power with the following conditions:

  • RPV pressure transmitter,36-07C, fails upscale

A. RPS Channel 11 half scram, only B. RPS Channel 12 half scram, only C. RPS Channel 11 half-scram and channel 11 ATS TROUBLE light D. RPS Channel 12 half scram and channel 12 ATS TROUBLE light Proposed Answer: C Explanation (Optional):

A. Incorrect - Because the PT signal has failed upscale, the ATS Cabinet master trip unit generates a gross failure output signal, which results in an A TS TROUBLE light (for RPS Channel 11) on the control room F Panel.

B. Incorrect - Because A and CPT's input to RPS Channel11. Because the PT signal has failed upscale, the ATS Cabinet master trip unit generates a gross failure output signal, which results in an ATS TROUBLE light (for RPS Channel 11) on the control room F Panel.

C. Correct - A and CPT's input to RPS Channel11. Because the PT signal has failed upscale, the A TS Cabinet master trip unit generates a gross failure output signal, which results in an ATS TROUBLE light (for RPS Channel 11) on the control room F Panel.

D. Incorrect - Because A and CPT's input to RPS Channel11.

Technical Reference{s): C-19859-C sht 2, N1-0P-40, Sect. (Attach if not previously provided)

B, C-18015-C, N1101212000C01 Proposed References to be provided to applicants during examination: None Learning Objective: N1-212000-RBO-11 (As available)

Question Source: Bank #

Modified Bank # ID: N1-212000-RBO- (Note changes or attach parent) 11-Q-03 New Question History: Last NRC Exam: Not used on 2008 or 2009 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 b(6) Design, components, and functions of reactivity control mechanisms and instrumentation.

55.43 Comments:

Nine Mile Point Unit 1 Level RO SRO Tier# 2 Group # 1 KJA# 300000 K2.02 Importance Rating 3.0 Knowleqge of electrical power supplies to the following: Emergency air compressor Proposed Question: RO Question # 3 The plant is operating at 100% power with the following Instrument Air Compressor (lAC) lineup:

  • lAC 13 is in service
  • lAC 11 is red flagged as the backup compressor
  • lAC 12 is green flagged as the standby compressor Then, Powerboard 16A is de-energized due to an electrical fault.

Which one of the following lists the safety-related IAC(s) available to supply header pressure, if any?

A. No safety-related lAC is available B. lAC 11 is available, only C. lAC 12 is available, only D. Both lAC 11 and 12 are available Proposed Answer: C Explanation (Optional):

A. Incorrect - lAC 12 is powered from PB 17A and remains available.

B. Incorrect - lAC 11 is powered from PB 16A. The loss of PB 16A makes lAC 11 unavailable.

C. Correct - lAC 12 is powered from PB 17A and remains available.

D. Incorrect -lAC 11 is powered from PB 16A. The loss of PB 16A makes lAC 11 unavailable. Plausible if candidate mistakes lAC 11 power supply as PB 16B vs. PB 16A.

Technical Reference(s): N1-0P-20, Sect B (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None Learning Objective: N1-27BOOO-RBO-4 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New x Question History: Last NRC Exam: Not used on 200B or 2009 Question Cognitive Level: Memory or Fundamental Knowledge x Comprehension or Analysis 10 CFR Part 55 Content: 55.41 b(07) Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features 55.43 Comments:

Nine Mile Point Unit 1 Level RO SRO Tier # 2 Group # 1

~----.---- .. ---~

KIA # 206000 K2.01 Importance Rating 3.2 Knowledge of electrical power supplies to the following: System valves: BWR-2,3,4 Proposed Question: RO Question # 4 The plant was operating at 100% power when a failure to scram resulted in the following:

  • The RO closed the following valves:
  • Feedwater FCV 12 Then, an electrical fault caused a loss of Powerboard 171 B.

Which valve CANNOT be manually re-opened from the Control Room due to this power loss?

A. Feedwater Isolation Valve 11 B. Feedwater Isolation Valve 12 C. Feedwater FCV 11 D. Feedwater FCV 12 Proposed Answer: B Explanation (Optional):

A. Incorrect - Feedwater Isolation Valve 11 is powered from PB 161 B B. Correct - Feedwater Isolation Valve 12 is a motor operated valve powered from PB 171B. With the loss of PB 171B, Feedwater Isolation Valve 12 fails as is.

C. Incorrect - Feedwater FCV 11 is an air-operated valve, with control power coming from RPS bus 11. The loss of PB 171 B does not affect RPS Bus 11.

D. Incorrect - Feedwater FCV 12 is an air-operated valve, with control power coming from

RPS bus 12. The loss of PB 171 B does not affect RPS Bus 12.

Technical Reference(s): N1-0P-16, Sect B. pg 14 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None Learning Objective: N1-259001-RBO-4 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam: No Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 b(07) Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features 55.43 Comments:

Nine Mile Point Unit 1 Level RO SRO Tier # 2 Group # 1 KIA # 239002 K3.03 Importance Rating 4.3 Knowledge of the effect that a loss or malfunction of the RELIEF/SAFETY VALVES will have on following: Ability to rapidly depressurize the reactor Proposed Question: RO Question # 5 The plant was operating at 100% power when the following events occurred:

  • An un-isolable steam leak in the Reactor Building has led to the need for an RPV Blowdown
  • Fuse failures caused four (4) ERVs to fail closed Which one of the following describes the effect of the ERV loss on the ability to rapidly depressurize the Reactor, in accordance with EOP-8, RPV Blowdown?

The Minimum Number of ERVs Required for Emergency Depressurization (1) and (2)

(1) (2)

A. is available the use of alternate Blowdown Systems to rapidly depressurize the Reactor is allowed, but NOT required B. is available the use of alternate Blowdown Systems to rapidly depressurize the Reactor is NEITHER required NOR allowed

c. is NOT available Turbine Bypass Valves may be used to rapidly depressurize the Reactor even if MSIV isolations must be defeated D. is NOT available Turbine Bypass Valves may be used to rapidly depressurize the Reactor but MSIV isolations must NOT be defeated Proposed Answer: C Explanation (Optional):

A. Incorrect - The Minimum Number of ERVs Required for Emergency Depressurization is

3. With four ERVs unavailable, only two remain available.

S. Incorrect - The Minimum Number of ERVs Required for Emergency Depressurization is

3. With four ERVs unavailable, only two remain available. EOP-8 step 15 allows use of alternate Slowdown systems since less than 3 ERVs can be opened.

C. Correct - The Minimum Number of ERVs Required for Emergency Depressurization is

3. With four ERVs unavailable, only two remain available. EOP-8 step 15 allows use of alternate Slowdown systems since less than 3 ERVs can be opened. This step and detail 0 also both give instructions allowing all isolation signals to be defeated.

D. Incorrect - EOP-8 step 15 allows use of alternate Slowdown systems since less than 3 ERVs can be opened. This step and detail 0 also both give instructions allowing all isolation signals to be defeated.

Technical Reference(s): EOP-8, EOP bases (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None Learning Objective: ID: N1-218000-RSO-04 (As available)

Question Source: Sank #

Modified Sank # (Note changes or attach parent)

New X Question History: Last NRC Exam: Not used on 2008 or 2009 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 b(07) Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features 55.43 Comments:

Facility: Nine Mile Point Unit 1 Vendor: GE Exam Date: 2010 Exam Type: R Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 KIA # 262002 K3.08 Importance Rating 2.7 Knowledge of the effect that a loss or malfunction of the UNINTERRUPTABLE POWER SUPPLY (AC.lD.C.) will have on following: Computer operation: Plant-Specific Proposed Question: RO Question # 6 The plant is operating at 100% power when MG Set 167 trips.

Which one of the following is the effect on plant operations?

A Loss of power to the Bypass Opening Jack B. Feedwater level control swaps to the backup power supply C. Loss of power to Emergency Diesel Generator 102 speed control D. Instantaneous core thermal power calculations will be unavailable Proposed Answer: D Explanation (Optional):

A Incorrect - Loss of MG 167 causes a loss of power to the EPR; however the BOJM is powered from a different power supply (BB 11)

B. Incorrect - The loss of the process computer results in the feedwater correction factor values being reset to 0.00. However feedwater control is NOT affected. MG 167 is the backup power supply to FWLC.

C. Incorrect - EDG 102 speed control power is from Battery Board 11, not MG 167.

D. Correct - Loss of MG Set 167 leads to a loss of the PPC. This makes the CTP

calculation unavailable.

N1-0P-48 sect B Technical Reference(s): N1-0P-42 sect D and H.2 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None Learning Objective: N1-262002-RBO-11 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam: Not used on 2008 or 2009 exams Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis be?) Design, components, and functions of control and safety systems, including instrumentation, signals, 10 CFR Part 55 Content: 55.41 interlocks, failure modes, and automatic and manual features.

55.43 Comments:

Facility: Nine Mile Point 1 Vendor: GE Exam Date: 2010 Exam Type: RO Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 KJA# 259002 K4.09 Importance Rating 3.1 Knowledge of REACTOR WATER LEVEL CONTROL SYSTEM design feature(s) and/or interlocks which provide for the following: Single element control (reactor water level provides the only input)

Proposed Question: RO Question # 7 The plant is operating at 10% power.

Which one of the following parameters is used by the Feedwater Level Control System (FWLC) to control RPV water level at this power level, in accordance with OP-16, Feedwater System Booster Pump to Reactor?

A. Reactor water level only B. Reactor water level and Steam flow, only C. Reactor water level and Feedwater flow, only D. Reactor water level, Steam flow and Feedwater flow Proposed Answer: A Explanation (Optional):

A. Correct - N1-0P-16 E2.0 and 7.12 WHEN power is greater than 25%, shift to 3 ELEMENT FEEDWATER MODE below 25% and on initial level control of the Vessel.

FWLC is in Single Element controlled by reactor water level alone.

B. Incorrect - Steam flow is only used when in three element control, which is not used until 25% power.

C. Incorrect - Feedwater flow is only used when in three element control, which is not used until 25% power.

D. Incorrect - This is three element control, which is not used until 25% power.

Technical Reference(s): N1-0P-16 Section B, E.2.0 and (Attach if not previously provided) 7.12 Proposed References to be provided to applicants during examination: None Learning Objective: N 1-259002-RBO-5 (As available)

Question Source: Bank # 1& C Bank ID: 17254 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: Not used on the 2008 or 2009 exams Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 5 55.43 Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.

Comments:

Facility: Nine Mile Point 1 Vendor: GE Exam Date: 2010 Exam Type: RO Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 KJA# 400000 K4.01 Importance Rating 3.4 Knowledge of CCWS design feature(s) and or interlocks which provide for the following:

Automatic start of standby pump Proposed Question: RO Question # 8 The plant is operating at 100% power with the following conditions:

  • TBCLC pump 11 is operating
  • TBCLC heat exchanger 12 is in service
  • TBCLC pressure is 90 psig
  • TBCLC heat exchanger outlet temperature is 94°F Then, an operator valving in TBCLC heat exchanger 11 causes TBCLC pressure to temporarily drop to 75 psig.

Which one of the following describes a consequence of the operator's action?

A. TBCLC pump 12 automatically starts, potentially tripping Instrument Air Compressor 12.

B. TBCLC pump 12 automatically starts, potentially causing heat exchanger tube damage.

C. TBCLC flow control valve 71-124 opens, causing TBCLC temperatures to rise above 94°F.

D. TBCLC temperature control valve 71-88 repositions, causing TBCLC temperatures to rise above 94°F.

Proposed Answer: B Explanation (Optional):

A. Incorrect - TBCLC Pump 12 will auto start when TBCLC pressure drops below 80 psig however Instrument Air Compressor 13 is the compressor that may be affected by the pump start.

B. Correct - TBCLC Pump 12 will auto start when TBCLC pressure drops below 80 psig.

Operating two pumps may cause Heat Exchanger tube damage.

C. Incorrect - Both operator actions (adding cooling with 11 heat exchanger and adding flow by auto starting the TBCLC pump) will cause the lube oil TCV,71-119, to throttle close. This will cause the bypass flow control valve,71-124, to throttle open. However, these valve movements will tend to lower heat addition to TBCLC.

D. Incorrect - Both operator actions (adding cooling with 11 heat exchanger and adding flow by auto starting the TBCLC pump) will tend to lower TBCLC temperatures. TBCLC temperature control valve repositions to maintain TBCLC temperature at setpoint.

Technical Reference(s): SN1-0. P-24'1 PO& L 1 and 3, Notes in (Attach if not previously provided) ect Ion F.. an d F..

20 Proposed References to be provided to applicants during examination: None Learning Objective: N1-274000-RBO-3 & 5 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam: Not used on 2008 or 2009 exams Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 4 55.43 Secondary coolant and auxiliary systems that affect the facility.

Comments:

Facility: Nine Mile Point Unit 1 Vendor: GE Exam Date: 2010 Exam Type: R Examination Outline Cross-reference: Level RO SRO Tier# 2 Group # 1 KJA# 215005 K5.01 Importance Rating 2.8 Knowledge of the operational implications of the following concepts as they apply to AVERAGE POWER RANGE MONITOR/LOCAL POWER RANGE MONITOR SYSTEM: LPRM detector operation Proposed Question: RO Question # 9 The plant is operating at 100% power with the following:

  • An Operator reports that LPRM 28-33A (input to APRM 11) indicates upscale Which one of the following states the impact of this failure on APRM 11 indication and indicated Core Thermal Power?

APRM 11 Core Thermal Power A. Indicates higher Indicates higher

8. Indicates higher No impact C. No impact Indicates higher D. No impact No impact Proposed Answer: 8 Explanation (Optional):

A. Incorrect - LPRM indicated power is not used in a CTP heat balance.

B. Correct - The APRM monitors the outputs of eight LPRMs and computes a signal which is proportional to the arithmetic average of the eight. The output of the APRM is recorded on a recorder on the E Panel. With one of the eight failed upscale, the reading of 125 averaged with the other 7 readings of 100 would cause the indicator to indicate higher than actual. There is no design feature which automatically detects LPRM failure to correct this. LPRM indicated power is not used in a heat balance.

C. Incorrect - The APRM would read higher because of the addition of an upscale (125) with the normal LPRM inputs. LPRM indicated power is not used in a heat balance.

D. Incorrect - The APRM would read higher because of the addition of an upscale (125) with the normal LPRM inputs.

Technical Reference(s): N1-REP-8, N1-0P-38C, Sect B (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None Learning Objective: N1-215000-RBO-5 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam: Not used on the 2008 or 2009 exams Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 2 55.43 General DeSign features of the core, including core structure, fuel elements, control rods, core instrumentation, and coolant flow.

Comments:

Facility: Nine Mile Point Unit 1 Vendor: GE Exam Date: 2010 Exam Type: RO Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 KIA # 207000 K5.03 Importance Rating 2.7 Knowledge of the operational implications of the following concepts as they apply to ISOLATION (EMERGENCY) CONDENSER: Heat transfer: BWR-2,3 Proposed Question: RO Question # 10 A steam leak in the Turbine Building resulted in the following:

  • Reactor power is 6% and steady
  • Reactor pressure is 950 psig and steady
  • Emergency Condenser (EC) loops 11 and 12 are in service
  • Then, a leak in the EC 111 shell results in EC 111 shell side water level continuously lowering Which one of the following describes the EC 11 heat transfer rate as shell side water level continues to drop and the appropriate Operator action to maintain Reactor pressure?

EC 11 Heat Transfer Rate Operator Action to Maintain Reactor Pressure A. Rises Secure EC 11 B. Rises Secure EC 12 C. Lowers Manually cycle ERVs D. Lowers Open Turbine Bypass Valves

Proposed Answer: C Explanation (Optional):

A. Incorrect - With lowering shell side water level, heat transfer from the Reactor coolant in the EC tubes to the shell side water lowers.

B. Incorrect - With lowering shell side water level, heat transfer from the Reactor coolant in the EC tubes to the shell side water lowers.

C. Correct - With lowering shell side water level, heat transfer from the Reactor coolant in the EC tubes to the shell side water lowers. With lower heat transfer, the EC will reject less heat from the RCS, resulting in Reactor pressure rising. Additional pressure control systems will need to be placed in service. With indications of a steam leak in the Turbine Building, MSIVs will not be able to be reopened to use the TBVs for pressure control. ERVs are available for pressure control.

D. Incorrect - With indications of a steam leak in the Turbine Building, MSIVs will not be able to be reopened to use the TBVs for pressure control.

. EOP-3 N1101207000C01 SDBD . . .

Technical Reference(s): 204 ' '(Attach If not previously provided)

Proposed References to be provided to applicants during examination: None Learning Objective: N1-207000-RBO-11 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam: Not used on 2008 or 2009 exams Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 14 55.43 Principles of heat transfer, thermodynamics and fluid mechanics.

Comments:

Facility: Nine Mile Point Unit 1 Vendor: GE Exam Date: 2010 Exam Type: R Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 KIA # 205000 K6.02 Importance Rating 2.7 Knowledge of the effect that a loss or malfunction of the following will have on the SHUTDOWN COOLING SYSTEM (RHR SHUTDOWN COOLING MODE): D.C. electrical power Proposed Question: RO Question # 11 The plant is shutdown with the following:

  • All Recirculation pumps are secured
  • A short circuit results in a loss of 125 VDC Battery Board 12 Which one of the following is the effect on SDC?

A. The system continues to operate. The only affect is that 38-02, SDC Inlet IV (Outside),

loses power.

B. A SDC Isolation signal occurs. SDC pump 12 does NOT trip and 38-02, SDC Inlet IV (Outside), does NOT close.

C. A SDC Isolation signal occurs. SDC pump 12 trips, however 38-02, SDC Inlet IV (Outside), does NOT close.

D. The system continues to operate. 38-02, SDC Inlet IV (Outside), loses power, and the ability to operate SDC Pump 12 from the Control Room is lost.

Proposed Answer: D Explanation (Optional):

A. Incorrect - SDC Pump 12 loses control power and cannot be remotely operated.

B. Incorrect - RPS supplies the isolation logic so with just a loss of DC no isolation signal occurs. A loss of 125 VDC does NOT cause a condition that will cause an SDC Isolation. The DC distribution system supplies 125 VDC to operate the DC motor operated isolation valve (38-02). On a loss of 125 VDC Bus 12 38-02 will fail as-is.

C. Incorrect - RPS supplies the isolation logic so with just a loss of DC no isolation signal occurs. A loss of 125 VDC does NOT cause a condition that will cause an SDC Isolation. The DC distribution system supplies 125 VDC to operate the DC motor operated isolation valve (38-02). On a loss of 125 VDC Bus 12 38-02 will fail as-is.

D. Correct - RPS supplies the isolation logic so with just a loss of DC no isolation signal occurs. A loss of 125 VDC does NOT cause a condition that will cause an SDC Isolation. The DC distribution system supplies 125 VDC to operate the DC motor operated isolation valve (38-02) and control power for the pumps. On a loss of the respective power supply, 38-02 will fail as-is. On a loss of DC control power, the ability to operate the pump from the control room will be lost.

N 110 1205000C01 ,Student Text pg 104 Technical Reference(s): C19845C Sht 1 (Attach if not previously provided)

C19439C Sht 10 SOP-47A.1 Proposed References to be provided to applicants during examination: None Learning Objective: N 1-205000-RBO-4 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam: Not used on 2008 or 2009 exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 4 55.43

Secondary coolant and auxiliary systems that affect the facility.

Comments:

Facility: Nine Mile Point Unit 1 Vendor: GE Exam Date: 2010 Exam Type: R Examination Outline Cross-reference: Level RO SRO Tier# 2 Group # 1 KIA # 215004 K6.01 Importance Rating 3.2 Knowledge of the effect that a loss or malfunction of the following will have on the SOURCE RANGE MONITOR (SRM) SYSTEM: RPS Proposed Question: RO Question # 12 Refueling is currently in progress, with the following:

  • REFUEL INST TRIP BYPASS CH 11 AND 12 keylock switches are in "NON COINCIDENT"
  • MG set 131 output breaker trips open
  • Reactor trip bl.ls 131 de-energizes Which one of the following describes the effect of this loss on both SRM 11 indication AND the RPS scram channels?

SRM 11 indication will ...

A. Fail downscale. A full scram will occur.

B. Remain accurate. A full scram will occur.

C. Fail downscale. A half scram will occur on RPS channel 11 only.

D. Remain accurate. A half scram will occur on RPS channel 11 only.

Proposed Answer: D Explanation (Optional):

A. Incorrect - The loss of power to the RPS solenoids (Reactor Trip Bus 131) will cause a half scram, however the SRM circuitry is not affected and because the SRM circuitry is

not affected the position of the NON-COINCIDENT switches does not result in a full scram.

B. Incorrect - The loss of power to the RPS solenoids (Reactor Trip Bus 131) will cause a half scram, however the SRM circuitry is not affected (does NOT de-energize) and because the SRM circuitry is not affected the position of the NON-COINCIDENT switches does not result in a full scram.

C. Incorrect - The loss of power to the RPS solenoids (Reactor Trip Bus 131) will cause a half scram, however the SRM circuitry is not affected (does NOT de-energize) and because the SRM circuitry is not affected, the SRM indication is not affected.

D. When both NON-COINCIDENT switches are placed in NON-COINCIDENT anyone SRM trip will cause a reactor scram, however Reactor Trip Bus 131 does NOT supply the power to the SRMs or the scram relays, they are powered from RPS Buses 11 and 12, which are powered by UPS 162A and 162B, so a loss of RPS Bus 131 power will not cause an SRM trip or Rod Block. The loss of power to the RPS solenoids (Reactor Trip Bus 131) will cause a half scram.

N1-0P-40, Att 4 N1-0P-48, Sect. B Technical Reference(s): (Attach if not previously provided)

C-19859C, Shts 3,4,5,6 C-19957C Proposed References to be provided to applicants during examination: None Learning Objective: N 1-212000-RBO-4 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam: Not used on 2008 or 2009 exams Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 6 55.43 Design, components, and function of reactivity control mechanisms and instrumentation.

Comments:

Facility: Nine Mile Point Unit 1 Vendor: GE Exam Date: 2010 Exam Type: R Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 KIA # 263000 A 1.01 Importance Rating 2.5 Ability to predict and/or monitor changes in parameters associated with operating the D.C.

ELECTRICAL DISTRIBUTION controls including: Battery charging/discharging rate Proposed Question: RO Question # 13 The plant is operating at 100% power with the following:

  • Static Battery Chargers (SBC) 161A and 171A are in service
  • Powerboard 16B is inadvertently de-energized then re-energized Which one of the following identifies the status of Batteries 11 and 12 two (2) minutes after Powerboard 16B is re-energized?

Battery 11 Battery 12 A. Discharging Discharging B. Discharging Charging C. Charging Discharging D. Charging Charging Proposed Answer: D Explanation (Optional):

A. Incorrect - Both batteries will be on their respective SBC and charging.

B. Incorrect - Both batteries will be on their respective SBC and charging.

C. Incorrect - Both batteries will be on their respective SBC and charging.

D. Correct - Approximately 100 seconds after AC power is restored to SBC161A (from powerboard 16B) the SBC will align itself to the Battery 11 and restore the normal float charge. Battery 12 is unaffected by the transient since its SBC 171A receives power from PB 17B, which is unaffected by the loss of PB 16B.

Technical Reference(s): N1-0P-47A, Sect B (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None Learning Objective: N 1-263000-RBO-3 (As available)

Question Source: Bank #

Modified Bank # 10: NRC 2006 RO 32 (Note changes or attach parent)

New Similar to #13 and #43 on 2008 NRC Exam, but sufficiently different Question History: Last NRC Exam: because it covers different batteries under a different situation.

No similar questions on 2009 exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Comments:

Facility: Nine Mile Point Unit 1 Vendor: GE Exam Date: 2010 Exam Type: R Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 KIA # 261000 A1.06 Importance Rating 2.7 Ability to predict and/or monitor changes in parameters associated with operating the STANDBY GAS TREATMENT SYSTEM controls including: Drywell and suppression chamber differential pressure: Mark-I Proposed Question: RO Question # 14 The plant is operating at 100% power with the following:

  • Drywell pressure is 1.9 pSig
  • Torus pressure is 1.2 psig
  • Drywell venting is in progress using RBEVS train 11, in accordance with N 1-0P-9 section H.1.0, Venting Primary Containment Through RBEVS During Normal Ops
  • Then, RBEVS fan 11 motor breaker trips Which one of the following describes the effect of the fan trip on RBEVS train 11 AND the resulting Drywell pressure once steady-state conditions are reached?

RBEVS train 11 ...

A. Remains unisolated. The Containment vent valves remain open, allowing Drywell pressure to lower below 1.2 psig.

B. Remains unisolated. The Containment vent valves remain open, allowing Drywell pressure to lower and stabilize at 1.2 psig.

C. Isolates on low flow. The Containment vent valves remain open, allowing Drywell pressure to lower and stabilize at 1.2 psig.

D. Isolates on low flow. The Containment vent valves automatically close, isolating the Drywell and stabilizing Drywell pressure at 1.9 psig.

Proposed Answer: A Explanation (Optional):

A. Correct - The only interlock is between the inlet and cooling dampers. The fan trip will not close any dampers. When venting the Drywell only the Drywell vent valves (201-32 and 201-31) are open. When the RBEVS fan trips Drywell pressure will vent through the idle fan lowering Drywell pressure below the Torus pressure of 1.2 pSig. If Drywell pressure lowers 0.5 psid below Torus pressure the vacuum breakers will open.

B. Incorrect - When venting the Drywell only the Drywell vent valves (201-32 and 201-31) are open. When the RBEVS fan trips Drywell pressure will vent through the idle fan lowering Drywell pressure below the Torus pressure of 1.2 psig.

C. Incorrect - There are no fan and damper trips the only interlock is between the inlet and cooling dampers. Fan trip will not close any dampers. When venting the Drywell only the Drywell vent valves (201-32 and 201-31) are open. When the RBEVS fan trips Drywell pressure will vent through the idle fan lowering Drywell pressure below the Torus pressure of 1.2 psig.

D. Incorrect - There are no fan and damper trips. The only interlock is between the inlet and cooling dampers. Fan trip will not close any dampers. The purge valves will not automatically close because no containment isolation or hi hi stack radiation signal is present.

N1-0P-9, H.1.0 Technical Reference(s): C-1B014-C (Attach if not previously provided)

C-1B013-C Proposed References to be provided to applicants during examination: None Learning Objective: N1-261000-RBO-3 & 5 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam: Not used in 200B or 2009 exams Question Cognitive Level: Memory or Fundamental Knowledge

Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 7 55.43 Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Comments:

Facility: Nine Mile Point Unit 1 Vendor: GE Exam Date: 2010 Exam Type: RO Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 KIA # 209001 A2.06 Importance Rating 3.2 Ability to (a) predict the impacts of the following on the LOW PRESSURE CORE SPRAY SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Inadequate system flow Proposed Question: RO Question # 15 The plant is operating at 100% power when a small break LOCA results in the following:

  • RPV pressure is 950 psig and stable
  • RPV water level is 60" and stable
  • Drywell pressure is 6 psig and stable
  • Core Spray Pump 111 has been running for three hours
  • Core Spray Topping Pump 111 has been running for three hours

A. All Core Spray Jumpers must be installed within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> to prevent RPV overfill and allow makeup to the Torus.

B. All Core Spray Pumps and Core Spray Topping Pumps must be placed in Pull to Lock within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> to prevent overheating the pumps.

C. Core Spray Pumps and Core Spray Topping Pumps must be swapped within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> to prevent overheating the Subsystem 111 pumps.

D. Jumpers for Core Spray Loop 11 only must be installed within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> to allow Core Spray flow through 40-06, Core Spray Test Valve 11.

Proposed Answer: C Explanation (Optional):

A. Incorrect - Based on a current pressure of 950 psig and the injection valves opening at 365 psig. It will take over an hour before injection is possible with a normal cooldown rate. Therefore the need is to protect the operating Core Spray and Core Spray Topping pumps before they operate for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> on minimum flow.

B. Incorrect - As long as Drywell pressure is above 3.5 psig, at least one train of Core Spray must be running to provide the SDC water seal. Based upon actions in N1-EOP 2 to cooldown, it will take over 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> to reach 365 psig.

C. Correct - System design allows each subsystem to operate up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> before pump overheating becomes a concern. N1-EOP-1, Attachment 4 directs swapping to an idle loop.

D. Incorrect There are no directions to establish flow through the test valve. Core Spray now may continue for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> on the minimum flow valve.

Technical Reference(s): N1-EOP-1 ATT. 4 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None Learning Objective: N1-209001-RBO-10 & 12 (As available)

Question Source: Bank #

Modified Bank # N1-209001C01-RBO- (Note changes or attach parent) 10-Q-04 New Question History: Last NRC Exam: Not used on the 2008 or 2009 NRC Exams Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 8 55.43 Components, capacity, and functions of emergency systems.

Facility: Nine Mile Point Unit 1 Vendor: GE Exam Date: 2010 Exam Type: R Examination Outline Cross-reference: Level RO SRO Tier# 2 Group # 1 KIA # 215003 A2.05 Importance Rating 3.3 Ability to (a) predict the impacts of the following on the INTERMEDIATE RANGE MONITOR (IRM) SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Faulty or erratic operation of detectors/system Proposed Question: RO Question # 16 A plant startup is in progress with the following:

  • The Mode Switch is in STARTUP with control rod withdrawal in progress
  • IRMs 11,12,15,16, and 18 read approximately 75 out of 125 on Range 2
  • IRMs 13, 14, and 17 read approximately 15 out of 125 on Range 3 Then, a malfunction in the IRM drive circuitry causes the IRM 13 detector to withdraw to the full-out position.

Which one of the following states the effect on the plant AND the required operator actions to continue withdrawing control rods?

This will result in panel annunciators ...

A. ONLY; withdrawing control rods may continue without any other control panel manipulations.

B. and a rod block from IRM downscale ONLY; bypassing IRM 13 is required to continue withdrawing control rods.

C. and a rod block from IRM downscale AND IRM detector position; bypassing IRM 13 is required to continue withdrawing control rods.

D. and a rod block. The IRM detector out of position cannot be bypassed and the startup must be halted until the detector is fully inserted or jumpers are installed.

Proposed Answer: C Explanation (Optional):

A. Incorrect - because it does not list rod blocks which need to be bypassed.

B. Incorrect - because it does not list all rod blocks. In addition to IRM Downscale a rod block will be caused by the detector not being fully inserted with the Mode Switch in STARTUP.

C. Correct - The detector moving out of the core will cause IRM 13 to go downscale as the detector moves out of the core. Additionally a rod block will be caused by the detector not being fully inserted with the Mode Switch in STARTUP. IRM 13 must be bypassed to permit clearing the rod block and continuing the startup.

D. Incorrect because the detector not inserted can be bypassed by bypassing IRM 13.

This will permit clearing the rod block and continuing the startup.

Technical Reference(s): N1-0P-38B (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None Learning Objective: N1-215000-RBO-11 (As available)

Question Source: Bank # 2009 NRC #37 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: 2009 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 7 55.43 Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Comments:

Facility: Nine Mile Point Unit 1 Vendor: GE Exam Date: 2010 Exam Type: R Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 KIA # 218000 A3.04 Importance Rating 3.7 Ability to monitor automatic operations of the AUTOMATIC DEPRESSURIZATION SYSTEM including: Primary containment pressure Proposed Question: RO Question # 17 A LOCA has resulted in the following:

  • RPV water level is -20 inches and slowly lowering
  • RPV pressure is 850 psig and slowly lowering
  • Drywell pressure rose to a high value of 5.5 psig, then Containment Spray initiated and quickly lowered Drywell pressure
  • Drywell pressure is now 2.5 psig and slowly lowering
  • The ADS timer has been active for 45 seconds Which one of the following describes the status of the ADS channel 11 white timer light and ERV 111 blue continuity light one (1) minute later?

ADS Channel 11 ERV 111 Blue White Timer Light Continuity Light A. Illuminated Illuminated B. Illuminated Extinguished C. Extinguished Illuminated D. Exting uished Extinguished Proposed Answer: A

Explanation (Optional):

A. Correct - The drywell pressure signal lowered below its trip setpoint (3.5 psig). The high pressure signal however "seals in". Therefore even though the condition has cleared, without an operator resetting the initiation, the timers will continue and the timer light will be lit. After one more minute, the initiation signal will have been present for 105 seconds, which is less than the 111 second time delay for ERV actuation. Therefore ERV 111 will still be closed, with the continuity light lit.

B. Incorrect - ERV 111 has not received an open signal from ADS yet, therefore the continuity light is still lit.

C. Incorrect - The ADS timer is still timing, therefore the ADS timer light is lit.

D. Incorrect - ERV 111 has not received an open signal from ADS yet, therefore the continuity light is still lit. Also, the ADS timer is still timing, therefore the ADS timer light is lit.

. N1-0P-2 Sect. B Technical Reference(s): DWG C-19859-C, Sheets 2, 18 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None Learning Objective: N1-218000-RBO-5 & 11 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam: Not used on 2008 or 2009 exams Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Comments:

Facility: Nine Mile Point Unit 1 Vendor: GE Exam Date: 2010 Exam Type: R Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 KIA # 223002 A3.02 Importance Rating 3.5 Ability to monitor automatic operations of the PRIMARY CONTAINMENT ISOLATION SYSTEM/NUCLEAR STEAM SUPPLY SHUT-OFF including: Valve closures Proposed Question: RO Question # 18 A small steam line break in the drywell has resulted in the following:

  • RPV water level dropped to a low of 22 inches before recovering to the normal band
  • Drywell pressure is 4.2 psig and slowly rising Which one of the following describes valves that have received a close signal in response to an automatic isolation?

A. Emergency Condenser Vents (05-02, 05-03)

B. Cleanup Isolation Valves (33-01,33-02,33-04)

c. Reactor Water Sample Return Valves (63-04, 63-05)

D. Drywell Sample System Isolation Valves (201.7-01, 201.7-02)

Proposed Answer: 0 Explanation (Optional):

A. Incorrect - A vessel isolation signal has NOT occurred. These valves are vessel isolations not containment isolations.

B. Incorrect - A vessel isolation signal has NOT occurred. These valves are vessel isolations not containment isolations.

C. Incorrect - A vessel isolation signal has NOT occurred. These valves are vessel isolations not containment isolations.

D. Correct - A Vessel Isolation is caused by:

  • LoLoLo Vacuum s 7" Hg
  • High Area Temperature ~ 200°F
  • High Steam Flow ~ 105 psid A Containment Isolation is caused by:
  • RPV LoLo Level 5"
  • High Drywell Pressure 3.5 psig Only a Containment isolation has occurred, caused by Drywell pressure. A Vessel isolation has not occurred because the only parameter given, RPV water level, is above the isolation setpoint.

Technical Reference(s): N1-S0P-40.2 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None Learning Objective: N 1-223002-RBO-5 (As available) 10: N1-223002 Question Source: Bank #

RBO-05-Q-09 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: Not used on the 2008 or 2009 exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Comments:

Facility: Nine Mile Point Unit 1 Vendor: GE Exam Date: 2010 Exam Type: R Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 KIA # 211000 A4.05 Importance Rating 4.1 Ability to manually operate and/or monitor in the control room: Flow indication: Plant-Specific Proposed Question: RO Question # 19 The plant was operating at 100% power when an inadvertent MSIV closure resulted in the following:

  • Reactor power is 15%
  • ERVs are cycling to control Reactor pressure
  • Liquid Poison pump 11 has been manually started from the Control Room Which one of the following describes indications that the explosive valves are open and there is flow into the RPV?

A. Discharge pressure of Liquid Poison pump is 1175 psig B. Discharge pressure of Liquid Poison pump is 1500 psig C. Current meter on Panel 1S-65 in Auxiliary Control Room indicates 0.15 amps D. Current meter on Pane11S-65 in Auxiliary Control Room indicates 2 amps Proposed Answer: A Explanation (Optional):

A. Correct - The ERVs will control RPV pressure at about 1100 psig. With the Liquid Poison Pump discharge pressure of 1175 pSig there is evidence of flow since the shut off head of the pump is 1675 psig and the relief valve is set at 1500 psig.

B. Incorrect - This indicates the explosive valve, or some other restriction exists and is blocking flow resulting in the pump operating at dead head pressure, this is also above the setpoint of the LP Pump Relief Valve.

C. Incorrect - This meter reading indicates there is continuity and valve has NOT fired.

D. Incorrect - A meter reading of 2 amps is normal firing current, but after successful firing, indication goes to 0 amps due to loss of continuity.

Technical Reference(s): N1-0P-12 Sect B (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None Learning Objective: N1-211000-RBO-6 & 10 (As available) 2009 NRC Exam Question Source: Bank #

  1. 34 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: 2009 NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 6 55.43 Comments:

Facility: Nine Mile Point Unit 1 Vendor: GE Exam Date: 2010 Exam Type: R Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 KIA # 264000 A4.01 Importance Rating 3.3 Ability to manually operate and/or monitor in the control room: Adjustment of exciter voltage Proposed Question: RO Question # 20 The plant is operating at 100% power with EDG 102 running loaded.

Which one of the following describes the effect of going to RAISE on the VOLTAGE ADJ RHEO GEN 102 control switch in the following conditions?

EDG 102 is operating with EDG 102 is operating PB 102 disconnected from the grid in parallel with the grid A. PB 102 voltage rises PB 102 voltage rises KVARs remain the same KVARs remain the same B. PB 102 voltage rises PB 102 voltage remains the same KVARs remain the same KVARs rise C. PB 102 voltage remains the same PB 102 voltage rises KVARs rise KVARs remain the same D. PB 102 voltage remains the same PB 102 voltage remains the same KVARs rise KVARs rise Proposed Answer: B Explanation (Optional):

A. Incorrect - When in parallel with the grid, raising EDG 102 excitation will cause the EDG to assume a higher MVAR loading, but will not change grid voltage.

B. Correct - With the EDG loaded but not connected to the grid, a change in excitation changes bus voltage, but MVAR loading is dependent solely on the bus loads. With the EDG loaded in parallel with the grid, a change in excitation causes MVAR loading to change, but is not enough to change grid voltage.

C. Incorrect - When disconnected from the grid, raising EDG 102 excitation directly affects bus voltage, but MVARs are dependent solely on bus loads. When in parallel with the grid, raising EDG 102 excitation will cause the EDG to assume a higher MVAR loading, but will not change grid voltage.

D. Incorrect - When disconnected from the grid, raising EDG 102 excitation directly affects bus voltage, but MVARs are dependent solely on bus loads.

Technical Reference(s): N1-0P-45, Sect G.1, E.4 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None Learning Objective: N1-264000-RBO 5 and 10 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam: Similar to 2009 #51, not used on 2008 Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Comments:

Facility: Nine Mile Point Unit 1 Vendor: GE Exam Date: 2010 Exam Type: RO Examination Outline Cross-reference: Level RO SRO Tier# 2 Group # 1 KIA # 218000 2.4.8 Importance Rating 3.8 Emergency Procedures I Plan: Knowledge of how abnormal operating procedures are used in conjunction with EOP's. (ADS)

Proposed Question: RO Question # 21 The plant is operating at 100% power when the following occur:

  • ERV 113 inadvertently opens
  • The crew enters SOP-1.4, Stuck Open ERV
  • Control Room actions to close the valve have NOT been successful
  • An operator has just been directed to ..IB Panels 11 and 12 on RB el 237
  • Torus water temperature is 86°F and rising Which one of the following sets of actions is correct, in accordance with CNG-PR-1.01-1009, Procedure Use and Adherence Requirements, and GAI-OPS-20, Transient Mitigation Guidelines?

A. Enter EOP-4, Primary Containment Control. Continue performing SOP-1.4. In the event of a conflict between the procedures, EOP-4 is the overriding document.

B. Enter EOP-4, Primary Containment Control. Continue performing SOP-1.4. In the event of a conflict between the procedures, SOP-1.4 is the overriding document.

C. Exit SOP-1.4 and enter EOP-4, Primary Containment Control. SOP-1.4 is re-entered at the step in-progress after exiting EOP-4.

D. Exit SOP-1.4 and enter EOP-4, Primary Containment Control. SOP-1.4 entry conditions are re-evaluated after exiting EOP-4.

Proposed Answer: A

Explanation (Optional):

A. Correct - Torus water temperature above 85°F requires entry into EOP-4. lAW CNG PR-1.01-1009, Sect. 5.4 1.1.b, examples of when Parallel Actions are used are the implementation of an abnormal operating procedure concurrently with an emergency operating procedure (EOP) when the abnormal operating procedure is supporting the recovery of safety functions within the EOP. The crew will continue in their efforts to close the ERV while executing the steps in the EOP{s). GAI-OPS-20 states that "Nothing shall supercede the proper implementation of the EOPs/SAPs". Therefore in the event a conflict between EOP-4 and SOP-1.4, EOP-4 is the overriding document.

B. Incorrect - EOP-4 is the higher tiered document, and therefore takes precedence in the event of a conflict with SOP-1.4.

C. Incorrect - CNG-PR-1.01-1009 allows parallel action in an EOP and SOP. SOP-1.4 does not have any specific exit criteria related to EOP-4. If SOP-1.4 was exited for some reason, re-entry would be required from the beginning of the procedure.

D. Incorrect - CNG-PR-1.01-1009 allows parallel action in an EOP and SOP. SOP-1.4 does not have any specific exit criteria related to EOP-4.

  • IR f T ec hnlca () CNG-PR-1.01-1009, Sect. 5.8, (Attach if not previously provided) e erence s: GAI-OPS-20 Proposed References to be provided to applicants during examination: None Learning Objective: NS101ADMTPSC38, #1 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam: Not used on the 2008 or 2009 exams Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

Facility: Nine Mile Point Unit 1 Vendor: GE Exam Date: 2010 Exam Type: RO Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 KIA # 209001 2.4.4 Importance Rating 4.5 Emergency Procedures I Plan: Ability to recognize abnormal indications for system operating parameters which are entry-level conditions for emergency and abnormal operating procedures (LPCS).

Proposed Question: RO Question # 22 The plant is operating at 100% power with N1-ST-Q1C, CS 112 PUMP AND VALVE OPERABILITY TEST in progress. While cycling 40-10, CORE SPRAY DISCHARGE IV 112 (INSIDE), the following occur:

  • Drywell floor drain input rises 0.55 gpm above the previous value when 40-10 is open
  • Drywell floor drain input remains 0.2 gpm above the previous value when 40-10 is closed
  • Drywell airborne radiation rises, then stabilizes at a higher level Which one of the following actions is required?

A. Enter EOP-4, Primary Containment Control B. Enter SOP-8.1, Increasing Drywell Leakage C. Enter SOP-40.2, Vessel/Containment Isolation D. Enter SOP-25.2, Fuel Failure or High Activity in RX Coolant or Off Gas Proposed Answer: B

Explanation (Optional):

A. Incorrect - There are no entry requirements to enter EOP-4. Plausible if leakage continued and had an effect on drywell pressure or torus water level.

B. Correct - The Core Spray Sparger DIP cell measures pressure in the sparger against that at the core plate. Normally the DIP is negative. If the sparger were to break inside the vessel or a leak occurred in a portion of the pipe or valve into the Drywell, the low pressure leg senses lowering pressure, causing the instrument to read some positive DIP causing the Diff Press Annunciator. The rising un-identified leakage rate exceeds the entry requirements for SOP-8.1. A probable cause for these conditions is a stem packing leak on Core Spray 40-10 or a pipe leak between 40-10 and the drywell penetration.

C. Incorrect - There is no indication that a vessel or containment isolation has occurred or is required. Indications show that closure of 40-10 has isolated leakage. Containment and vessel isolation plausible if leakage continued and had an effect on drywell pressure.

D. Incorrect - The rising Drywell Airborne radiation is not an entry requirement for SOP 25.2.

N 1-ARP-K3-4-1 Technical Reference(s): N1-S0P-8.1 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None Learning Objective: N1-209001-RBO-5 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam: Not used on 2008 or 2009 exams Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

Facility: Nine Mile Point Unit 1 Vendor: GE Exam Date: 2010 Exam Type: RO Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 KJA# 223002 A2.06 Importance Rating 3.0 Ability to (a) predict the impacts of the following on the PRIMARY CONTAINMENT ISOLATION SYSTEM/NUCLEAR STEAM SUPPLY SHUT-OFF; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Containment instrumentation failures Proposed Question: RO Question # 23 The plant is shutdown with the following:

  • SDC pumps 11 and 12 are operating
  • Drywell Cooling fans 11, 12 and 13 are in service
  • Containment Venting is in progress Then, a maintenance error causes the following Drywell pressure instruments to indicate upscale:
  • Drywell Pressure to RPS Channel 11-1
  • Drywell Pressure to RPS Channel 12-1 Which one of the following describes the impact of this failure on in-service systems, if any, and what actions are required?

Impact on In-Service Systems. if any Required Actions A. Drywell Cooling isolates Enter EOP-4, Primary Containment Control and place Containment Spray pumps in pull-to-Iock B. Containment Venting isolates Enter the High Drywell Pressure ARP and manually reset the Drywell high pressure alarm once the error is fixed

C. Shutdown Cooling isolates Enter SOP-6.1, Loss of SFP/RX Cavity Level/Decay Heat Removal and perform feed and bleed D. No impact on in-service systems Enter SOP-40.2, Vessel/Containment Isolation and verify isolation status Proposed Answer: B Explanation (Optional):

A. Incorrect - Drywell cooling does not isolate on a containment isolation.

B. Correct - With one DW pressure instrument upscale in both RPS channels 11 and 12, the logic is made up for a containment isolation. Of the listed systems, only Containment venting isolates on a containment isolation. The ARP gives guidance on resetting the high OW pressure condition.

C. Incorrect - SOC isolates on a vessel isolation, but not on a containment isolation.

D. Incorrect - Containment venting is affected by the containment isolation. Plausible if the candidate believes this combination of upscale DW pressure instruments is not sufficient to actuate a containment isolation.

N 1-S0P-40.2 Technical Reference(s): C-19859-C shts 2,4, 12 (Attach if not previously provided)

C-18014-C sht 1 Proposed References to be provided to applicants during examination: None Learning Objective: N 1-223002-RBO-5 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam: Not used on the 2008 or 2009 exams Question Cognitive Level: Memory or Fundamental Knowledge

Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 7 55.43 Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Comments:

Facility: Nine Mile Point Unit 1 Vendor: GE Exam Date: 2010 Exam Type: R Examination Outline Cross-reference: level RO SRO Tier # 2 Group # 1 KIA # 300000 A3.02 Importance Rating 2.9 Ability to monitor automatic operations of the INSTRUMENT AIR SYSTEM including: Air temperature Proposed Question: RO Question # 24 The plant is operating at 100% power with the following:

  • 11 Instrument Air Compressor (lAC) is red flagged as the backup compressor
  • 12 Instrument Air Compressor is tagged out and NOT available
  • 13 Instrument Air Compressor is red flagged Then, a TBClC TCV failure results in the associated lAC intercooler and aftercooler air temperatures exceeding 450oF.

Which one of the following describes the automatic operations of the Instrument Air system?

A. lAC 11 trips on high air temperature. 94-91, IA Inter-Tie BV, closes when lAC 11 trips.

B. lAC 13 trips on high air temperature. 94-91, IA Inter-Tie BV, closes when lAC 13 trips.

C. lAC 11 trips on high air temperature. 94-91, IA Inter-Tie BV, closes on lowering IA pressure.

D. lAC 13 trips on high air temperature. 94-91, IA Inter-Tie BV, closes on lowering IA pressure.

Proposed Answer: B Explanation (Optional):

A. Incorrect - RBClC supplies compressors 11 & 12, TBClC supplies compressor 13 and

BV 94-91 auto-closes when lAC 13 trips.

B. Correct - RBCLC supplies compressors 11 & 12, TBCLC supplies compressor 13. On a high TBCLC temperature, lAC 13 will operate until it trips on high air temperature.

When lAC 13 trips BV 94-91 separates a NSR portion of the lAS from the rest of the system. The valve auto-closes upon a LOOP or a trip of lAC 13 to conserve air for use by SR end users. In addition, with its control switch in AUTO, 94-91 will isolate if downstream instrument air pressure lowers to approximately 89 psig. However in this situation IA pressure will not lower until after lAC 13 trips.

C. Incorrect - RBCLC supplies compressors 11 & 12, TBCLC supplies compressor 13 and BV 94-91 auto-closes when lAC 13 trips.

D. Incorrect - When lAC 13 trips BV 94-91 separates a NSR portion of the lAS from the rest of the system.

Technical Reference(s): N1-0P-20, Sect B.2.1 and B.2.2 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None Learning Objective: N1-278001-RBO 5 and 8 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam: Not used on the 2008 or 2009 exams Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content 55.41 4 55.43 Secondary coolant and auxiliary systems that affect the facility.

Comments:

Facility: Nine Mile Point Unit 1 Vendor: GE Exam Date: 2010 Exam Type: RO Examination Outline Cross-reference: Level RO SRO Tier# 2 Group # 1 KIA # 262002 A4.01 Importance Rating 2.8 Ability to manually operate and/or monitor in the control room: Transfer from alternative source to preferred source Proposed Question: RO Question # 25 The plant is operating at 100% power with the following:

  • Lines 1 and 4 de-energize
  • Both EDGs restore power to their respective powerboards Which one of the following describes the effect of this transient on UPS 162 and the corresponding Control Room indication for UPS 162?

UPS 162 ...

A. Must be manually transferred back to the normal power source. Annunciator A3-1-2, RPS UPS 162 Trouble, automatically clears when the transfer occurs.

B. Must be manually transferred back to the normal power source. Annunciator A3-1-2, RPS UPS 162 Trouble, clears once the local trouble alarm is manually reset.

C. Automatically transfers back to the normal power source. Annunciator A3-1-2, RPS UPS 162 Trouble, automatically clears when the transfer occurs.

D. Automatically transfers back to the normal power source. Annunciator A3-1-2, RPS UPS 162 Trouble, clears once the local trouble alarm is manually reset.

Proposed Answer: D Explanation (Optional):

A. Incorrect - UPS 162 automatically transfers back to PB 16B approximately 30 seconds after PB 16B re-energizes. A3-1-2 does not reset until an operator manually resets the local trouble alarm.

B. Incorrect - UPS 162 automatically transfers back to PB 168 approximately 30 seconds after PB 16B re-energizes.

C. Incorrect - A3-1-2 does not reset until an operator manually resets the local trouble alarm.

D. Correct - UPS 162 automatically transfers back to PB 16B approximately 30 seconds after PB 16B re-energizes. A3-1-2 does not reset until an operator manually resets the local trouble alarm.

. N 1-ARP-A3-1-2 Technical Reference(s): OP-40 section B (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None Learning Objective: N 1-262002-RBO-5 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam: Not used on 2008 or 2009 exams Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Comments:

Facility: Nine Mile Point Unit 1 Vendor: GE Exam Date: 2010 Exam Type: RO Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 KIA # 215003 A1.05 Importance Rating 3.9 Ability to predict and/or monitor changes in parameters associated with operating the INTERMEDIATE RANGE MONITOR (IRM) SYSTEM controls including: SCRAM and rod block trip setpoints Proposed Question: RO Question # 26 A plant startup and heatup are in progress with the following:

  • IRM channel 11 is reading 86 out of 125 on range 8
  • REFUEL INST TRIP BYPASS 11 switch is in COINCIDENT
  • REFUEL INST TRIP BYPASS 12 switch is in COINCIDENT Then, the operator down ranges IRM '11 to range 7.

Which one of the following identifies the plant response?

A An upscale rod block is received, only B. A downscale rod block is received, only C. A rod block and a half scram is received D. A rod block and a full scram is received Proposed Answer: C Explanation (Optional):

A Incorrect - The lower scale (range 7) is upscale based on the range 8 reading; alarm, rod block, half scram.

B. Incorrect - The lower scale (range 7) is upscale based on the range 8 reading; alarm, rod block, half scram.

C. Correct - Upscale neutron flux level half scram. The trips of four of the IRM monitors are incorporated in logic channel 11 and the trips of the other four IRM monitors are incorporated in logic channel 12. With the REFUEL INST TRIP BYPASS 11 and 12 switches in the NON-COINCIDENT position, a single IRM channellNOP or UPSCALE TRIP will cause a Reactor Scram. In the COINCIDENT position, both channel 11 and channel 12 RPS must trip. When the IRM is down ranged, it is above the trip setpoint of the lower range.

D. Incorrect -If in NON-coincident a full scram would have occurred.

C-19859-C Technical Reference(s): N1-0P-38A (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None Learning Objective: N1-215000-RBO 7 (As available)

Question Source: Bank # 1& C Bank# 176 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: Not used in 2009 exam, similar to #6 on 2008 exam.

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 7 55.43 DeSign, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Comments:

Facility: Nine Mile Point Unit 1 Vendor: GE Exam Date: 2010 Exam Type: RO Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 2 KIA # 259001 K1.05 Importance Rating 3.2 Knowledge of the physical connections and/or cause- effect relationships between REACTOR FEEDWATER SYSTEM and the following: Condensate system Proposed Question: RO Question # 27 The plant is operating at 45% power with the following:

  • Condensate pumps 11 and 12 are running
  • Feedwater Booster pumps (FWBPs) 12 and 13 are running
  • Condensate pump 13 is tagged out Then, Condensate pump 11 trips due to an electrical fault.

Which one of the following describes the plant response and the required operator action?

Plant Response Required Operator Action A. FWP 13 trips on low Immediately scram the Reactor suction pressure due to lowering RPV water level B. FWP 13 trips on low Start FWPs 11 and 12 to prevent suction pressure a scram on low RPV water level C. FWP 13 suction pressure Perform an Emergency Power lowers but no FWP trip occurs Reduction to prevent damage to FWP 13 D. FWP 13 suction pressure Monitor FWBP suction pressure lowers but no FWP trip occurs and lower power if necessary to maintain Condensate pump 12 < 135 amps

Proposed Answer: D Explanation (Optional):

A. Incorrect - FWP will not trip on low suction pressure, there is no 13 FWP low suction pressure trip. This was part of the mod that made 13 FWP 100% capacity. Because the Condensate pumps are 50% capacity pumps there is no need for a reactor scram.

B. Incorrect - FWP will not trip on low suction pressure, there is no 13 FWP low suction pressure trip. This was part of the mod that made 13 FWP 100% capacity. Because the Condensate pumps are 50% capacity pumps there is no trip of FWP 13 or need to start FWPs 11 and 12.

C. Incorrect - FWP will not trip on low suction pressure, there is no 13 FWP low suction pressure trip. This was part of the mod that made 13 FWP 100% capacity. Because the Condensate pumps are 50% capacity pump there is no need for an emergency power reduction based on damaging FWP 13.

D. Correct - FWP will not trip on low suction pressure, there is no 13 FWP low suction pressure trip. This was part of the mod that made 13 FWP 100% capacity. Additionally the Condensate Pumps are 50% capacity pumps that will provide adequate flow at this power level.

N1-0P-16, Sect. B C-19853-C sht 1 Technical Reference(s): (Attach if not previously provided)

ARP H3-1-3 N1-0P-15A, Sect B Proposed References to be provided to applicants during examination: None Learning Objective: N1-259001-RBO 5 (As available)

Question Source: Bank #

Modified Bank # (l\Iote changes or attach parent)

New X Question History: Last NRC Exam: Not used on the 2008 or 2009 exams Question Cognitive Level: Memory or Fundamental Knowledge

Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 5 55.43 Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.

Comments:

Facility: Nine Mile Point Unit 1 Vendor: GE Exam Date: 2010 Exam Type: RO Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 2 KIA # 219000 K2.01 Importance Rating 2.5 Knowledge of electrical power supplies to the following: Valves (RHRlLPCI Torus Cooling)

Proposed Question: RO Question # 28 The plant is operating at 85% power with the following:

  • ERV 111 inadvertently opened and is now closed
  • Torus water temperature is 86°F and steady
  • Powerboard 167 is NOT available Which one of the following states the Containment Spray loops that can be placed in Torus Cooling from the Control Room, if any?

A. 111, 112, 121 or 122 B. 111 or 112, only C. 121 or 122, only D. None are available Proposed Answer: D Explanation (Optional):

A. Incorrect - BV-80-118 is the only valve (common to placing any of the 4 loops in service) that restricts the ability to place at least one of the loops in service. All other valves to be operated in aligning for torus cooling are DC-solenoid operated, AOVs.

B. Incorrect - BV-80-118 is the only valve (common to placing any of the 4 loops in service) that restricts the ability to place at least one of the loops in service. All other

valves to be operated in aligning for torus cooling are DC-solenoid operated, AOVs.

C. Incorrect - BV-80-118 is the only valve (common to placing any of the 4 loops in service) that restricts the ability to place at least one of the loops in service. All other valves to be operated in aligning for torus cooling are DC-solenoid operated, AOVs.

D. Correct - Without PB 167 (600 VAC), can't open BV-80-118, which is necessary to place any of the 4 loops in torus cooling.

Technical Reference(s): ~1-0P-14,C-18012-C, SOP-1.4 att (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None Learning Objective: N 1-22600 1-RBO-4 (As available)

Question Source: Bank # Water Sys # 125 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: Not used on 2008 exam, vaguely similar to question #58 on 2009 exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 7 55.43 Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Comments:

Facility: Nine Mile Point Unit 1 Vendor: GE Exam Date: 2010 Exam Type: RO Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 2 KIA # 271000 K3.01 Importance Rating 3.5 Knowledge of the effect that a loss or malfunction of the OFFGAS SYSTEM will have on following: Condenser vacuum Proposed Question: RO Question # 29 The plant is operating at 100% power when the following annunciators are received:

  • H1-1-7, OFF GAS HIGH RADIATION
  • H1-2-7, OFF GAS RAD MON 11-12 FILTER DP SAMPLE FLOW
  • H1-3-7, MN CONDR OFF GAS TIMER STARTED The Off Gas rad monitors are confirmed to have tripped on high-high level.

Which one of the following describes the resulting automatic action(s) and the required manual action?

Automatic Action(s) Required Manual Action A. Stack Blocking Valve, Immediately scram the reactor due to rapidly 77 -03, closes lowering Main Condenser vacuum B. Interstage Blocking Valves, Immediately scram the reactor due to rapidly 76-12 and 76-13, close lowering Main Condenser vacuum C. Stack Blocking Valve, Perform an emergency power reduction to attempt 77 -03, closes to control radiation levels and stabilize Main Condenser vacuum D. Interstage Blocking Valves, Perform an emergency power reduction to attempt 76-12 and 76-13, close to control radiation levels and stabilize Main Condenser vacuum

Proposed Answer: C Explanation (Optional):

A. Incorrect - The vacuum loss following closure of 77-03 is not rapid. Sufficient volume exists in the offgas system piping to continue to remove noncondensible gases from the main condenser for a period of time, during which vacuum will slowly degrade.

B. Incorrect - Blocking Valves 76-12/76-13 do NOT close. Hi pressure/hi temp provides auto closure signals to Blocking Valves 76-12/76-13 (interstage blocking valves).

C. Correct - BV 77-03 closes on high-high offgas radiation. The vacuum loss following closure of 77-03 is not rapid. Sufficient volume exists in the offgas system piping to continue to remove noncondensible gases from the main condenser for a period of time, during which vacuum will slowly degrade. SOP-25.1 and 25.2 direct an emergency power reduction to control radiation/vacuum.

D. Incorrect - Blocking Valves 76-12/76-13 do not close. Hi pressure/hi temp provides auto closure signals to Blocking Valves 76-12176-13 (interstage blocking valves),

Technical Reference(s): N1-S0P-25.1, 25,2 (Attach if not previously provided)

N1-ARP-H 1-3-7 Proposed References to be provided to applicants during examination: None Learning Objective: N 1-271 000-RBO-5 (As available)

Question Source: Bank #

'fi dBan k # 05-Q-01 Mod lie ID: N1-271000-RBO- (N 0 t e changes or attach paren t)

New Question History: Last NRC Exam: Not used on 2008 exam or 2009 exams similar to 2009 #71 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 11 55.43

Purpose and operation of radiation monitoring systems, including alarms and survey equipment.

Comments:

Facility: Nine Mile Point Unit 1 Vendor: GE Exam Date: 2010 Exam Type: RO Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 2 KIA # 272000 K4.03 Importance Rating 3.6 Knowledge of RADIATION MONITORING System design feature(s) and/or interlocks which provide for the following: Fail safe tripping of process radiation monitoring logic during conditions of instrument failure.

Proposed Question: RO Question # 30 The plant is operating at 100% power with the following:

  • Off Gas Effluent Stack Monitoring System (OGESMS) radiation monitor RN10a fails downscale
  • A technician sent to investigate the problem inadvertently moves the RN 10-'2 mode switch from OPERATE to STANDBY
  • The white INOP light for RN10-'2 illuminates Which one of the following describes the resulting plant response?

A. OGESMS isolation signal is received, only B. Containment Purge valve isolation signal is received, only C. OGESMS and Containment Purge valve isolation signal is received D. No automatic isolation signals are received, only the stack gas trouble alarm is received Proposed Answer: B Explanation (Optional):

A. Incorrect - OGESMS does not have any operable rad monitors, however the system does NOT isolate automatically.

B. Correct - The OGESMS includes trips and alarms for downscale, INOP, and high-high level stack radiation. All trips are initiated by the RN10A and RN10B Monitors, and result in an isolation signal to the 20" and 24" vent and purge valves. The logic is arranged so that any two-out-of-two combinations of high-high, INOP, or downscale from both Monitors will complete the trip function. In addition, a high stack radiation level sensed by RN10AlB will generate an alarm only.

C. Incorrect - OGESMS does not have any operable rad monitors, however the system does NOT isolate automatically.

D. Incorrect - The logic is arranged so that any two-out-of-two combinations of high-high, INOP, or downscale from both Monitors will complete the trip function. The stack gas trouble alarm is received.

Technical Reference(s): N1-0P-50B, Sect. B.4.4 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None Learning Objective: N1-272000-RBO-11 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New x Not used on 2008 or 2009 exam.

Question History: Last NRC Exam: 2008 #17 involved RN10 monitors Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 11 55.43 Purpose and operation of radiation monitoring systems, including alarms and survey equipment.

Comments:

Facility: Nine Mile Point Unit 1 Vendor: GE Exam Date: 2010 Exam Type: RO Examination Outline Cross-reference: Level RO SRO Tier# 2 Group # 2 KIA # 288000 K5.01 Importance Rating 3.1 Knowledge of the operational implications of the following concepts as they apply to PLANT VENTILATION SYSTEMS: Airborne contamination control Proposed Question: RO Question # 31 Following a steam line break in the Secondary Containment, the following events occurred:

  • Reactor Building Emergency Ventilation System (RBEVS) automatically initiated
  • Reactor Building (RB) Ventilation remains in operation Which one of the following is the significance of these conditions?

A. RBEVS flow is short-cycling the Reactor Building.

B. The operating RB Ventilation system prevents flow through RBEVS.

C. The operating RB Ventilation system is an unfiltered radioactive discharge.

D. Parallel operation of RBEVS and RB Ventilation will cause Secondary Containment pressure to become positive.

Proposed Answer: C Explanation (Optional):

A. Incorrect - The RB ventilation fans would be operating in parallel with the RBEVS fans, there is no path back through the RB vent fans. Short cycling or recirculation through the RBEVS will not occur.

B. Incorrect - The RBEVS parallel with the normal RB Ventilation systems will not prevent

RBEVS flow.

C. Correct - If Reactor Building Ventilation exhaust radiation exceeds the isolation setpoint the normal ventilation should automatically isolate and EVS should automatically start.

RBEVS realigns to send potential radioactive release particles through the filtered train associated with RBEVS and prevent unfiltered release through the normal RB Vent Discharge. These actions should be verified to limit radioactivity release to the environment and maintain the desired negative differential pressure.

D. Incorrect - The Secondary Containment is maintained at a negative pressure with both systems operating.

EOP bases for N1-EOP-5 Technical Reference(s): C18013C (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None Learning Objective: N1-288000-RBO-5 and RBO-8 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam: Not used on the 2008 or 2009 exams Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 13 55.43 Procedures and equipment available for handling and disposal of radioactive materials and effluents.

Comments:

Facility: Nine Mile Point Unit 1 Vendor: GE Exam Date: 2010 Exam Type: R Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 2 KIA # 201002 K6.01 Importance Rating 2.5 Knowledge of the effect that a loss or malfunction of the following will have on the REACTOR MANUAL CONTROL SYSTEM: Select matrix power Proposed Question: RO Question # 32 The plant is operating at 35% power with the following:

  • A plant shutdown is in progress
  • Control rod 26-11 is being continuously inserted using the normal Rod Movement Control Switch
  • Then, Rod Select Power is lost while control rod 26-11 is passing through position 24 Which one of the following describes the affect of this loss?

When Rod Select Power is lost, the control rod motion wilL ..

A. Stop. The Settle Directional Control Valve, SOV-120, will NOT open.

B. Stop. The Settle Directional Control Valve, SOV-120, will open to settle the rod.

C. Continue until the Rod Movement Control Switch is released. The Settle Directional Control Valve, SOV-120, will NOT open.

D. Continue until the Rod Movement Control Switch is released. The Settle Directional Control Valve, SOV-120, will open to settle the rod.

Proposed Answer: A Explanation (Optional):

A. Correct - The CONTROL ROD POWER switch 4S4 on Panel E is used to remove power from the rod select and positioning circuits when the system is not required for rod movement. Removing power from the select bus removes power from the timer and the control positioning valves at the HCU.

B. Incorrect - The Settle DCV will not be opened by the timer.

C. Incorrect - Rod motion will immediately be stopped as the Insert DCVs are de energized.

D. Incorrect - Rod motion will immediately be stopped as the Insert DCVs are de energized.

N1-0P-5, Sect B Technical Reference(s): C-22030-C, Sht 1 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None Learning Objective: N 1-20 1002-RBO-5 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam: Not used on the 2008 or 2009 exams Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 6 55.43 Design, components, and function of reactivity control mechanisms and instrumentation.

Comments:

Facility: Nine Mile Point Unit 1 Vendor: GE Exam Date: 2010 Exam Type: RO Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 2 KIA # 204000 A 1.03 Importance Rating 2.8 Ability to predict and/or monitor changes in parameters associated with operating the REACTOR WATER CLEANUP SYSTEM controls including: Reactor water temperature Proposed Question: RO Question # 33 A plant transient has resulted in the following:

  • RPV pressure and temperature are being controlled using EOP-1 attachment 9, RPV Press Control Thru RWCU Temperature
  • RPV pressure is 800 psig and steady
  • Recirc suction temperature is 520°F and steady Then, the following occur:
  • An Operator further opens 70-85, BV-CU NON REGEN HX RBCLC OUTLET
  • A momentary pressure detector malfunction causes the RWCU system pressure signal to rise to 118 psig before returning to normal Which one of the following describes the effect of these events on Recirc suction temperature and RWCU system status?

Recirc suction temperature ...

A. Rises, because RWCU isolated on high system pressure B. Rises, because more RBCLC flow is bypassing the NRHX C. Lowers, because more RWCU flow is being supplied to the NRHX D. Lowers, because more RBCLC flow is being supplied to the NRHX

Proposed Answer: D Explanation (Optional):

A. Incorrect - RWCU isolates on high system pressure of >130 psig for 9 seconds, therefore no isolation occurs in this situtation. Additionally, EOP-1 Att 9 bypasses all RWCU isolations.

B. Incorrect - Opening 70-85, BV-CU NONREGEN HX RBCLC OUTLET will increase RBCLC flow through the H/X further cooling the reactor coolant.

C. Incorrect 85, BV-CU NON REGEN HX RBCLC OUTLET throttles the RBCLC water not the RWCU water. RWCU flow is un-affected.

D. Correct - Opening 70-85, BV-CU NONREGEN HX RBCLC OUTLET will increase RBCLC flow through the H/X further cooling the reactor coolant.

. EOP-1 Att 9 Techmcal Reference(s): C-18009-C (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None Learning Objective: N1-204000-RBO-1 0 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam: Not used on 2008 or 2009 exams Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 5 55.43 Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.

Comments:

Facility: Nine Mile Point Unit 1 Vendor: GE Exam Date: 2010 Exam Type: RO Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 2 KIA # 201003 A2.04 Importance Rating 3.5 Ability to predict and/or monitor changes in parameters associated with operating the CONTROL ROD AND DRIVE MECHANISM controls including: Single control rod SCRAM Proposed Question: RO Question # 34 The plant has experienced a failure to scram with the following:

  • RPV pressure is 900 psig and steady
  • N1-EOP-3.1, Alternate Control Rod Insertion, has been entered
  • RPS jumpers have been inserted
  • ARI has been overridden
  • The scram has been reset Which one of the following methods, in accordance with N1-EOP-3.1, will result in the largest overall differential pressure across the CRD piston for inserting control rods?

A. Open individual scram test switches B. Vent the control rod over piston area C. Drive control rods using maximum drive pressure D. Maximize CRD cooling water differential pressure Proposed Answer: A Explanation (Optional):

A. Correct - Of the choices presented, scram test switch operation would result in the largest ilp across the CRDM piston. The ilp is the difference between the HCU

accumulator at charging pressure of 1390 to 1510 psig and the atmospheric pressure in the SDV.

B. Incorrect - Venting the over piston area places the Ap between the reactor and a hose, across the piston. The higher charging header pressure will not be seen by the piston because the scram is reset. Maximum Ap would be -900 psig.

C. Incorrect - CRD drive water pressure is regulated between CRD pressure and reactor pressure; the maximum Ap would be -500 psig.

D. Incorrect - CRD cooling water pressure is regulated between CRD pressure and reactor pressure; the maximum Ap would be -500 psig.

Technical Reference(s): N1-0P-5 (Attach if not previously provided)

C-1B016-C Proposed References to be provided to applicants during examination: None Learning Objective: N1-201001-RBO-6 (As available)

ID: N1-201001-RBO-Question Source: Bank #

07-Q-01 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: Not used on 200B or 2009 exams Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 6 55.43 Design, components, and function of reactivity control mechanisms and instrumentation.

Comments:

Facility: Nine Mile Point Unit 1 Vendor: GE Exam Date: 2010 Exam Type: RO Examination Outline Cross-reference: Level RO SRO Tier# 2 Group # 2 KIA # 223001 A3.05

- - - - .... --.~

Importance Rating 4.3 Ability to monitor automatic operations of the PRIMARY CONTAINMENT SYSTEM AND AUXILIARIES including: Drywell pressure Proposed Question: RO Question # 35 A LOCA has resulted in the following:

  • Drywell (DW) pressure is 4.0 psig and slowly lowering
  • Torus pressure is 4.8 psig and slowly lowering
  • Reactor Building (RB) pressure is -0.25" H20 and steady Which one of the following identifies the status of the Torus to DW Vacuum Breakers and the RB to Torus Vacuum Breakers?

Torus to DW Vacuum Breakers RB to Torus Vacuum Breakers A. OPEN OPEN B. OPEN CLOSED C. CLOSED OPEN D. CLOSED CLOSED Proposed Answer: B Explanation (Optional):

A. Incorrect - RBltorus vacuum breakers do NOT open because torus pressure is higher than RB pressure.

B. Correct - Torus/OW vacuum breakers open because torus pressure is greater than OW pressure (by more than 0.5 psid). RB/torus vacuum breakers do NOT open because torus pressure is higher than RB pressure.

C. Incorrect - Torus/OW vacuum breakers open because torus pressure is greater than OW pressure. The RB to Torus Vacuum Breakers would remain closed because the Torus pressure is above the Reactor Building pressure.

O. Incorrect - Torus/OW vacuum breakers open because torus pressure is greater than OW pressure.

N1-0P-9, Sect. 0, P & L 10 and 12 Technical Reference(s): (Attach if not previously provided)

N1-ARP-K1,4-6 TS 3.3.6 basis Proposed References to be provided to applicants during examination: None Learning Objective: N1-223001-RBO-6 (As available) 10: N1-223001-RBO-Question Source: Bank #

06-Q-01 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: Not used on the 2008 or 2009 exams Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 5 55.43 Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.

Comments:

Facility: Nine Mile Point Unit 1 Vendor: GE Exam Date: 2010 Exam Type: RO Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 2 KIA # 201006 A4.04 Importance Rating 3.3 Ability to manually operate and/or monitor in the control room: Rod withdrawal error indication:

P-Spec(Not-BWR6) (Rod Worth Minimizer System (RWM))

Proposed Question: RO Question # 36 A plant startup is in progress with the following:

  • The current rod group requires twelve (12) rods to be withdrawn from position 00 to 08
  • Eight (8) of the rods have been withdrawn to position 08
  • The ninth rod has been withdrawn to position 10
  • Then, the Operator selects the tenth rod Which one of the following describes the response of the Rod Worth Minimizer?

A. One withdraw error is indicated, and further control rod withdrawal is allowed.

B. One select error is indicated, only, and further control rod withdrawal will be blocked.

C. One withdraw error is indicated, only, and further control rod withdrawal will be blocked.

D. One withdraw error and one select error is indicated, and further control rod withdrawal will be blocked.

Proposed Answer: D Explanation (Optional):

A. Incorrect - If the operator attempts a rod withdrawal that deviates by one notch from the selected program, the RWM blocks such action.

B. Incorrect - A withdraw error is indicated because a rod has been withdrawn past the group limit.

C. Incorrect - A Select error is indicated because a rod other than the offending rod has been selected.

D. Correct - A withdraw error is indicated because a rod has been withdrawn past the group limit. A Select error is indicated because a rod other than the offending rod has been selected. Further control rod withdrawal is blocked with a single withdraw error.

Technical Reference(s): N1-0P-37, Sect. B (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None Learning Objective: N 1-201 003-RBO-5 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam: Not used on the 2008 or 2009 exams Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 6 55.43 Design, components, and function of reactivity control mechanisms and instrumentation.

Comments:

Facility: Nine Mile Point Unit 1 Vendor: GE Exam Date: 2010 Exam Type: RO Examination Outline Cross-reference: Level RO SRO Tier# 2 Group # 2 KIA # 214000 2.2.44 Importance Rating 4.2 Equipment Control: 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. (RPIS)

Proposed Question: RO Question # 37 Which one of the following describes the rod position indication on the full core display immediately following a scram and after the operator resets the scram?

Immediately following the scram, control rods would have a (1) . After the scram is reset, the control rods would have a (2) o.

(1 ) (2)

A. Green-green backlight with Green-green backlight with blank digits indication 00 digits indication B. Green-green backlight with No backlight with blank digits indication 00 digits indication C. Green-green backlight with Green-green backlight with 00 digits indication 00 digits indication D. Green-green backlight with No backlight with 00 digits indication 00 digits indication Proposed Answer: A Explanation (Optional):

A. Correct - The un-reset scram signal applies CRD/Reactor pressure under the CRDM

piston to hold the rod in beyond the 00 position. The number 51 reed switch keeps the green-green background lights energized as the control rod is inserted beyond the 00 position, which results in blank windows with the green-green background lights illuminated. When the operator resets the scram, the scram valves close and the control rod drive will settle on the collet fingers for position 00 and the 00 indication will illuminate. Another reed switch (switch 52) will illuminate the green-green background at position 00.

B. Incorrect - A reed switch (switch 52) will illuminate the green-green background at position 00 once the scram is reset.

C. Incorrect - The un-reset scram signal applies CRD/Reactor pressure under the CRDM piston to hold the rod in beyond the 00 position.

D. Incorrect - The un-reset scram signal applies CRD/Reactor pressure under the CRDM piston to hold the rod in beyond the 00 position. The number 51 reed switch keeps the green-green background lights energized as the control rod is inserted beyond the 00 position. Another reed switch (switch 52) will illuminate the green-green background at position 00 once the scram is reset.

. SOP-1 Technical Reference(s): N11 01201 002C01 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None Learning Objective: N1-201002-RBO-10 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam: not used on the 2008 or 2009 exams Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 6 55.43 Design, components, and function of reactivity control mechanisms and instrumentation.

Comments:

Facility: Nine Mile Point Unit 1 Vendor: GE Exam Date: 2010 Exam Type: RO Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 2 KIA # 256000 K2.01 Importance Rating 2.7 Knowledge of electrical power supplies to the following: System pumps (Condensate System)

Proposed Question: RO Question # 38 The plant is operating at 75% power with the following:

  • Condensate pumps 11 and 12 are running
  • Then, a Reactor scram occurs Which one of the following describes the electrical power supplies to the Condensate Pumps five (5) minutes after the Reactor scram?

Condensate Pump 11 Condensate Pump 12 A. Powerboard 101 from a Powerboard 12 from a Reserve Transformer Reserve Transformer B. Powerboard 11 from a Powerboard 101 from a Reserve Transformer Reserve Transformer C. Powerboard 101 from the Powerboard 12 from the Station Service Transformer Station Service Transformer D. Powerboard 11 from the Powerboard 101 from the Station Service Transformer Station Service Transformer Proposed Answer: B Explanation (Optional):

A. Incorrect - Condensate Pump 11 is powered by P.B. #11, Condensate Pump 12 is powered by P.B. #101.

B. Correct -Immediately following the scram, bus transfers result in Reserve Transformers powering auxiliary loads. Condensate Pump 11 is powered by P.B. #11, Condensate Pump 12 is powered by P.B. #101, Condensate Pump 13 is powered by PB#12, C. Incorrect - Condensate Pump 11 is powered by P.B. #11, Condensate Pump 12 is powered by P.B. #101. Bus transfers result in Reserve Transformers powering auxiliary loads.

D. Incorrect - Bus transfers result in Reserve Transformers powering auxiliary loads.

Technical Reference(s): C-19409-C, Sht 1b (Attach if not previously provided)

N1-0P-30, Sect. B.1.0 Proposed References to be provided to applicants during examination: None Learning Objective: N1-256000-RBO-1 0 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam: Not used on 2008 or 2009 exams Question Cognitive Level: Memory or Fundamental Knowledge x Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Comments:

Facility: Nine Mile Point Unit 1 Vendor: GE Exam Date: 2010 Exam Type: RO Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 KIA # 295030 EK1.03 Importance Rating 3.8 Knowledge of the operational implications of the following concepts as they apply to LOW SUPPRESSION POOL WATER LEVEl: Heat capacity Proposed Question: RO Question # 39 Given the following conditions:

  • Torus water level is 0.5 ft below the low Torus water level EOP entry condition
  • Torus water temperature is 135°F Which one of the following lists (1) the HIGHEST of the given RPV pressures at which an RPV blowdown can be initiated within the Heat Capacity Temperature limit AND (2) the basis for this limit?

IMI Heat Capacity Temperature Limit 1&0 BAD 170"F@15~ '"""""" ......... ......... ....

I ........ ......... .........

G010D ......... ~ r-.... 100.. r--.... ..... T 10.5-11.25!1 r--. ......

100 H Use EO""'lAtt23 S:lr to'U$ veter IeoIBls hebvO,Qft or above11O!!

I I

111"F@ IO!!Opeig ,Q-105and rl'11.2S-13.0iI 80 o laD 200 300 -4OD SIlO 600 700 BOO 00<l 1000 1100 RPV Pfessure (psig)

Figure 3-11: Heat Capacity Temperature Limit

(2)

A. 4S0 psig Maintain Drywell temperature within its design limit B. 700 psig Maintain Drywell temperature within its design limit C. 4S0 psig Maintain Torus pressure below the Primary Containment Pressure Limit D. 700 psig Maintain Torus pressure below the Primary Containment Pressure Limit Proposed Answer: C Explanation (Optional):

A. Incorrect - The basis for the limit is to maintain Torus pressure below the Primary Containment Pressure Limit. By maintaining this limit the plant is assured Drywell pressure will not exceed the design basis.

B. Incorrect - 700 psig is above the limit for this Torus level. The basis for the limit is to maintain Torus pressure below the Primary Containment Pressure Limit. By maintaining this limit the plant is assured Drywell pressure will not exceed the design basis.

C. Correct - Answer based on Torus level of 10ft (O.S ft below EOP-4 entry condition of 10.S ft), so the lower HCTL curve is the limiting curve. For 13soF in the Torus, the highest pressure that can be within the HCTL is -SOO psig. Therefore 4S0 psig is the highest of the given pressures that is within the HCTL.

The Heat Capacity Temperature Limit (Figure 3-11) is the highest torus water temperature from which a blowdown will not raise Torus pressure above the Primary Containment Pressure Limit D. Incorrect - 700 psig is above the limit for this Torus level.

Technical Reference(s): EOP Bases, Sect 3.6 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None Learning Objective: 01-0PS-006-344-1-04, EO 1.S (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Not used on 2008, 2009 NRC #15 has Question History: Last NRC Exam: the same KIA but asks a different question.

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

Facility: Nine Mile Point Unit 1 Vendor: GE Exam Date: 2010 Exam Type: RO Examination Outline Cross-reference: Level RO SRO Tier# 1 Group # 1 KIA # 295024 EK1.01 Importance Rating 4.1 Knowledge of the operational implications of the following concepts as they apply to HIGH DRYWELL PRESSURE: Drywell integrity: Plant-Specific Proposed Question: RO Question # 40 A LOCA has resulted in the following:

  • Torus pressure is 42 psig and rising
  • Drywell pressure is 44 psig and rising
  • Drywell temperature is 280"F and rising
  • Reactor pressure is 75 psig and stable
  • Containment water level is 14 feet and rising
  • Containment Spray has been verified in service Which one of the following describes the operator action necessary to preserve the Primary Containment integrity and the bases for the action?

Execute N1-EOP-8, RPV Blowdown ...

A. and then vent the Torus to prevent exceeding the Pressure Suppression Pressure.

B. and then vent the Torus to prevent exceeding the Primary Containment Pressure Limit.

C. to prevent exceeding the Pressure Suppression Pressure. Containment venting will NOT be required.

D. to prevent exceeding the Primary Containment Pressure Limit. Containment venting will NOT be required.

Proposed Answer: B

Explanation (Optional):

A. Incorrect - The basis for blowdown and venting in these conditions is to avoid exceeding the pressure limit of the Containment. PSP has already been violated.

B. Correct - The given conditions are about to violate PCPL. EOP-4 requires RPV Blowdown and then venting the containment to prevent exceeding PCPL. With water level 14', the Torus is used for venting. The basis for this region of the PCPL curve is to avoid exceeding the pressure limit of the Containment.

C. Incorrect - Containment pressure is about to reach the PCPL curve, therefore venting is required. With Reactor pressure already down to 75 psig, the blowdown is not going to have any appreciable effect on preventing PCPL violation. PSP has already been violated.

O. Incorrect - Containment pressure is about to reach the PCPL curve, therefore venting is required. With Reactor pressure already down to 75 psig, the blowdown is not going to have any appreciable effect on preventing PCPL violation.

EOP Bases Technical Reference(s): EOP-4 (Attach if not previously provided)

EOP-4 OW pressure Proposed References to be provided to applicants during examination: leg, PSP and PCPL curves Learning Objective: 01-0PS-006-344-1-04-EO-1.3 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam: Not used on the 2008 or 2009 exams Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

Facility: Nine Mile Point Unit 1 Vendor: GE Exam Date: 2010 Exam Type: RO Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 KIA # 295005 AK1.03 Importance Rating 3.5 Knowledge of the operational implications of the following concepts as they apply to MAIN TURBINE GENERATOR TRIP: Pressure effects on reactor level Proposed Question: RO Question # 41 The plant is operating at 17% power with the following:

  • RPV water level is 70" and stable Then, a Main Turbine trip occurs. Total Feedwater flow stabilizes at 1.3 Mlbm/hr once steady state conditions are achieved.

Which one of the following describes the effect on Reactor water level immediately after the trip and once steady-state conditions are reached?

Immediate RPV Level Response RPV level at steady-state A. Lowers Above 70" B. Lowers Below 70" C. Rises Above 70" D. Rises Below 70" Proposed Answer: A Explanation (Optional):

A. Correct. The sudden pressure rise from the TSV closure collapses voids and causes reactor level to shrink, lowering reactor level. The shaft-driven feedwater pump can provide feedwater flow of greater than 3800 gpm for approximately 3.2 minutes during pump coastdown. The turbine trip will signal the motor-driven feedwater pump to start.

Whenever a turbine trip or reactor low level scram occurs (+53"), the reactor feedwater control system will control FCVs 11 and 12 in the High Pressure Coolant Injection (HPCI) mode of control. The control system associated with each of the motor-driven feed pumps (when operating as HPCI) consists basically of a controller which operates either as a maximum flow controller or as a vessel level controller. Signal values are computed so that with the setpoint applied to the controller, flow will be held at a maximum of 3420 gpm or vessel level will be maintained at 72" under FW 12 control.

B. Incorrect - Signal values are computed so that with the setpoint applied to the controller, flow will be held at a maximum of 3420 gpm or vessel level will be maintained at 72" under FW 12 control.

C. Incorrect - The sudden pressure rise from the TSV closure collapses voids and causes reactor level to shrink, lowering reactor level.

D. Incorrect - The sudden pressure rise from the TSV closure collapses voids and causes reactor level to shrink, lowering reactor level. Signal values are computed so that with the setpoint applied to the controller, flow will be held at a maximum of 3420 gpm or vessel level will be maintained at 72" under FW 12 control.

N1-S0P-31.1 N1-0P-16, Sect. B Technical Reference(s): (Attach if not previously provided)

UFSAR, Sect. XV 3.14.1 N11012390001C01 pg 148 Proposed References to be provided to applicants during examination: None Learning Objective: N 1-245000-RBO-1 0 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam: Not used on the 2008 or 2009 exams Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X

10 CFR Part 55 Content: 55.41 5 55.43 Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.

Comments:

Facility: Nine Mile Point Unit 1 Vendor: GE Exam Date: 2010 Exam Type: RO Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 KIA # 29S028 EK2.04 Importance Rating 3.6 Knowledge of the interrelations between HIGH DRYWELL TEMPERATURE and the following:

Drywell ventilation Proposed Question: RO Question # 42 The plant is operating at 100% power with the following:

  • Six (6) Drywell cooling fans are running
  • Drywell average temperature is 134°F and stable Then, Drywell cooling fan 11 trips and cannot be restarted.

Which one of the following describes the expected magnitude of the Drywell temperature change and, based on this expected change, whether entry into EOP-4, Primary Containment Control, will be required?

Expected Drywell Temperature Change Entry into EOP-4 A. SOF Will be required B. 5°F Will NOT be required C. 20°F Will be required D. 20°F Will NOT be required Proposed Answer: B Explanation (Optional):

A. =

Incorrect - 134 + 5 139°F, which is below the 150°F Drywell average temperature EOP-4 entry condition B. Correct - OP-8 states the expected rise in Drywell average temperature from securing a

=

single cooling fan is 5°F. 134 + 5 139°F, which is below the 150°F Drywell average temperature EOP-4 entry condition.

C. Incorrect - OP-8 states the expected rise in Drywell average temperature from securing

=

a single cooling fan is 5°F. 134 + 5 139°F, which is below the 150°F Drywell average temperature EOP-4 entry condition.

D. Incorrect - OP-8 states the expected rise in Drywell average temperature from securing a single cooling fan is 5°F.

Technical Reference(s): N1-0P-8 section H.3.0 Note 1 (Attach if not previously provided)

N1-EOP-4 Proposed References to be provided to applicants during examination: None Learning Objective: N1-223001-RBO-10 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 5 55.43 Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.

Comments:

Facility: Nine Mile Point Unit 1 Vendor: GE Exam Date: 2010 Exam Type: RO Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 KIA # 295006 AK2.06 Importance Rating 4.2 Knowledge of the interrelations between SCRAM and the following: Reactor power Proposed Question: RO Question # 43 The plant is operating at 20% power with the following:

  • Drywell pressure is slowly rising
  • The CRS directs scramming the Reactor in accordance with SOP-1, Reactor Scram
  • An Operator places the Mode Switch in SHUTDOWN
  • APRM and RPIS indications are unavailable due to equipment failures
  • RPV water level lowers to 62 inches before recovering to normal
  • No EOPs have been entered
  • Four (4) minutes after the scram, the IRMs have been fully inserted and indicate upscale on Range 9 Which one of the following describes this indication and the required Operator actions?

A. This is a normal indication due to decay heat. Continue actions of SOP-1.

B. This is a normal indication due to delayed neutrons. Continue actions of SOP-1.

C. This is NOT a normal indication. Immediately enter EOP-3, Failure to Scram, on high Reactor power. No entry into EOP-2, RPV Control, is required.

D. This is NOT a normal indication. Immediately enter EOP-2, RPV Control, on high Reactor power and then transition to EOP-3, Failure to Scram.

Proposed Answer: D Explanation (Optional):

A. Incorrect - Decay heat would result in a thermal output of -3% at 100 seconds. The given CTP indication is approximately 8.6%.

B. Incorrect - Decay heat would result in a thermal output of -3% at 100 seconds. The given CTP indication is approximately 8.6%.

C. Incorrect - There are no directions to enter EOP-3 in SOP-1. At a minimum, EOP-2 entry is required before transitioning to EOP-3. EOP-2 contains the diagnostic steps to determine if EOP-3 entry is warranted.

D. Correct - Decay heat would result in a thermal output of -3% at 100 seconds. In this case, reactor power is above the 6% entry condition for EOP-2, RPV Control, since the upscale trip on IRM range 9 is calibrated for approximately 12% power. There are no directions to enter EOP-3 in SOP-1. Entry is made into EOP-2 which then directs entry into EOP-3.

Technical Reference(s): N1-EOP-2, UFSAR section V.E.1, (Att h'f t . I 'd d)

LSSS 2.1.1.b and bases ac I no prevIous y provi e Proposed References to be provided to applicants during examination: None Learning Objective: 01-0PS-006-342-1-01, EO-1.3 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam: Not used on the 2008 or 2009 exams Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 5 55.43 Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.

Comments:

Facility: Nine Mile Point Unit 1 Vendor: GE Exam Date: 2010 Exam Type: RO Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 KIA # 295025 EK2.01 Importance Rating 4.1 Knowledge of the interrelations between HIGH REACTOR PRESSURE and the following: RPS Proposed Question: RO Question # 44 A plant startup is in progress with the following:

  • Reactor pressure rises rapidly to a peak of 1085 psig
  • Reactor pressure is restored to a normal value within 7 seconds of the start of the transient Which one of the following describes the automatic plant response to this event?

A. Multiple alarms actuate. No Reactor scram, ERV actuation or Emergency Condenser initiation occurs.

B. Multiple alarms actuate and the Reactor scrams. No ERV actuation or Emergency Condenser initiation occurs.

C. Multiple alarms actuate, the Reactor scrams and two ERVs open. No Emergency Condenser initiation occurs.

D. Multiple alarms actuate, the Reactor scrams and Emergency Condensers initiate. No ERV actuation occurs.

Proposed Answer: B Explanation (Optional):

A. Incorrect - The Reactor scrams based on pressure exceeding 1080 psig. There is no time delay on this signal.

B. Correct - The Reactor scrams based on pressure exceeding 1080 psig. There is no time delay on this signal. The lowest ERV actuation setpoint is 1090 psig, which is not reached. The Emergency Condensers initiation only if pressure is above 1080 psig for more than 12 seconds. In this question, pressure is only above normal for 7 seconds, so no Emergency condenser initiation occurs. Alarms F1-1-2, F1-2-1, F2-3-4, F4-1-7 and F4-2-8 come in directly based on high pressure and the auto scram. Many other alarms will also be received due to integrated plant response to the scram.

C. Incorrect - The lowest ERV actuation setpoint is 1090 pSig, which is not reached.

D. Incorrect - The Emergency Condensers initiation only if pressure is above 1080 psig for more than 12 seconds. In this question, pressure is only above normal for 7 seconds, so no Emergency condenser initiation occurs.

N1-0P-13, section B N1-0P-1, section B Technical Reference(s): (Attach if not previously provided)

N1-0P-40, section B ARPs F1, F2, F4 Proposed References to be provided to applicants during examination: None Learning Objective: N 1-212000-RBO-5 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam: Not used on the 2008 or 2009 exams Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Comments:

Facility: Nine Mile Point Unit 1 Vendor: GE Exam Date: 2010 Exam Type: RO Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 KIA # 700000 AK3.02 Importance Rating 3.6 Knowledge of the reasons for the following responses as they apply to GENERATOR VOLTAGE AND ELECTRIC GRID DISTURBANCES: Actions contained in abnormal operating procedure for voltage and grid disturbances.

Proposed Question: RO Question # 45 The plant is operating at 100% power when a grid disturbance causes the following:

  • Computer points F432, F433, F434 indicate 115 KV Bus phase voltages are 113.2 KV and stable
  • 115 KV Bus frequency is 58.8 Hz and stable
  • The Load Tap Changers (LTCs) for T101 Nand T101S are in MANUAL at the 0-1 and 1r positions, respectively
  • The CRS has directed that the LTCs for T1 01 Nand T1 01 S be repositioned to the 1r and 2r positions, respectively Which one of the following is the reason for repositioning the T1 01 Nand T1 01 S LTCs?

A. Prevent Breakers R10 and R40 from opening on low voltage.

B. Prevent Breakers R10 and R40 from opening on low frequency.

C. Raise the voltage being supplied to Powerboards 101, 102 and 103.

D. Raise the frequency being supplied to Powerboards 101, 102 and 103.

Proposed Answer: C Explanation (Optional):

A. Incorrect - The LTCs affect voltage out of the transformers, a low voltage trip of R 10 and R40 occurs with a sustained low voltage into the transformers.

B. Incorrect - The Reserve Station Transformers 101 Nand 101 S are equipped with Load Tap Changers (LTC) capable of changing output voltage, not frequency.

C. Correct - The Reserve Station Transformers 101 Nand 101 S are equipped with Load Tap Changers (LTC) capable of changing output voltage over a range of +10% of 4160 volt busses. The Tap Changers are normally operated in the AUTO mode but can be operated in MANUAL. Each LTC is equipped with 17 positions; eight in the raise direction, eight in the lower direction, and one neutral position. The neutral position, represented as "0-1" on the digital display, will result in supplying 4160 volts to affected busses with 115 kV on the 115 kV bus. Raising the tap position will result in supplying a higher voltage to the 4160 volt busses.

D. Incorrect - The Reserve Station Transformers 101 Nand 101S are equipped with Load Tap Changers (LTC) capable of changing output voltage, not frequency.

. N 1-0P-33A, Sect B Technical Reference(s): N1-S0P-33A.3 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None Learning Objective: N1-262000-RBO-6 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam: Not used on the 2008 or 2009 exams Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

Facility: Nine Mile Point Unit 1 Vendor: GE Exam Date: 2010 Exam Type: RO Examination Outline Cross-reference: Level RO SRO Tier# 1 Group # 1 KIA # 295004 AK3.02 Importance Rating 2.9 Knowledge of the reasons for the following responses as they apply to PARTIAL OR COMPLETE LOSS OF D.C. POWER: Ground isolation/fault determination Proposed Question: RO Question # 46 The plant is operating at 100% power when A3-4-3, BATTERY BOARD 12 GROUND, alarms.

Which one of the following describes (1) the indication used to check the magnitude of the ground and (2) why is it necessary to quickly locate and repair the ground, in accordance with N1-0P-47A?

Ground Indication Reason for Quickly Locating and Repairing the Ground A. Ground lights in Multiple grounds could cause failure or initiation of safety Battery Room 12 equipment B. Ground lights in Presence of a ground on Battery Board 12 makes EDG Battery Room 12 103 inoperable due to degraded field flashing capability C. Ground voltmeter on Multiple grounds could cause failure or initiation of safety Control Room panel A3 equipment D. Ground voltmeter on Presence of a ground on Battery Board 12 makes EDG Control Room panel A3 103 inoperable due to degraded field flashing capability Proposed Answer: C Explanation (Optional):

A. Incorrect - Battery 14 has only local ground indication, but the ground indication for Batteries 11 and 12 is in the Control Room.

B. Incorrect - Battery 14 has only local ground indication, but the ground indication for Batteries 11 and 12 is in the Control Room. No requirement exists to declare EDG 103 inop due to solely a ground on Battery Board 12. OP-47A P&L #10 requires the EDG to be declared inop on a total loss of the battery due to insufficient field flashil1g capability on the SSC.

C. Correct - lAW Section H of N 1-0P-47A, to check the magnitude of the ground observe ground voltmeter on panel A3. lAW P & L 8.0, It is important to expedite trouble shooting and repair of a ground on the DC system because multiple grounds could cause failure or initiation of safety equipment.

D. Incorrect - No requirement exists to declare EDG 103 inop due to solely a ground on Battery Soard 12. OP-47A P&L #10 requires the EDG to be declared inop on a total loss of the battery due to insufficient field flashing capability on the SBC.

N1-0P-47A, P & L 8 & 10 and Technical Reference(s): Sect. 8.0, ARP A3-4-3, (Attach if not previously provided)

N11 01263000CO 1 Proposed References to be provided to applicants during examination: None Learning Objective: N1-263000-RBO-11 (As available)

Question Source: Sank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam: not used on the 2008 or 2009 exams Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

Facility: Nine Mile Point Unit 1 Vendor: GE Exam Date: 2010 Exam Type: RO Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 KIA # 295016 AK3.03 Importance Rating 3.5 Knowledge of the reasons for the following responses as they apply to CONTROL ROOM ABANDONMENT: Disabling control room controls Proposed Question: RO Question # 47 The Control Room has been evacuated due to a fire with the following:

  • The immediate actions of SOP-21.2, Control Room Evacuation, have been taken

o CHANNEL 11 CONTROL TRANSFER keylock switch in the EMERG position o EMERGENCY COOLING ISOLATION keylock switch in the BYPASS position Which one of the following describes the reason for these actions at RSP 11?

A. Enables manual operation of Emergency Condenser 11 from RSP 11 and disables the high steam flow isolation to ensure safe shutdown capability from outside the Control Room.

8. Disables manual operation of Emergency Condenser 11 from the Control Room to prevent a hot-short from closing the Steam Supply Valves and/or Condensate Return Valve.

C. Enables Emergency Condenser 11 automatic initiation logic located outside of the Control Room to ensure Emergency Condensers function as credited in the safety analysis.

D. Disables Emergency Condenser 11 automatic initiation logic located in the Control Room and disables the high steam flow isolation to prevent spurious initiation or isolation of the Emergency Condenser 11.

Proposed Answer: A Explanation (Optional):

A. Correct - To provide for safe shutdown of the plant in the event the Main Control Room must be evacuated, primary Emergency Cooling System controls and indication are provided at Remote Shutdown Panels. These are redundant remote manual controls at the Remote Shutdown Panels (RSP) and are enabled by placing the control transfer keylock switch in EMERG. This action does not disable the corresponding control room controls. The EMERGENCY COOLING ISOLATION keylock switch disables the high steam flow isolation (high steam flow isolation from hot shorts in control room is prevented by the confirmatory logic arrangement).

Redundant auto-initiation logics and "confirmatory" logics are provided outside the Control Room to ensure automatic initiation and continued operation of the system in the event of a Control Room fire. These logics are always in-service, and do not require manual action to lineup in the event of a fire.

B. Incorrect - Placing Channel 11 Control Transfer keylock switch in the EIVIERG position does not disable the controls from the Control Room, the Control Room controls are still operational.

C. Incorrect - Redundant auto-initiation logics and "confirmatory" logics are provided outside the Control Room to ensure automatic initiation and continued operation of the system in the event of a Control Room fire. These logics are always in-service, and do not require manual action to lineup in the event of a fire.

O. Incorrect - Redundant auto-initiation logics and "confirmatory" logics are provided outside the Control Room to ensure automatic initiation and continued operation of the system in the event of a Control Room fire. These logics are not overridden by this switch manipulation.

C-19859-C, Sht 8 N1-S0P-21.2 Technical Reference(s): (Attach if not previously provided)

SOBO-204, Emergency Cooling System Proposed References to be provided to applicants during examination: None Learning Objective: N1-296000-RBO-7 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New x Question History: Last NRC Exam: Not used on the 2008, similar to 2009 exam question #6 but different Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Comments:

Facility: Nine Mile Point Unit 1 Vendor: GE Exam Date: 2010 Exam Type: RO Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 KIA # 295031 EA1.10 Importance Rating 3.6 Ability to operate and/or monitor the following as they apply to REACTOR LOW WATER LEVEL: Control rod drive Proposed Question: RO Question # 48 A small break LOCA has resulted in the following:

  • RPV water level is 65 inches and lowering slowly
  • RPV pressure is 800 psig and lowering slowly The CRS has directed maximizing CRD system flow to the RPV using EOP-HC attachment 13.

Which one of the following describes the CRD parameter that must be monitored as RPV pressure lowers and how that parameter is to be controlled?

Monitored Parameter Controlled by Adjusting A. CRD pump motor current CRD Flow Control Valve B. CRD pump discharge pressure CRD Flow Control Valve C. CRD pump motor current CRD Drive Water Control Valve D. CRD pump discharge pressure CRD Drive Water Control Valve Proposed Answer: A Explanation (Optional):

A. Correct -lAW N1-EOP-HC attachment 13, adjust FIC 44-146B OR 28-18 as required to achieve maximum attainable flow, WHILE keeping pump amps below 245 amps. As

RPV pressure lowers the resistance to CRD flow will lower, permitting higher flows and consequently higher motor current. This must be controlled using CRD FCV in Manual or 28-18, the bypass around the FCV.

B. Incorrect - CRD Pump flow and CRD Pump motor current are the limiting parameters.

C. Incorrect - As RPV pressure lowers the resistance to CRD flow will lower permitting higher flows and consequently higher motor current. This must be controlled using CRD FCV in Manual or 28-18, the bypass around the FCV.

D. Incorrect - CRD Pump flow and CRD Pump motor current are the limiting parameters.

As RPV pressure lowers the resistance to CRD flow will lower permitting higher flows and consequently higher motor current. This must be controlled using CRD FCV in Manual or 28-18, the bypass around the FCV.

Technical Reference(s): N1-EOP-HC attachment 13 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None learning Objective: N1-201001-RBO-12 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: last NRC Exam: Not used on 2008 or 2009 exams Question Cognitive level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

Facility: Nine Mile Point Unit 1 Vendor: GE Exam Date: 2010 Exam Type: R Examination Outline Cross-reference: Level RO SRO Tier# 1 Group # 1 KIA # 295037 EA1.10 Importance Rating 3.7 Ability to operate and/or monitor the following as they apply to SCRAM CONDITION PRESENT AND REACTOR POWER ABOVE APRM DOWNSCALE OR UNKNOWN: Alternate boron injection methods: Plant-Specific Proposed Question: RO Question # 49 The plant is operating at 100% power when an ATWS results in the following:

  • Reactor power is 18%
  • Liquid Poison (LP) system 11 is initiated from the Control Room but fails to inject
  • An Operator in the field reports that 41-05, BV-LlQUID POISON TANK OUTLET is closed and cannot be opened Which one of the following methods can be used to inject boron into the RPV, in accordance with EOP-3.2, Alternate Boron Injection?

A. Align the drain of the LP tank to the suction of either LP pump.

B. Align the drain of the LP tank to the suction of the RWCU pumps.

C. Add Sodium Pentaborate to a RWCU filter and inject with the RWCU pumps.

D. Add Sodium Penta borate to a RWCU demineralizer and inject with the RWCU pumps.

Proposed Answer: C Explanation (Optional):

A. Incorrect - There is no flowpath from the LP tank drain to the suction of the LP pumps.

Also, there is no procedural guidance for connecting any type of temporary hose, such

as there is for the hydro pump.

B. Incorrect - There is no flowpath from the LP tank drain to the suction of the RWCU pumps. Also, there is no procedural guidance for connecting any type of temporary hose, such as there is for the hydro pump.

C. Correct - Per N1-EOP-3.2 Att 2, Boron Injection Using Reactor Water Cleanup System, Enriched Boron is mixed/added in the Filter Precoat Tank before being injected into RWCU filter, and then the RPV with the RWCU System pumps.

D. Incorrect - The RWCU filters are used, not the demineralizers.

. N1-EOP-3.2 Att 2 Technical Reference(s): C-18019-C (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None Learning Objective: 01-0PS-006-344-1-12, EO 1.3 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam: Not used on the 2008 or 2009 exams Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

Facility: Nine Mile Point Unit 1 Vendor: GE Exam Date: 2010 Exam Type: R Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 KIA # 295021 AA1.02 Importance Rating 3.5 Ability to operate and/or monitor the following as they apply to LOSS OF SHUTDOWN COOLING: RHR/shutdown cooling Proposed Question: RO Question # 50 The plant is shutdown with Shutdown Cooling loop 12 in service when the following occur:

  • Reactor pressure rises to 130 psig
  • Reactor water temperature rises to 355°F Which one of the following actions, if any, must be taken to start an alternate Shutdown Cooling pump?

A. Reactor water temperature must be reduced below 350°F, only.

B. The Shutdown Cooling Isolation Valves must be re-opened, only.

C. Nothing must be done; an alternate SDC pump may be started under present conditions.

D. Reactor water temperature must be reduced below 350°F AND the Shutdown Cooling Isolation Valves must be re-opened.

Proposed Answer: A Explanation (Optional):

A. Correct - Isolation Valves 38-01, SDC SYSTEM IN IV SC-11 (INSIDE) and 38-02, SDC SYSTEM OUT IV SC-12 (OUTSIDE) are interlocked so that only one valve can be

opened when Reactor pressure is above 120 psig. Below 120 psig, both valves can be opened. However, there is no close signal to these valves if pressure rises back above 120 psig. There are no other isolation signals given in the stem of the question.

The Shutdown Cooling System Isolation Valves will close on the following signals:

  • Reactor Vessel Level Low-Low (~ + 5")
  • High Area Temperature (170°F T.S. Limit)
  • Manual Isolation Since no condition caused closure of the Shutdown Cooling IVs, they do not need to be reopened. The Shutdown Cooling pumps have a 350°F start permissive, so reactor water temperature must be lowered below that value.

B. Incorrect - Since no condition caused closure of the Shutdown Cooling IVs, they do not need to be reopened.

C. Incorrect - The Shutdown Cooling pumps have a 350°F start permissive, so reactor water temperature must be lowered below that value.

D. Incorrect - Since no condition caused closure of the Shutdown Cooling IVs, they do not need to be reopened.

Technical Reference(s): N1-0P-4, P & L 3,4 and 5 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None Learning Objective: N1-205000-RBO-5 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam: Not used on the 2008 or 2009 exams Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7

55.43 Design, components, and fUnction of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Comments:

Facility: Nine Mile Point Unit 1 Vendor: GE Exam Date: 2010 Exam Type: R Examination Outline Cross-reference: Level RO SRO Tier # 1

~- ...-

Group # 1 KIA # 295003 AA2.04 Importance Rating 3.5 Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF AC. POWER: System lineups Proposed Question: RO Question # 51 The plant is operating at 100% power when the following occurs:

  • A grid disturbance results in a loss of 115 KV power
  • Breakers R10 and R40 open
  • Protective relays at Lighthouse Hill clear necessary busses
  • Bennetts Bridge auto-transfers to re-energize Line 4 Which one of the following describes the device that automatically closes to restore 115 KV power to the plant?

A Breaker R10 B. Breaker R40 C. Disconnect 178 D. Disconnect 8106 Proposed Answer: B Explanation (Optional):

A Incorrect - There is no power available on Line 1 for R-10 B. Correct - Per N1-0P-33A section B system description, R-40 closes following the disturbance

C. Incorrect - Disconnect 178 will not open unless a transformer fault is detected. With only a grid disturbance causing the loss of offsite lines, 178 will remain closed through the transient.

D. Incorrect - IF reclosure fails, bus sectionalizing disconnect SW 8106 opens AND THEN R 10 and R40 attempt another reclosure to re-energize the unfaulted section of the 115 kV bus. However in this case with Bennetts Bridge re-energizing Line 4 and no indication of bus faults, 8106 will not open.

Technical Reference(s): N1-0P-33A Sect B (Attach if not previously provided)

C-19409-C sht 1b Proposed References to be provided to applicants during examination: None Learning Objective: N1-262001-RBO-10 (As available)

Question Source: Bank # 2004 NRC #2 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: Not use on 2008 or 2009 exams Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 4 55.43 Secondary coolant and auxiliary systems that affect the facility.

Comments:

Facility: Nine Mile Point Unit 1 Vendor: GE Exam Date: 2010 Exam Type: RO Examination Outline Cross-reference: Level RO SRO Tier# 1 Group # 1 KIA # 295019 AA2.01 Importance Rating 3.5 Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF INSTRUMENT AIR: Instrument air system pressure Proposed Question: RO Question # 52 The plant is operating at 100% power with the following:

  • An Instrument Air (IA) line break results in IA pressure slowly and continuously lowering
  • The following sequence of events occurs:

Time (mm:ss) Event 00:10 Instrument Air Compressor (lAC) 13 loads 00:20 lAC 11 loads 00:30 lAC 12 starts and loads 00:45 Annunciator L 1-4-7, INST AIR BACK - UP VALVE OPEN alarms 00:57 Breathing Air Blocking Valve 114-02 closes; Automatic Dryer Bypass valves94-164 and 94-206 open 02:30 Annunciator F3-3-2, CRD CONTROL AIR PRESSURE HI-LO alarms

  • IA pressure has still been continuously lowering throughout this transient Which one of the following describes Instrument Air pressure at time 01 :OO?

At time 01 :00, Instrument Air pressure is between ...

A. 93 and 98 psig

B. 90 and 93 psig C. 85 and 90 psig D. 65 and 85 psig Proposed Answer: D Explanation (Optional):

A. Incorrect - IA pressure is below 85 psig (the setpoint for the Breathing Air Blocking Valve 114-02 closure and the Automatic Dryer Bypass valves94-164 and 94-206 opening). IA pressure is above 65 psig (Annunciator F3-3-2, CRD CONTROL AIR PRESSURE HI-LO setpoint).

B. Incorrect - IA pressure is below 85 psig (the setpoint for the Breathing Air Blocking Valve 114-02 closure and the Automatic Dryer Bypass valves94-164 and 94-206 opening). IA pressure is above 65 psig (Annunciator F3-3-2, CRD CONTROL AIR PRESSURE HI-LO setpoint).

C. Incorrect - IA pressure is below 85 psig (the setpoint for the Breathing Air Blocking Valve 114-02 closure and the Automatic Dryer Bypass valves94-164 and 94-206 opening). IA pressure is above 65 psig (Annunciator F3-3-2, CRD CONTROL AIR PRESSURE HI-LO setpoint).

D. Correct - Annunciator L 1-4-7, INST AIR BACK - UP VALVE OPEN alarms at 90 psig.

Breathing Air Blocking Valve 114-02 closes and the Automatic Dryer Bypass valves 94 164 and 94-206 open at 85 psig. Annunciator F3-3-2, CRD CONTROL AIR PRESSURE HI-LO alarms at low pressure of 65 psig. This means at time 01 :00, pressure is between 65 and 85 psig.

N1-0P-20, Sect. B Technical Reference(s): N1-ARP-L 1-4-7 (Attach if not previously provided)

N 1-ARP-F3-3-2 Proposed References to be provided to applicants during examination: None Learning Objective: N1-278001-RBO-5 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam: Not used on 2008 or 2009 exams Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 4 55.43 Secondary coolant and auxiliary systems that affect the facility.

Comments:

Facility: Nine Mile Point Unit 1 Vendor: GE Exam Date: 2010 Exam Type: RO Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 KIA # 295026 EA2.03 Importance Rating 3.9 Ability to determine and/or interpret the following as they apply to SUPPRESSION POOL HIGH WATER TEMPERATURE: Reactor pressure Proposed Question: RO Question # 53 The plant was operating at 100% power when an ATWS resulted in the following:

  • Reactor power is 3% and stable
  • Reactor pressure is 965 psig controlled using ERVs
  • Torus water temperature is 109°F and rising Which one of the following describes the required action and the basis for this action?

Action Required Basis A. Trip all Recirc Pumps Prevent exceeding PSP B. Start Liquid Poison injection Prevent exceeding PSP C. Trip all Recirc Pumps Prevent exceeding HCTL D. Start Liquid Poison injection Prevent exceeding HCTL Proposed Answer: D Explanation (Optional):

A. Incorrect - Recirc pumps are only tripped if power is above 6%. Torus water level and torus pressure are not applicable factors for this event, therefore there is no immediate challenge to PSP.

B. Incorrect - Torus water level and torus pressure are not applicable factors for this event, therefore there is no immediate challenge to PSP.

C. Incorrect - Recirc pumps are only tripped if power is above 6%.

D. Correct - lAW EOP-3, Before torus temperature reaches 11 O°F start LP injection.

A scram failure with MSIV closure results in torus heatup due to steam discharged through the ERVs. If torus temperature cannot be maintained below the Heat Capacity Temperature Limit, EOP-4 will require a blowdown. Based on the given pressure and pressure control mechanism, it can be determined that continued torus heatup will occur. Since it is desirable to shut down the reactor before the blowdown is required, boron is injected when the torus heatup begins. If power is relatively low (less than approximately 3%), boron injection may be completed and the reactor shut down before torus temperature reaches the Heat Capacity Temperature Limit.

Technical Reference(s): EOP-3, Bases, pg 157 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None Learning Objective: 01-0PS-006-344-1-03 EO-1.2 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam: Not used on the 2008 or 2009 exams.

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

Facility: Nine Mile Point Unit 1 Vendor: GE Exam Date: 2010 Exam Type: RO Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 KlA# 295023 2.4.21 Importance Rating 4.0 Emergency Procedures I Plan: Knowledge of the parameters and logic used to assess the status of safety functions, such as reactivity control, core cooling and heat removal, reactor coolant system integrity, containment conditions, radioactivity release control, etc. (Refueling Accidents)

Proposed Question: RO Question # 54 During refueling, the following conditions exist:

  • An irradiated LPRM string has been inadvertently raised partially out of the water
  • The Fuel Pool High Range Refuel/Bypass Switch is in the REFUEL position
  • The Fuel Pool High Range Radiation Monitor reads 1500 mr/hr Which one of the following describes the status of Reactor Building Ventilation?

A. One Reactor Building Emergency Ventilation fan started, Reactor Building Supply and Exhaust Fans are operating in a normal lineup.

B. One Reactor Building Emergency Ventilation fan started, Reactor Building Supply and Exhaust Fans tripped and the Reactor Building isolated.

C. Both Reactor Building Emergency Ventilation fans started, Reactor Building Supply and Exhaust Fans are operating in a normal lineup.

D. Both Reactor Building Emergency Ventilation fans started, Reactor Building Supply and Exhaust Fans tripped and the Reactor Building isolated.

Proposed Answer: D Explanation (Optional):

A. Incorrect - Both RBEVS fans start and the RB ventilation system trips and the Reactor Building isolates.

B. Incorrect - Both RBEVS fans start.

C. Incorrect - The Reactor Building isolated.

D. Correct - The Reactor Building Emergency Ventilation system will auto-initiate on Fuel Pool High Range Rad Monitor Hi at 1000 mr/hr with Refuel/Bypass Switch in REFUEL position.

When an automatic initiation signal is received, the RBEV fans 11 and 12 both start, the following valves open: 202-36, 202-37, 202-34,202-38,202-35 and the cooling valves, 202-74 and 202-75 close, placing the RBEVS system in service. At the same time trip signals are sent to both RB supply and exhaust fans and close signals are sent to RB isolation valves. This removes the RB ventilation system from service and isolates it from the Reactor Building.

Technical Reference(s): N1-0P-10, Sect. B.3.0 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None Learning Objective: N1-261000 -RBO-10 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam: Not used on 2008 or 2009 exam.

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Comments:

Facility: Nine Mile Point Unit 1 Vendor: GE Exam Date: 2010 Exam Type: R Examination Outline Cross-reference: level RO SRO Tier # 1 Group # 1 KIA # 295018 2.4.46 Importance Rating 4.2 Emergency Procedures I Plan: Ability to verify that the alarms are consistent with the plant conditions (Partial or Total loss of CCW).

Proposed Question: RO Question # 55 The plant is operating at 100% power with the following conditions:

  • An Operator reports from Turbine Building elevation 351' that the Reactor Building Closed loop Cooling (RBClC) high point vent line is discharging a steady stream of water Which one of the following is a possible cause of the problem?

Tube leak in the ...

A. Reactor Recirculation Pump seal cooler B. EDG 102 Jacket Water Heat Exchanger C. Reactor Building equipment drain tank cooler D. Reactor Water Cleanup regenerative heat exchanger Proposed Answer: A Explanation (Optional):

A. Correct - System pressure must be higher than the RBClC pressure at the location of the leak. The recirc pump seal cooler is the only leak identified above that has ample

pressure to cause in-leakage to the RBClC system and cause leakage at the elevated drain in the turbine building (RBClC cannot flow into the Makeup Tank because there are check valves on lines into RBClC).

B. Incorrect - Cooled by EDG Raw Water, not RBClC C. Incorrect - This would leak RBClC water into the drain tank D. Incorrect - The RHX is cooled by RWCU flow returning to the Reactor. Only the NRHX is cooled by RBClC. A tube leak in the RHX would cause water entering RWCU to leak into the RWCU discharge and short-cycle the system.

Technical Reference(s): N1-0P-11. Sect. H.16.0 (Attach if not previously provided)

C-18022-C sht 2 Proposed References to be provided to applicants during examination: None learning Objective: N1-208000-RBO-11 (As available)

Question Source: Bank #

ID: N1-208000-RBO-Modified Bank # (Note changes or attach parent) 11-Q-02 New Question History: last NRC Exam: Not used on 2008 or 2009 exams Question Cognitive level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 4 55.43 Secondary coolant and auxiliary systems that affect the facility.

Comments:

Facility: Nine Mile Point Unit 1 Vendor: GE Exam Date: 2010 Exam Type: RO Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 KIA # 295038 2.1.27 Importance Rating 3.9 Conduct of Operations: Knowledge of system purpose and I or function (High Off-site Release Rate).

Proposed Question: RO Question # 56 The plant is operating at 100% power with the following:

  • The running Turbine Building Ventilation fans have tripped
  • The Shift Manager declares an ALERT condition based on off-site release rates Which one of the following operator actions is required in accordance with EOP-6, Radioactivity Release Control, and why?

A. Restart the Turbine Building Ventilation system to direct any radioactivity release through an elevated, monitored path.

B. Verify the Turbine Building Ventilation system isolated to minimize the radiological release from the Turbine Building.

C. Restart the Turbine Building Ventilation system to prevent transferring air between the Reactor Building and the Turbine Building.

D. Verify the Turbine Building Ventilation system isolated to prevent transferring air between the Reactor Building and the Turbine Building.

Proposed Answer: A

Explanation (Optional):

A. Correct - The basis document for N1-EOP-6 states the turbine building ventilation is restarted to prevent an unmonitored ground release. N1-EOP-6 is entered when the ALERT condition based on oft-site release rates is exceeded. The Turbine Building ventilation system maintains a negative pressure in the Turbine Building to ensure releases from or through systems that pass through secondary containment are captured for release through the plant stack.

B. Incorrect because the turbine building is not air tight, not restarting the turbine building ventilation would result in an unmonitored, ground release of radioactivity.

C. Incorrect because starting the turbine building ventilation would not prevent air from transferring to the reactor building.

D. Incorrect because isolating the turbine building ventilation would not prevent air from transferring to the reactor building.

Technical Reference(s): EOP-Bases for EOP-6 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None Learning Objective: N1-288002-RBO-12 (As available)

Question Source: Bank # ID: NRC 2006 RO 67 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: Not used on 2008, similar to 2009 #18 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

Facility: Nine Mile Point Unit 1 Vendor: GE Exam Date: 2010 Exam Type: R Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 KIA # 600000 AA 1.09 Importance Rating 2.5 Ability to operate and I or monitor the following as they apply to PLANT FIRE ON SITE: Plant fire zone panel (including detector location)

Proposed Question: RO Question # 57 The plant is operating at 100% power with the following:

  • The Fire Brigade Leader reports a fire at the Recirculation MG Sets
  • The corresponding Fire Zone, C-2092MG, is NOT yet in alarm
  • This Fire Zone is in ALARM ONLY
  • The Local Fire Panel is NOT accessible
  • The Fire Brigade Leader has requested fire suppression be initiated from the Control Room
  • The Fire Brigade Leader reports that the applicable fire suppression source has been verified lined up to the header Which one of the following describes the method of suppression for this Fire Zone AND the required operator actions to initiate suppression at Main Fire Panel 2 (MFP2), in accordance with OP-21 E, Fire Protection System Fire Detection?

Method of Fire Suppression Operator Actions at MFP2 A. Place Zone Control Switch in AUTO B. Water Place Zone Control Switch in AUTO C. Place Zone Control Switch in DISCHARGE D. Water Place Zone Control Switch in DISCHARGE Proposed Answer: C

Explanation (Optional):

A. Incorrect - lAW OP-21 E section H.3.0, for manual initiation of an automatic fire suppression system, place the control switch in DISCHARGE. Since the Fire Zone is not yet in alarm, the AUTO position will not result in suppression.

B. Incorrect - C stands for C02. lAW OP-21 E section H.3.0, for manual initiation of an automatic fire suppression system, place the control switch in DISCHARGE. Since the Fire Zone is not yet in alarm, the AUTO position will not result in suppression.

C. Correct - C stands for C02. lAW OP-21 E section H.3.0, for manual initiation of an automatic fire suppression system, place the control switch in DISCHARGE.

D. Incorrect - C stands for C02.

Technical Reference(s): ~11CO:e~~;~;; ~tt 3, H.3.0, N1-0P- (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None Learning Objective: N 1-286000-RBO-5 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam: Not used on 2008 or 2009 exams.

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Comments:

Facility: Nine Mile Point Unit 1 Vendor: GE Exam Date: 2010 Exam Type: R Examination Outline Cross-reference: Level RO SRO Tier# 1 Group # 1 KIA # 295001 AA2.06 Importance Rating 3.2 Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF FORCED CORE FLOW CIRCULATION: Nuclear boiler instrumentation Proposed Question: RO Question # 58 The plant is operating at 100% power with the following:

Time (hh:mm) Condition 08:00 All Reactor Recirc Pumps (RRPs) are running Total Core Flow is 62 Mlbm/hr 08:05 RRP 11 trips 08:07 Individual RRP flows stabilize at the following indicated values:

RRP 11: 5 Mlbm/hr RRP 12: 14 Mlbm/hr RRP 13: 14 Mlbm/hr RRP 14: 14 Mlbm/hr RRP 15: 14 Mlbm/hr Which one of the following is the approximate value for actual core flow?

A. 61 Mlbm/hr B. 56 Mlbm/hr C. 51 Mlbm/hr D. 46 Mlbm/hr

Proposed Answer: C Explanation (Optional):

A. Incorrect - This would be the total of all pump flows, assuming the flow for the tripped pump is still forward flow. Reverse flow occurs in this situation and the summing network assumes all flow is in the forward direction.

B. Incorrect - This value incorrectly subtracts only the 5 Mlbm/hr that is indicated for RRP 11 from the total indicated core flow.

C. Correct - RRP 11 indicated flow is actually reverse flow through the idle loop. You can determine actual core flow 2 ways:

1. Double the indicated flow from RRP 11, 5x2=10 Mlbm/hr. This would then be subtracted from total indicated core flow (4*14+5 = 61). 61-10= 51 Mlbm/hr.
2. Since the 5 Mlbm/hr flow from RRP 11 is actually reverse flow, it is being supplied from the running 4 pumps, bypassing the core. Therefore the 5 Mlbm/hr reverse flow in the tripped loop must be subtracted from the total flow of the 4 pumps that are still running. 14*4-5= 51 Mlbm/hr.

D. Incorrect - This would be the total of all pump flows without adding the flow in the tripped Recirc loop, minus double the tripped Recirc loop flow (4*14-10=46).

Technical Reference(s): OP-1 P & L 21 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None Learning Objective: N 1-20200 1-RBO-11 (As available)

Question Source: Bank # NRC 2006 RO 49 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: Not used on 2008 or 2009 exams Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x

10 CFR Part 55 Content: 55.41 2 55.43 General Design features of the core, including core structure, fuel elements, control rods, core instrumentation, and coolant flow.

Comments:

Facility: Nine Mile Point Unit 1 Vendor: GE Exam Date: 2010 Exam Type: RO Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 2 KIA # 295017 AK1.02 Importance Rating 3.8 Knowledge of the operational implications of the following concepts as they apply to HIGH OFF-SITE RELEASE RATE: Protection of the general public Proposed Question: RO Question # 59 An un-isolable primary system rupture has occurred, causing rising Offsite release rates.

Which one of the following lists the condition that requires entry into EOP-6, Radioactivity Release Control, AND the condition that requires entry into EOP-8, RPV Blowdown, to ensure protection of the general public?

Entry into EOP-6, Radioactivity Release Control, is required when Offsite release rate exceeds the (1) Emergency Action Level value. Entry into EOP-8, RPV Blowdown, is required before Offsite release rate exceeds the (2) Emergency Action Level value.

(1) (2)

A. Unusual Event Site Area Emergency B. Unusual Event General Emergency C. Alert Site Area Emergency D. Alert General Emergency Proposed Answer: D Explanation (Optional):

A. Incorrect - Entry into EOP-6 is required at the Alert level. Entry into EOP-8 is required before the General Emergency level.

B. Incorrect - Entry into EOP-6 is required at the Alert level.

C. Incorrect - Entry into EOP-8 is required before the General Emergency level.

D. Correct - Entry into EOP-6 is required at the Alert level. Entry into EOP-8 is required before the General Emergency level.

Technical Reference(s): EOP-6, EOP Bases (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None Learning Objective: 01-0PS-006-344-1-06, EO 1.3 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam: Not used on the 2008 or 2009 exams Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

Facility: Nine Mile Point Unit 1 Vendor: GE Exam Date: 2010 Exam Type: RO Examination Outline Cross-reference: Level RO SRO Tier# 1 Group # 2 KIA # 295009 AK2.01 Importance Rating 3.9 Knowledge of the interrelations between LOW REACTOR WATER LEVEL and the following:

Reactor water level indication Proposed Question: RO Question # 60 A LOCA has resulted in the following:

  • Actual Reactor water level is zero (0) inches
  • Fuel Zone water level instruments indicate zero (0) inches 3
  • Core Spray Loop 11 is injecting at 190 x 10 Ibmlhr 3
  • Core Spray Loop 12 is injecting at 190 x 10 Ibmlhr
  • Reactor pressure is 50 psig and stable
  • Drywell pressure is 8 psig and slowly lowering
  • Drywell temperature elevation 319' is 302°F and slowly lowering
  • Drywell temperature elevation 263' is 275°F and slowly lowering
  • Drywell temperature elevation 230' is 250°F and slowly lowering
  • Drywell average temperature is 275°F and slowly lowering In addition to the Fuel Zone water level instruments, which one of the following level instruments is available, if any?

A. Lo-Lo-Lo instruments B. Wide Range instruments C. HilLo - LolLo Rosemount instruments D. No other instruments can be used Proposed Answer: D

Explanation (Optional):

A. Incorrect - Lo-Lo-Lo instruments are above the minimum usable level in Detail A, but with both Core Spray loops injecting, they will be indicating upscale due to the location of the variable leg tap.

B. Incorrect - Wide Range instruments are below the minimum usable level and cannot be used.

C. Incorrect - Rosemount instruments are below the minimum usable level but could be inferred as at or above the minimum usable level if the incorrect temperature is used.

D. Correct - Wide Range and Rosemount instruments are below the minimum usable levels in EOP Detail A for the given DWelevation 319' temperature. Lo-Lo-Lo instruments cannot be used because both Core Spray loops are injecting.

Level Instruments Restrictions Technical Reference(s): EOP Detail A, C-35843-C (Attach if not previously provided)

Level Instruments Proposed References to be provided to applicants during examination: Restrictions, EOP Detail A Learning Objective: N1-216000-RBO-12 (As available)

Question Source: Bank # EOP Bank#10 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: Not used on the 2008 or 2009 exams Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

Facility: Nine Mile Point Unit 1 Vendor: GE Exam Date: 2010 Exam Type: RO Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 2 KIA # 295032 EK3.01 Importance Rating 3.5 Knowledge of the reasons for the following responses as they apply to HIGH SECONDARY CONTAI NMENT AREA TEMPERATU RE: Emergency/normal depressurization Proposed Question: RO Question # 61 An un-isolable primary system rupture has resulted in two Reactor Building General Area temperatures exceeding the Maximum Safe Value.

Which one of the following is a reason EOP-5, Secondary Containment Control, directs an RPV Blowdown on high Secondary Containment temperature?

A. Prevent a failure of the Primary Containment B. Preserve operability of Electromatic Relief Valves C. Ensure site boundary dose limits are not exceeded D. Minimize damage to equipment required for safe shutdown Proposed Answer: D Explanation (Optional):

A. Incorrect - The high temperatures pose a direct and immediate threat to secondary containment, equipment in the secondary containment, and safe operation of the plant.

Not the primary containment, which is protected by EOP-4.

B. Incorrect - ERV operability is the reason for RPV blowdown at 300°F in the Drywell in EOP-4. However, the component of concern is located in the Drywell, and not directly affected by high temperatures in the Reactor Building.

C. incorrect - Plausible because this is the reason for depressurizing in EOP-6, Rad Release.

D. Correct - A parameter above the Maximum Safe Value in two separate areas is indicative of a wide-spread problem posing a direct and immediate threat to secondary containment, equipment in the secondary containment, and safe operation of the plant.

if a primary system is discharging into the secondary containment and area temperatures, radiation levels, or water levels are above their Maximum Safe Values in two or more areas, a blowdown must therefore be performed in accordance with EOP-8.

The blowdown minimizes flow through the break, rejects heat to the suppression pool in preference to outside the containment, and places the primary system in the lowest possible energy state.

Technical Reference(s): EOP Bases, EOP-5 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None Learning Objective: N 1-290001-RBO-12 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam: Not used on the 2008 or 2009 exams Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

Facility: Nine Mile Point unit 1 Vendor: GE Exam Date: 2010 Exam Type: RO Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 2 KJA# 295036 EA 1.04 Importance Rating 3.1 Ability to operate and/or monitor the following as they apply to SECONDARY CONTAINMENT HIGH SUMP/AREA WATER LEVEL: Radiation monitoring: Plant-Specific Proposed Question: RO Question # 62 The plant is operating at 80% power with the following:

Time (hh:mm) Condition 08:25 Annunciator H2-2-1, R BLDG FL DR SUMPS 11-16 AREA WTR LVL LEVEL HIGH, alarms Computer point B130, RBFDT 12 (NE) LVL HIGH, is in alarm 08:30 Computer point F188, NE RB CORNER RM WTR LVL HIGH, is in alarm 08:35 Annunciator H1-4-8, AREA RADIATION MONITORS, alarms A computer printout confirms RX BLDG - NE CORNER, EL 198 is at 11 mr/hr Which one of the following describes the significance of the high radiation alarm?

A. This is an indication of a second General Area in alarm, only.

B. This is an additional entry condition for EOP-5, Secondary Containment Control, only.

C. This confirms that one General Area is above a Maximum Safe Value, requiring a shutdown per OP-43C.

D. This is indication of one General Area above two Maximum Safe Values, requiring a shutdown per OP-43C.

Proposed Answer: B Explanation (Optional):

A. Incorrect - The alarm for F188 is actuated at a water level of S feet (Table S N1-EOP-S) in the NE corner room, which is the maximum safe value. This is the same area as the radiation monitor (RB198 NE RB EDT AREA).

B. Correct - The alarm for F188 is actuated at a water level of S feet (Table S N1-EOP-S) in the NE corner room, which is the maximum safe value. This is the same area as the radiation monitor (RB198 NE RB EDT AREA). The radiation monitor is a second entry condition for EOP-S. The radiation levels (11 mr/hr) are well below the Max Safe value of 8 R/hr.

C. Incorrect - There is only one area above the maximum safe value for water level. The radiation alarm is a second entry condition for EOP-S. The radiation levels (11 mr/hr) are well below the Max Safe value of 8 R/hr. EOP-S directs waiting until 2 or more General areas are above Maximum Safe Values of the same parameter before shutdown is required.

D. Incorrect - There is only one area above the maximum safe value for water level. The radiation alarm is a second entry condition for EOP-S. The radiation levels (11 mr/hr) are well below the Max Safe value of 8 Rlhr. EOP-S directs waiting until 2 or more General areas are above Maximum Safe Values of the same parameter before shutdown is required.

EOP-S Technical Reference{s): ARP-H1, H2 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: EOP-S ONLY Learning Objective: 01-0PS-006-344-1-0S EO-1.3 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam: Not used on 2008 or 2009 exams Question Cognitive Level: Memory or Fundamental Knowledge

Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

Facility: Nine Mile Point Unit 1 Vendor: GE Exam Date: 2010 Exam Type: RO Examination Outline Cross-reference: Level RO SRO Tier# 1 Group # 2 KIA # 295012 AA2.01 Importance Rating 3.8 Ability to determine and/or interpret the following as they apply to HIGH DRYWELL TEMPERATURE: Drywell temperature Proposed Question: RO Question # 63 A small steam leak in the Drywell has resulted in the following:

  • Drywell average temperature is 210°F and rising slowly
  • Drywell pressure is 5 psig and rising slowly
  • Containment parameters are in the NO SPRAY area of the Containment Spray Initiation Limit
  • Torus water temperature is 78°F and stable
  • Reactor pressure is 830 psig and slowly lowering Which one of the following actions is required for these conditions?

A. Start all available Drywell Cooling fans.

B. Enter N1-EOP-8 and blowdown the Reactor.

C. Enter N1-EOP-4.1 and vent the Containment.

D. Initiate Containment Spray in Torus Cooling mode.

Proposed Answer: A Explanation (Optional):

A. Correct - lAW EOP-4, drywell cooling is required to attempt to maintain drywell temperature less than 300°F.

B. Incorrect - There is no requirement to blowdown based on the conditions above.

Temperature may be lowered or drywell pressure may rise allowing containment spray prior to reaching a requirement for a blowdown.

C. Incorrect - EOP-4.1 entry is only directed from EOP-4 after Torus pressure has exceeded 13 psig and if PCPL is being challenged. In this case, with Orywell pressure at 5 psig and riSing slowly, neither of these conditions is met.

O. Incorrect - The requirement to initiate torus cooling is 85°F.

Technical Reference(s): EOP-4, EOP-1 att 16 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None Learning Objective: N1-223001-RBO-12 (As available)

Question Source: Bank # 10: AUO 2008 RO 61 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: Not used on 2008 or 2009 exams Same KIA as 2009 #14 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

Facility: Nine Mile Point Unit 1 Vendor: GE Exam Date: 2010 Exam Type: RO Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 2 KIA # 295029 2.4.1 Importance Rating 4.6 Emergency Procedures I Plan: Knowledge of EOP entry conditions and immediate action steps (High Suppression Pool Water Level).

Proposed Question: RO Question # 64 A plant startup was in progress when a Feedwater line break in the Drywell resulted in the following:

  • The Reactor scrammed on low RPV water level
  • RPV water level reached a low of 2 inches
  • RPV water, level has been restored to 72 inches
  • RPV pressure is 150 psig and lowering slowly
  • Drywell pressure is 2.5 psig and rising slowly
  • Drywell average temperature is 145°F and rising slowly
  • Torus water temperature is 81°F and rising slowly

Enter EOP-4 on high Torus water (1) . Place all Containment Spray pump control switches in pull-to-Iock (2)

(1 ) (2)

A. Level Immediately B. Level Only if Drywell pressure exceeds 3.5 psig C. Temperature Immediately D. Temperature Only if Drywell pressure exceeds 3.5 psig

Proposed Answer: A Explanation (Optional):

A. Correct - Torus water level is above the 11.25' EOP-4 entry condition. Containment Spray pump control switches are placed in pull-to-Iock immediately upon entry into EOP-4, as long as the pumps are not already spraying the Containment. The given conditions would not have provided an automatic start signal for Containment Spray, because Drywell pressure has not yet reached 3.5 psig.

B. Incorrect - Containment Spray pump control switches are placed in pull-to-Iock immediately upon entry into EOP-4, as long as the pumps are not already spraying the Containment. The given conditions would not have provided an automatic start signal for Containment Spray, because Drywell pressure has not yet reached 3.5 psig.

C. Incorrect - Torus water temperature is NOT above the 85°F EOP-4 entry condition.

D. Incorrect - Torus water temperature is NOT above the 85°F EOP-4 entry condition.

Technical Reference(s): EOP-4 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None Learning Objective: 01-0PS-006-344-1-04, EO-1.3 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam: Not used on 2008,2009 Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

Facility: Nine Mile Point Unit 1 Vendor: GE Exam Date: 2010 Exam Type: RO Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 2 KJA# 295002 AA2.01 Importance Rating 2.9 Ability to determine and/or interpret the following as they apply to LOSS OF MAIN CONDENSER VACUUM: Condenser vacuum/absolute pressure Proposed Question: RO Question # 65 SOP-25.1, Unplanned Loss Of Condenser Vacuum, contains the following decision step:

No Yes Which one of the following describes the significance of the listed annunciator?

If F3-4-6, First Stage Bowl Press, is in alarm ...

A. The turbine will NOT trip on low condenser vacuum.

B. A turbine trip will directly cause an automatic Reactor scram.

C. The Reactor must be manually scrammed before vacuum falls to 22.1" Hgv.

D. Condenser vacuum can approach 10" Hgv. before the Reactor must be manually scrammed.

Proposed Answer: D Explanation (Optional):

A. Incorrect - Low first stage bowl pressure bypasses the reactor scram on a turbine trip.

The turbine trip is not bypassed.

B. Incorrect - A turbine trip will cause a reactor scram when turbine first stage pressure is above 310 psig (-45% power). This annunciator in alarm indicates that turbine first stage pressure is below 310 psig therefore no Reactor scram will occur.

C. Incorrect - With the alarm OFF the reactor is scrammed prior to 22.1" Hgv.

D. Correct - With the alarm ON a turbine trip will NOT cause a reactor scram. Therefore the reactor scram can be delayed until condenser vacuum approaches 10" Hgv.

N 1-S0P-25.1 Technical Reference(s): N1-ARP-F3-4-6 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None Learning Objective: 01-0PS-006-342-1-01, EO-1.3 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam: Not used on the 2008 or 2009 exams Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or AnalysiS 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

Facility: Nine Mile Point Unit 1 Vendor: GE Exam Date: 2010 Exam Type: RO Examination Outline Cross-reference: Level RO SRO Tier # 3 Group # 1 KIA # G1 2.1.34 Importance Rating 2.7 Conduct of Operations: Knowledge of primary and secondary plant chemistry limits.

Proposed Question: RO Question # 66 The plant is operating at 30% power with the following:

  • A2-1-1, GENERATOR STAT. WATER OUTLET TEMP. HIGH, is in alarm
  • A2-1-2, GENERATOR STAT. WATER INLET PRESS. LOW, is in alarm
  • A2-2-1, GENERATOR STAT. WATER HIGH CONDUCTIVITY, is in alarm Which one of the following warrants an immediate trip of the Turbine?

A. Stator water inlet flow at 400 gpm B. Stator water inlet pressure at 22 psig C. Stator water outlet temperature at 87°C D. Stator water conductivity at 10.2 ""mho/cm Proposed Answer: D Explanation (Optional):

A. Incorrect - Although this flow is low (runback at 442 gpm), it indicates that there has not been a complete loss of SWC flow. With power at 30%, generator amps are below the value at which a runback would occur. SOP-32 does not require a Turbine trip under these conditions.

B. Incorrect - Although this pressure is below the alarm setpoint of 23 psig, it indicates that there has not been a complete loss of SWC flow. With power at 30%, generator amps are below the value at which a runback would occur. SOP-32 does not require a Turbine trip under these conditions.

C. Incorrect - This temperature is below the generator runback temperature (95°C). It indicates that there has not been a complete loss of SWC flow. With power at 30%,

generator amps are below the value at which a runback would occur. SOP-32 does not require a Turbine trip under these conditions.

O. Correct - Per ARP A2-2-1, conductivity above 9.9 requires a turbine trip.

ARP A2-2-1, A2-2-2 Technical Reference(s): OP-44, Sect. B, Controls (Attach if not previously provided)

SOP-32 Proposed References to be provided to applicants during examination: None Learning Objective: N1-253000-RBO-12 (As available)

Question Source: Bank # Turb Sys 1052712 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: Not used on the 2008 or 2009 exams Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 5 55.43 Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.

Comments:

Facility: Nine Mile Point Unit 1 Vendor: GE Exam Date: 2010 Exam Type: RO Examination Outline Cross-reference: Level RO SRO Tier # 3 Group # 1 KIA #

_G1 __ -2.1.8 Importance Rating 3.4 Conduct of Operations: Ability to coordinate personnel activities outside the control room.

Proposed Question: RO Question # 67 Following an evacuation of the Control Room, you direct a Plant Operator to take local manual control of Feedwater Flow Control Valve (FCV) 11 using SOP-21.2, Control Room Evacuation, , Feedwater Pumps 11112 FCV Manual Operation.

Which one of the following actions must be taken to place Feedwater FCV 11 in local manual?

A. De-energize Feedwater FCV 11 controller.

B. Place the pneumatic controller at the FCV in manual.

C. Pin the Feedwater FCV 11 and vent the valve operator.

D. Remove the HPCI Fuses, FU8 and FU9, in Auxiliary Control Cabinet 1S34.

Proposed Answer: C Explanation (Optional):

A. Incorrect - De-energizing the controller will cause the FCV to lock up in the AS-IS condition.

B. Incorrect - There are no manual methods for operating the FCV using the air cylinder.

C. Correct - lAW SOP-21.2, Take manual control of 11 FCV as follows:

  • At FW Pump 11 FCV, align upper hole in manual control collar to hole in Feedwater FCV stem by rotating manual control handwheel.
  • Insert control pin into manual control collar hole.
  • Close operator air supply valve
  • Uncap AND open operator top AND bottom piston vents:
  • WHEN directed by CRO, throttle rotating manual control handwheel D. Incorrect - This will allow throttling of HPCI from the Control Room, not local control.

Technical Reference(s): SOP-21.2, Att 6 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None Learning Objective: N1-259001-RBO-11 (As available)

Question Source: Bank # SYSID 36053 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: Not used on 2008 or 2009 exams Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

Facility: Nine Mile Point Unit 1 Vendor: GE Exam Date: 2010 Exam Type: RO Examination Outline Cross-reference: Level RO SRO Tier# 3 Group # 2 KIA # G2 2.2.13 Importance Rating 4.1 Equipment Control: Knowledge of tagging and clearance procedures.

Proposed Question: RO Question # 68 Which one of the following is permitted under an Operating Permit tag, in accordance with CNG-OP-1.01-1007, Clearance and Safety Tagging?

A. Intermittent operation of tagged equipment.

B. Use of non-site personnel as Tagout Holders.

C. Manipulation of components while tagged with a Danger tag.

D. Unrestricted operation of tagged equipment over multiple shifts.

Proposed Answer: A Explanation (Optional):

A. Correct - Operating Permit tags shall be attached to equipment or switches, or controls where testing or for intermittent personnel protection B. Incorrect - Non-site personnel shall not sign onto a Tagout that has an Operating Permit Tag as a Tagout Holder.

C. Incorrect - A device tagged with an Operating Permit tag shall not be tagged with a Danger tag at the same time.

D. Incorrect - Multiple restrictions apply to Operating Permit tags such as band c above plus others listed in Attachment 14 of CNG CNG-OP-1.01-1007. In particular, operation of equipment under an Operating Permit tag over multiple shifts requires re-notification

of the Tagging Authority each shift.

Technical Reference(s): CNG-OP-1.01-1007, Sects. 5.4 (Attach if not previously provided) and Att 14 Proposed References to be provided to applicants during examination: None Learning Objective: GAP-OP-02-T001 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam: Not used on the 2008 or 2009 exams Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

Facility: Nine Mile Point Unit 1 Vendor: GE Exam Date: 2010 Exam Type: R Examination Outline Cross-reference: Level RO SRO Tier # 3 Group # 2 KIA # G2 2.2.20 Importance Rating 2.6 Equipment Control: Knowledge of the process for managing troubleshooting activities.

Proposed Question: RO Question # 69 The plant is operating at 100% power with the following:

  • Area Radiation Monitor (ARM) #1, SE Plant Entrance TB 261', has an intermittent fault and is periodically alarming upscale.
  • A troubleshooting plan has been developed that contains two phases.
  • Phase 1 will cause multiple nuisance alarms on annunciator H1-4-8, Area Radiation Monitors.

Which one of the following describes the flagging tool color used to mark annunciator H1-4-8 during each phase of this troubleshooting plan, in accordance with CNG-OP-1.01-2003, Alarm Response and Control?

Phase 1 Phase 2 A. Black Black B. Black Yellow C. Yellow Black D. Yellow Yellow Proposed Answer: B Explanation (Optional):

A. Incorrect - During phase 2, the annunciator will have one of multiple inputs removed from service. This requires a yellow flag per CNG-OP-1.01-2003 section 5.2.0.1.

B. Correct - During phase 1, the annunciator must be flagged for maintenance/nuisance.

This requires a black flag per CNG-OP-1.01-2003 section 5.2.D.1. During phase 2, the annunciator will have one of multiple inputs removed from service. This requires a yellow flag per CNG-OP-1.01-2003 section 5.2.D.1.

C. Incorrect - During phase 1, the annunciator must be flagged for maintenance/nuisance.

This requires a black flag per CNG-OP-1.01-2003 section 5.2.D.1. During phase 2, the annunciator will have one of multiple inputs removed from service. This requires a yellow flag per CNG-OP-1.01-2003 section 5.2.D.1.

D. Incorrect - During phase 1, the annunciator must be flagged for maintenance/nuisance.

This requires a black flag per CNG-OP-1.01-2003 section 5.2. D.1.

CNG-OP-1.01-2003 Technical Reference(s): N1-0P-50A (Attach if not previously provided)

N1-ARP-H1 Proposed References to be provided to applicants during examination: None N 1-272000-RBO-5, S-ODP-OPS-Learning Objective: (As available) 0001-T001 Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam: Not used on the 2008 or 2009 exams Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

Facility: Nine Mile Point Unit 1 Vendor: GE Exam Date: 2010 Exam Type: RO Examination Outline Cross-reference: Level RO SRO Tier # 3 Group # 3 KIA # G3 2.3.5 Importance Rating 2.9 Radiation Control: Ability to use radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc.

Proposed Question: RO Question # 70 The plant is operating at 100% power when annunciator L 1-4-3, React Bldg Vent Rad Monitor Off Normal, alarms.

Which one of the following describes the possible cause for this alarm?

A. RWCU leak in the Drywell B. Recirculation Pump seal failure C. Emergency Condenser tube leak D. Spent Fuel Pool Sludge Tank overfill Proposed Answer: D Explanation (Optional):

A. Incorrect - This leak is contained in the drywell, and would only cause containment rad level changes.

B. Incorrect - This failure would only cause containment rad level changes.

C. Incorrect - This leak would result in an EC rad alarm, it is not vented to the stack.

D. Correct - The spent fuel pool sludge tank overflows into the Reactor Building Ventilation system. This would cause 5 mr/hr in the Reactor Building Ventilation exhaust, which

causes the given annunciator.

ARP L 1-4-3 Technical Reference(s): (Attach if not previously provided)

N1-0P-6, Sect. D.3 Proposed References to be provided to applicants during examination: None Learning Objective: N1-272000-RBO-8 (As available)

Question Source: Bank # AUD 2008 RO 52 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: Not used on the 2008 or 2009 exams Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 11 55.43 Purpose and operation of radiation monitoring systems, including alarms and survey equipment.

Comments:

Facility: Nine Mile Point Unit 1 Vendor: GE Exam Date: 2010 Exam Type: RO Examination Outline Cross-reference: Level RO SRO Tier # 3 Group # 3 KJA# G3 2.3.14 Importance Rating 3.4 Radiation Control: Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or emergency conditions or activities.

Proposed Question: RO Question # 71 Which one of the following precautions is required to prevent contamination of the Refuel Floor, in accordance with OP-34, Refueling Procedure?

A. Stow the refueling mast when it is not in use.

B. Keep the steam dryer wetted during and after removal.

C. Maintain forced or natural circulation while shutdown with fuel in the vessel.

D. Avoid storing freshly discharged fuel within 2.5 feet of the Spent Fuel Pool gates.

Proposed Answer: B Explanation (Optional):

A. Incorrect - This precaution is not related to contamination. Refuel bridge operations using tools other than the refuel mast have the potential for inadvertent contact between the refuel mast and the SFP wall or other components or structures. Consideration should be given to stowing the refuel mast when it will not be used.

B. Correct - Failure to maintain steam dryer wetted may result in increased airborne activity levels and personnel contamination.

C. Incorrect - This precaution is not related to contamination. Failure to maintain forced or natural circulation while shutdown with fuel in the vessel may result in thermal stratification and inaccurate temperature indications.

D. Incorrect - This precaution (FHP-25 4.1.12) is related to minimize streaming radiation exposure to personnel on the refuel floor, not contamination.

N1-0P-34 Technical Reference(s): N1-FHP-25 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None Learning Objective: N 1-234000-RBO-9 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam: Not used on the 2008 or 2009 exams Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 12 55.43 Radiological safety principles and procedures.

Comments:

Facility: Nine Mile Point Unit 1 Vendor: GE Exam Date: 2010 Exam Type: RO Examination Outline Cross-reference: Level RO SRO Tier # 3 Group # 4 KIA # G4 2.4.41 Importance Rating 2.9

- - - ...... ~ -

Emergency Procedures I Plan: Knowledge of the emergency action level thresholds and classifications.

Proposed Question: RO Question # 72 Which one of the following describes the lowest Emergency Action Level at which the full Emergency Response Organization (ERO) is required to be staffed?

A. Unusual Event B. Alert C. Site Area Emergency D. General Emergency Proposed Answer: B Explanation (Optional):

A. Incorrect - The Emergency Response Organization is not required to be activated at the UE level.

B. Correct - The OSC and EOF are activated during an Alert, Site Area Emergency or General Emergency. This requires the full Emergency Response Organization.

C. Incorrect - The full Emergency Response Organization is required to be staffed at the lower Alert level.

D. Incorrect - The full Emergency Response Organization is required to be staffed at the lower Alert level.

Technical Reference(s): EPIP-EPP-18 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None 03-0PS-006-350-3-30, EO-1.1 Learning Objective: (As available) 03-0PS-006-350-3-35, EO-1.4 Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam: Not used on the 2008 or 2009 exams Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

Facility: Nine Mile Point Unit 1 Vendor: GE Exam Date: 2010 Exam Type: RO Examination Outline Cross-reference: Level RO SRO Tier # 3 Group # 4 KIA # G4 2.4.45 Importance Rating 4.1 Emergency Procedures I Plan: Ability to prioritize and interpret the significance of each annunciator or alarm.

Proposed Question: RO Question # 73 The plant is operating at 100% power with the following:

Time (hh:mm) Condition 00:00 CRD pump 11 is in service 00:01 Annunciator F3-2-2, CONTROL ROD DRIVE PUMP 11 SUCT PRESS LOW, alarms Annunciator F3-1-2, CONTROL ROD DRIVE PUMP 11 TRIP-VIB, alarms Annunciator F3-1-5, CRD CHARGING WTR PRESSURE HIILO, alarms 00:05 Annunciator F3-2-5, CRD ACCUMULATOR LEVEL HIGH PRESS LOW, alarms Control rod 22-19 accumulator pressure is 990 psig and slowly lowering Which one of the following describes the required operator action(s) and the reason for the action(s), in accordance with SOP-5.1 , Loss of Control Rod Drive?

A. Immediately scram the Reactor to prevent a potential unanalyzed rod pattern.

B. Immediately scram the Reactor to ensure rod insertion times are not exceeded.

C. Start a CRD pump by time 00:25 and then insert a control rod one notch, to prevent spurious control rod drifts.

D. Start a CRD pump by time 00:25 and then insert a control rod one notch, to prevent the need for a manual Reactor scram.

Proposed Answer: D Explanation (Optional):

A. Incorrect - N1-S0P-5.1 only directs the operator to immediately scram the reactor if reactor pressure is less than 900 psig and at rated power reactor pressure is -1025 psig.

B. Incorrect - N1-S0P-5.1 only directs the operator to immediately scram the reactor if reactor pressure is less than 900 psig and at rated power reactor pressure is -1025 psig.

C. Incorrect - N1-S0P-5.1 directs the operator to start a CRD pump to restore charging water pressure and avoid a manual reactor scram. Spurious control rod drifts are the concern in N1-S0P-20.1 when CRD instrument air header pressure is degraded.

D. Correct - The alarms that were received tell the operator that CRD pump 11 has tripped and are entry conditions for N1-S0P-5.1, LOSS OF CONTROL ROD DRIVE. The conditions provided are that the plant is at rated power, therefore reactor pressure can be assumed to be greater than 900 psig. The initial alarms indicate the trip of the CRD pump, the accumulator low pressure alarm completes the scram requirement if the CRD pump is not restarted within 20 minutes. (If no CRD pump is running and any accumulator alarm(s) is received while reactor pressure is greater than 900 pSig, the CRD pump must be restarted within 20 minutes and one Control Rod must be inserted at least one notch OR a Reactor scram is required.) The 20 minute completion time was selected to provide a reasonable time to place a CRD pump in service to restore charging header pressure and recognizes the ability of Reactor pressure of greater than 900 psig alone to fully insert all control rods. If Reactor pressure were less than 900 psig under these same conditions, an immediate scram would be required since Reactor pressure can no longer be counted on to fully insert all control rods.

N1-ARP-F3 Technical Reference(s): (Attach if not previously provided)

N1-S0P-5.1 Proposed References to be provided to applicants during examination: None Learning Objective: N1-201001-RBO-11 (As available)

Question Source: Bank #

Modified Bank # ID: NRC 2006 RO 10 (Note changes or attach parent)

New

Question History: Last NRC Exam: Not used on 2008, similar to 2009 #25 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility_

Comments:

Facility: Nine Mile Point Unit 1 Vendor: GE Exam Date: 2010 Exam Type: R Examination Outline Cross-reference: Level RO SRO Tier # 3 Group # 4 KIA # G4 2.4.34 Importance Rating 4.2 Emergency Procedures I Plan: Knowledge of RO tasks performed outside the main control room during an emergency and the resultant operational effects.

Proposed Question: RO Question # 74 The plant is operating at 100% power with the following:

  • Investigation reveals both Recirculation Pump (RRP) 15 seals are beginning to fail
  • RRP 15 is operating in "Local Lock"
  • SOP-1.2, Recirc Pump Seal Failure, is entered
  • Drywell pressure is 1.2 psig and stable
  • Drywell floor drain leak rate is stable
  • It has been determined that the failures are non-catastrophic Which one of the following actions is required?

A. Immediately place RRP 15 MG Set control switch to STOP and isolate the pump. A plant shutdown is NOT required.

B. Immediately place RRP 15 MG Set control switch to STOP, isolate the pump and start an orderly plant shutdown per OP-43C.

C. Manually reduce RRP 15 flow to 6 to 8 X 106 Ibm/hr using the scoop tube positioner, close the discharge valve, place RRP 15 MG Set control switch to STOP and isolate the pump. A plant shutdown is NOT required.

D. Manually reduce RRP 15 flow to 6 to 8 X 106 Ibm/hr using the scoop tube positioner, close the discharge valve, place RRP 15 MG Set control switch to STOP, isolate the pump and start an orderly plant shutdown per OP-43C.

Proposed Answer: C Explanation (Optional):

A. Incorrect -Only catastrophic failures require immediate tripping of the RRP MG Set in SOP-1.2.

B. Incorrect -Only catastrophic failures require immediate tripping of the RRP MG Set in SOP-1.2.

C. Correct - Only catastrophic failures require immediate tripping of the RRP MG Set in SOP-1.2. The correct action is to reduce RRP flow with the local control, and then shutdown the pump. No requirement exists to perform a plant shutdown based on the given conditions. With initial power at 100%, at least four RRPs must have been in service. That leaves a minimum of three RRPs in operation, which allows continued operation.

D. Incorrect - No plant shutdown is required.

Technical Reference(s): N1-ARP-F2-1-5 (Attach if not previously provided)

N1-S0P-1.2 Proposed References to be provided to applicants during examination: None Learning Objective: N1-202001-RBO-10 (As available)

Question Source: Bank #

Modified Bank # 2008 RO Audit #64 (Note changes or attach parent)

New Question History: Last NRC Exam: Not used on the 2008 or 2009 exams Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

Facility: Nine Mile Point Unit 1 Vendor: GE Exam Date: 2010 Exam Type: R Examination Outline Cross-reference: Level RO SRO Tier # 3 Group # 3 KIA # G3 2.3.4 Importance Rating 3.2 Radiation Control: Knowledge of radiation exposure limits under normal or emergency conditions.

Proposed Question: RO Question # 75 The plant is shutdown with the following:

  • A 31 year old Operator is entering the Drywell for a job
  • General area dose rate is 3 Rem/hr
  • The Operator's TEDE for the year is 1710 mRem
  • The Operator's lifetime exposure is 25 Rem
  • No dose extension is obtained
  • It takes 45 minutes to complete the job Which one of the following states the dose limit(s) exceeded, if any, in order to complete the job, in accordance with GAP-RPP-07, Internal and External DOSimetry Program?

A. None B. Federal only

c. Administrative only D. Federal and Administrative Proposed Answer: C Explanation (Optional):

A. Incorrect - The administrative dose limit without any extensions is 2000 mRem. The worker will have 3960 mRem when the job is completed.

B. Incorrect - The federal yearly limit is 5000 mRem.

C. Correct - The administrative dose control limit without any extensions is 2000 mRem TEDE. The total dose the worker will have received after completing the job will be 1710+3000(45/60)=3960 mRem. A worker can receive up to 4000 mRem TEDE with all of the approved extensions. The NRC yearly limit is 5000 mRem.

D. Incorrect - The federal yearly limit is 5000 mRem.

Technical Reference(s): GAP-RPP-07 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None Learning Objective: GAP-RPP-07 -T001 (As available)

Question Source: Bank # 2007 NRC RO #7 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: Not used on 2008 or 2009 exams Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 12 55.43 Radiological safety principles and procedures.

Comments:

Facility: Nine Mile Point Unit 1 Vendor: GE Exam Date: 2010 Exam Type: SRO ONLY Examination Outline Cross-reference: Level RO SRO Tier# 1 Group # 1 KIA # 295024 EA2.02 Importance Rating 4.0 Ability to determine and/or interpret the following as they apply to HIGH DRYWELL PRESSURE: Drywell temperature Proposed Question: SRO Question # 76 A steam leak in the Drywell has resulted in the following:

  • The mode switch is in SHUTDOWN
  • Drywell average temperature is 299°F and rising slowly
  • Drywell pressure is 11 psig and rising slowly
  • Torus pressure is 9 psig and rising slowly
  • Torus water level is 12 feet and rising slowly
  • RPV pressure is 875 psig and lowering slowly
  • All available Drywell Cooling is in service Which one of the following describes the next action required to be taken?

A. Enter EOP-8, RPV Blowdown, and open three ERVs.

B. Enter EOP-1 Attachment 15 and lower Torus water level.

C. Enter EOP-1 Attachment 17 and initiate Containment Spray.

D. Enter EOP-2, RPV Control, and rapidly depressurize the RPV.

Proposed Answer: C Explanation (Optional):

A. Incorrect - Three less drastic actions are performed before the blowdown is required:

1. Verify reactor scram
2. Enter EOP-2, RPV Control
3. Initiate containment sprays Therefore containment sprays must be initiated to attempt to lower Orywell Temperature before the blowdown determination is made.

B. Incorrect - While torus level is high, we are still below the spray ring (13.5'). No action, beyond monitoring, would be expected at this time, as use of Containment Spray for lowering Containment pressures and temperatures would be the priority. Additionally, Containment Spray valves80-114 and 80-115 receives close signals when OW pressure is above 3.5 psig, which would prevent establishing the lineup in EOp-1 attachment 15.

C. Correct - With OWT at 299°F and rising, Containment Spray is required. Orywell pressure at 11 psig is within the SPRAY region of the Containment Spray Initiation Limit. Therefore Containment Sprays must be initiated to attempt to lower Orywell Temperature before the blowdown is performed. If following use of Containment Spray, the determination is made that you cannot restore and maintain Orywell temperature below 300°F, a blowdown must be performed.

O. Incorrect - Rapid depressurization of the RPV is only allowed when RPV Blowdown is anticipated. RPV Blowdown should not be anticipated until all EOP actions designed to prevent Blowdown have been exhausted.

EOP-4 Technical Reference(s): EOP-1 (Attach if not previously provided)

GAI-OPS-20 Section 1.2.3.1 Containment Spray Proposed References to be provided to applicants during examination: Initiation Limit (K) curve Learning Objective: N 1-223001-RBO-12 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New x Question History: Last NRC Exam: Not used on 2008 or 2009 NRC exams Question Cognitive Level: Memory or Fundamental Knowledge

Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 55.43 5 Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Comments:

Facility: Nine Mile Point Unit 1 Vendor: GE Exam Date: 2010 Exam Type: SRO ONLY Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 KIA # 295004 AA2.04 Importance Rating 3.3 Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF D.C. POWER: System lineups Proposed Question: SRO Question # 77 The plant is operating at 100% power with the following:

  • Static Battery Charger (SBC) 161B develops a fault
  • SBC 161A cannot be placed in service
  • Electrical Maintenance is troubleshooting the condition and reports the expected restoration time is unknown Which one of the following actions is required for these conditions?

A. Restore either battery charger to service or place the plant in a cold shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

B. Place MG 167 in Battery Charger mode and continue plant operations until the SBC is restored to service.

C. Restore either battery charger to service within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or place the plant in a cold shutdown condition within 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />.

D. Restore either battery charger to service within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or reduce Reactor pressure to less than 110 psig within 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />.

Proposed Answer: D Explanation (Optional):

A. Incorrect - A battery charger must be returned to service within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> per TS.

3.6.3.h or take the action required by T.S. 3.1.5, which requires reactor coolant pressure be reduced to 110 pSig or less and reactor coolant temperature be reduced to saturation temperature or less within 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />.

B. Incorrect - Although MG 167 can be used as a battery charger, it is not safety related and cannot be used to avoid entry into TS 3.6.3.h LCO.

C. Incorrect - TS 3.1.5 requires reactor coolant pressure be reduced to 110 psig or less and reactor coolant temperature be reduced to saturation temperature or less within 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />. Cold shutdown is defined as 212°F, which would equate to 0 psig and be more restrictive than the actual specification.

D. Correct - The loss of SBC 161B with SBC 161A means no battery charger is available to Battery Board 11. Per OP-47 A P&L 9, this is treated as a loss of the battery, requiring a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> LCO for return of a SBC. If no SBC is returned within the 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, TS 3.1.5 is entered, requiring reactor coolant pressure be reduced to 110 psig or less and reactor coolant temperature be reduced to saturation temperature or less within 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />.

Technical Reference(s): N1-0P-47A, T.S. 3.1.5, 3.6.3 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: T.S. 3.1.5, 3.6.3.

Learning Objective: N 1-263000-RBO-14 (As available)

ID: N1-263000-RBO-Question Source: Bank #

14-Q-10 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: Not used on the 2008 or 2009 exams Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 55.43 2 Facility operating limitations in the technical specifications and their bases.

Comments:

Facility: Nine Mile Point Unit 1 Vendor: GE Exam Date: 2010 Exam Type: SRO ONLY Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 KIA # 295001 AA2.02 Importance Rating 3.2 Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF FORCED CORE FLOW CIRCULATION: Neutron monitoring Proposed Question: SRO Question # 78 The plant is operating at 90% power with the following:

  • 5 Recirc Loops are operating
  • Recirc Pump (RRP) 14 trips
  • eRS has directed the RO to close RRP 14 discharge valve Which one of the following describes the impact on the Flow-Biased Scram setpoint as the RRP 14 discharge valve is being closed and the impact on Technical Specifications?

The Flow-Biased Scram setpoint becomes ... Technical SpeCification Impact A. Less conservative APRMs will become operable B. More conservative APRMs will become operable C. Less conservative Reactor power limit will be 90.5%

D. More conservative Reactor power limit will be 90.5%

Proposed Answer: B Explanation (Optional):

A. Incorrect - When the RRP 14 discharge valve is shut the indicated flow will lower to match actual core flow, which lowers the flow-biased scram setpoint {makes it more

conservative) and makes the APRMs operable.

B. Correct. With RRP 14 discharge valve open, reverse flow through the idle loop is measured as positive flow which makes the indicated core flow higher than actual. This makes the APRM flow biased scram setpoints less conservative. When the RRP 14 discharge valve is shut, the indicated flow will lower to match actual core flow, which lowers the flow-biased scram setpoint (makes it more conservative) and makes the APRMs operable. T.S. 3.1.7 states when operating 4 loop with the non-operating loop un-isolated, the plant can operate at rated power. Having only the discharge valve shut means the loop is idle, not isolated. In this condition, the loop is maintained warm.

C. Incorrect - Reactor power is only limited to 90.5% if a RRP loop is isolated. TS. 3.1.7 states when operating 4 loops with the non-operating loop un-isolated, the plant can operate at rated power. Having only the discharge valve shut means the loop is idle, not isolated.

D. Incorrect - Reactor power is only limited to 90.5% if a RRP loop is isolated. TS.3.1.7 states when operating 4 loops with the non-operating loop un-isolated, the plant can operate at rated power. Having only the discharge valve shut means the loop is idle, not isolated.

Technical Reference(s): TS 3.1.7 (Attach if not previously provided)

N1-S0P-1.3 Proposed References to be provided to applicants during examination: TS.3.1.7 Learning Objective: N 1-20200 1-RBO-14 (As available)

ID: N1-202001-RBO-Question Source: Bank #

14-Q-04 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: Not used on 2008 or 2009 exams (2009 #90 is three loop power)

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 1 Conditions and limitations in the facility license Comments:

Facility: Nine Mile Point Unit 1 Vendor: GE Exam Date: 2010 Exam Type: SRO ONLY Examination Outline Cross-reference: Level RO SRO Tier# 1 Group # 1 KIA # 295038 2.4.18 Importance Rating 4.0 Emergency Procedures I Plan: Knowledge of the specific bases for EOPs. (High Off-Site Release Rate)

Proposed Question: SRO Question # 79 N1-EOP-2, RPV Control, has been entered with Reactor water level continuing to drop.

STEAM COOl:

~~re ... Entl!r'EOP-G

~-14 BEFORE BLOW DOWN:

En1er EOP-8 while ....

oontn.i!'!Q hel1i!.

L-16 Verify Liquid Poison iosec;iol' (EOP-l Att 13}.

- njeet emire OPnlen'!i af L(Ida Poison tank.

L-17 Which one of the following describes the basis for injecting Liquid Poison per Step L-17 of EOP-2?

A. Provides an additional source of injection to the Reactor.

B. Minimizes radioactive release by making the Torus pH higher (less acidic).

C. Minimizes radioactive release by making the Torus pH lower (more acidic).

D. Minimizes radioactive release by borating the coolant inside the Reactor vessel.

Proposed Answer: B Explanation (Optional):

A. Incorrect - Establishing Liquid Poison in Step L-17 is not as a means for level control but as a means to control radiological dose following a loss of coolant accident involving core damage. Since Liquid Poison is identified as an Alternate Injection System it would likely be started to augment RPV injection in an earlier step of the Level branch, before RPV water level reaches the top of the active fuel (Element L-3, L-7, or L-9).

This is a plausible distracter for those candidates that do not recognize the radiological impact from Liquid Poison injection once the TAF has been reached.

B. Correct - Design basis analyses credit Liquid Poison injection for limiting the radiological dose following loss of coolant accidents involving core damage. Radiation induced reactions are predicted to convert large fractions of dissolved ionic iodine into elemental iodine and organic iodides which can escape into the containment atmosphere. The rate of these reactions is strongly dependent on suppression pool pH. If the pH is maintained greater than 7, very little of the dissolved iodine will be converted to volatile forms and most of the iodine fission products will be retained in the suppression pool.

Over time, the pH in the torus will tend to decrease due to the addition of acidic chemicals. The sodium pentaborate solution used in the Liquid Poison system is derived from a strong base and therefore raises suppression pool pH.

C. Incorrect - As described above, Liquid Poison is injected to control and raise Torus pH following the onset of a LOCA. This is a plausible distracter for those candidates who are unsure of the addition of Liquid Poison raises or lowers pH in the Torus.

D. Incorrect - Boration of the reactor coolant is performed to reduce power levels in the core by neutron moderation. This is plausible distracter for those candidates who believe that dose mitigation is achieved with boration of the coolant in the vessel versus the torus volume.

Technical Reference(s): EOP-2, NER-1 M-095 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None Learning Objective: 01-0PS-006-344-1-03, EO-1.2 & 1.3 (As available)

Question Source: Bank # 2009 NRC #98 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: 2009 NRC #98 Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 5 Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Comments:

Facility: Nine Mile Point Unit 1 Vendor: GE Exam Date: 2010 Exam Type: SRO ONLY Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 KIA # 295026 2.2.38 Importance Rating 4.5 Equipment Control: Knowledge of conditions and limitations in the facility license (Suppression Pool High Water Temp).

Proposed Question: SRO Question # 80 A plant startup is in progress with the following:

  • N1-ST-C2, Solenoid-Actuated Pressure Relief Valves Operability and Flow Verification Test, is being performed
  • Torus average water temperature is rising Which one of the following describes the Torus average water temperature limits during and after the test, in accordance with Technical Specifications?

Torus average water temperature shall NOT exceed (1) during performance of the test and must be reduced below the normal power operation limit within (2)

(2)

A. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> B. 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> C. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> D. 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Proposed Answer: D Explanation (Optional):

A. Incorrect - Temperature is allowed up to 95 degrees during ERV testing, and 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> are allowed to return temperature below 85 degrees.

B. Incorrect - Temperature is allowed up to 95 degrees during ERV testing.

C. Incorrect - 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> are allowed to return temperature below 85 degrees.

D. Correct - Per TS 3.3.2.d, during testing of relief valves that add heat to the torus the operating limit of 85°F is raised 10°F to 95°F. Twenty-four hours are permitted to return the temperature to less than 85°F.

Technical Reference(s): TS 3.3.2.d (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None Learning Objective: N1-223001-RBO-14 (As available)

Question Source: Bank #

M dT dB k # NRC RETAKE 2005 (Note changes or attach parent) o lie an #14 New Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 1 Conditions and limitations in the facility license Comments:

Facility: Nine Mile Point Unit 1 Vendor: GE Exam Date: 2010 Exam Type: SRO ONLY Examination Outline Cross-reference: Level RO SRO Tier# 1 Group # 1 KIA # 295037 2.2.39 Importance Rating 4.5 Equipment Control: Knowledge of less than or equal to one hour technical specification action statements for systems (SCRAM Conditions Present and Reactor Power Above APRM Downscale or Unknown).

Proposed Question: Question # 81 A plant startup is in progress with the following:

  • The mode switch is in STARTUP
  • Reactor pressure is 926 psig
  • All other IRMs are midscale on range 9
  • All APRM downscales have cleared Then, the range switches for IRMs 12 and 13 are mistakenly moved to range 7.
  • No other alarms occur
  • The range switch positions are corrected
  • Annunciator F2-3-6 clears Which one of the following describes an associated Technical Specification requirement, if any?

A. No further actions are required.

B. Immediately place RPS channel 11 in the tripped condition.

C. One hour is allowed to verify sufficient IRM channels are operable or tripped.

D. Twelve hours are allowed to place RPS channel 11 in the tripped condition.

Proposed Answer: C Explanation (Optional):

A. Incorrect - A half-scram should have been generated with IRMs 12 and 13 on range 7.

With these two channels inoperable for the IRM neutron flux scram, Tech Spec actions are required.

B. Incorrect - There are no requirements to immediately place the RPS channel in the tripped condition.

C. Correct - A half-scram should have occurred, because with the IRMs midscale on range 9, when the range switch is positioned to range 7 the reading will be above the top of the scale. The failure to trip and cause a half-scram makes IRMs 12 and 13 inoperable.

Since IRM 11 in Channel 11 is also inoperable there are 3 inoperable channels in the trip system. T.S. Table 3.6.2.a requires 3 operable channels. With only IRM channel 14 operable, the plant is 2 channels below this requirement. Note (0) states:

With two or more channels required by Table 3.6.2a inoperable in one or more Parameters:

1. Within one hour, verify sufficient channels remain Operable or tripped* to maintain trip capability for the Parameter, and
2. Within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, place the inoperable channel(s) in one trip system and/or that trip system** in the tripped condition*, and
3. Within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, restore the inoperable channels in the other trip system to an Operable status or tripped*.

IRM 14 is the only remaining channel in Trip system 11, and therefore must be verified operable or tripped within one hour.

D. Incorrect - The time requirement is more restrictive than 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

Technical Reference(s): T.S. 3.6.2.a (Attach if not previously provided)

TS 3.6.2.a and Proposed References to be provided to applicants during examination:

tables Learning Objective: N1-215000-RBO-14 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam: Not used on the 2008 or 2009 exams Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 55.43 2 Facility operating limitations in the technical specifications and their bases.

Comments:

Facility: Nine Mile Point Unit 1 Vendor: GE Exam Date: 2010 Exam Type: SRO ONLY Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 KIA # 295005 2.2.37 Importance Rating 4.6 Equipment Control: Ability to determine operability and I or availability of safety related equipment (Main Turbine Generator Trip).

Proposed Question: SRO Question # 82 The plant is operating at 100% power. I & C reports that a document review has revealed the Turbine Stop Valve closure scram setpoints are set according to the table below.

Turbine Stop Valve Channel 11 Setpoint Channel 12 Setpoint

(% valve closure) (% valve closure) 11 9 10 12 8 11 13 9 10 14 8 12 Which one of the following describes the significance of these setpoints and their effects following a Turbine Trip?

The ChanneL ..

A. 12 setpoints are too high. This will narrow the margin to MCPR following a Turbine Trip.

B. 11 setpoints are too low. This will narrow the margin to MCPR following a Turbine Trip.

C. 12 setpoints are too high. This will result in an excessive RPV level transient following a Turbine Trip.

D. 11 setpoints are too low. This will result in an excessive RPV level transient following a Turbine Trip.

Proposed Answer: A

Explanation (Optional):

A. Correct - The channel 12 setpoint for TSV 12 is above 10% which is outside the requirements of TS 2.1.2.i and TS 3.6.2 of equal to or less than 10%. The turbine stop valve closure scram is provided to anticipate the rapid increase in pressure and neutron flux resulting from fast closure of the turbine control valves due to the worst case transient of a load rejection and subsequent failure of the bypass. The scram setpoints are chosen to ensure MCPR is not violated during the transient.

B. Incorrect - The channel 11 setpoints are in spec.

C. Incorrect - There are no concerns with these setpoints and the change in RPV water level.

D. Incorrect - The channel 11 setpoints are in spec. There are no concerns with these setpoints and the change in RPV water level.

  • IR ~

Tec hnlca () T.S. 2.1.2.i and Bases (Attach if not previously provided) e.erence s: TS 3.6.2 Proposed References to be provided to applicants during examination: None Learning Objective: N1-245000-RBO-14 (As available)

Question Source: Bank #

Modified Bank # 88331 (Note changes or attach parent)

New Question History: Last NRC Exam: not used on the 2008 or 2009 exams Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 55.43 2 Facility operating limitations in the technical specifications and their bases.

Comments:

Facility: Nine Mile Point Unit 1 Vendor: GE Exam Date: 2010 Exam Type: SRO ONLY Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 2 KIA # 295020 AA2.03 Importance Rating 3.7 Ability to determine and/or interpret the following as they apply to INADVERTENT CONTAINMENT ISOLATION: Reactor power Proposed Question: SRO Question # 83 The plant is operating at 100% power with the following:

  • The cause of the inadvertent isolation has been corrected
  • RWCU restoration is commencing Which one of the following restrictions apply prior to placing RWCU back in service?

Lower Reactor power...

A. to less than 90% using OP-438, Normal Power Operations.

8. to less than 90% using SOP-1.1, Emergency Power Reduction.

C. approximately 15 MWth using OP-438, Normal Power Operations.

D. approximately 15 MWth using SOP-1.1, Emergency Power Reduction.

Proposed Answer: C Explanation (Optional):

A. Incorrect Power needs to be lowered approximately 15 MWth.

B. Incorrect - Power needs to be lowered approximately 15 MWth. A normal, planned

power reduction, such as this, is performed using OP-43B, not SOP-1.

C. Correct - Per OP-3, prior to restoring RWCU to service when at rated power, power must be lowered approximately 15 MWth. A normal, planned power reduction, such as this, is performed using OP-43B, not SOP-1.1.

D. Incorrect - The procedure for changing power at these conditions is N1-0P-43B.

  • IR f Techmca () N1-0P-3, Sect. H.1, E.2 and P & L (Att h'f t . I 'd d) eerence s: 10, N1-0P-43B, N1-S0P-1.1 ac I no prevlousy provi e Proposed References to be provided to applicants during examination: None Learning Objective: N1-204000-RBO-9 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam: Not used on 2008 or 2009 exams Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 5 Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Comments:

Facility: Nine Mile Point Unit 1 Vendor: GE Exam Date: 2010 Exam Type: SRO ONLY Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 2 KIA # 295007 2.4.6 Importance Rating 4.7 Emergency Procedures I Plan: Knowledge of EOP mitigation strategies (High Reactor Pressure).

Proposed Question: SRO Question # 84 A plant startup is in progress with the following:

  • Reactor power is 2%
  • Reactor pressure is 500 psig Then, a LOCA occurs and the following conditions exist:
  • ADS is bypassed
  • Core Spray pumps automatically started and are running
  • No additional injection systems are available
  • RPV water level is -75" and lowering slowly
  • RPV pressure is 400 psig and lowering slowly Which one of the following EOP strategies is appropriate under these conditions?

A. Enter EOP-8, RPV Slowdown, and open three ERVs.

S. Enter EOP-9, Steam Cooling, and continue to monitor RPV water level.

C. Install Core Spray jumpers per EOP-1 Attachment 4, Throttling Core Spray.

D. Remain in EOP-2, RPV Control, and lower pressure with Emergency Condensers.

Proposed Answer: D

Explanation (Optional):

A. Incorrect - EOP-8, RPV Blowdown would only be entered if level was less than -84".

B. Incorrect - Steam cooling would only be entered if no injection source was available and level was less than -84".

C. Incorrect - This would be done to prevent Core Spray injection not needed for core cooling. With RPV water level dropping below the top of active fuel, Core Spray is needed for core cooling. Installation of Core Spray jumpers would bypass the automatic opening of Core Spray IVs at 365 psig under the exact conditions for which Core Spray is designed. Core Spray jumper installation is only appropriate after the IVs have stroked fully open.

D. Correct - Blowdown is not currently allowed because RPV water level is above -84".

EOP-2 gives direction to maximize injection sources to maintain water level above -84".

RPV pressure is too high for Core Spray to inject, but Emergency Condensers are available to lower pressure to within the capability of Core Spray. RPV pressure was low enough at the beginning of the transient to allow pressure to be immediately lowered below 365 psig without violating the 1OOoF/hr cooldown rate.

Technical Reference(s): N1-EOP-2, GAI-OPS-20, NER (Attach if not previously provided) 1M-095 Proposed References to be provided to applicants during examination: None Learning Objective: 01-0PS-006-342-03 EO-1.2 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam: not used on 2008 or 2009 exams Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 5

Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Comments:

Facility: Nine Mile Point Unit 1 Vendor: GE Exam Date: 2010 Exam Type: SRO ONLY Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 2 KIA # 295014 AA2.03 Importance Rating 4.3 Ability to determine and/or interpret the following as they apply to INADVERTENT REACTIVITY ADDITION: Cause of reactivity addition Proposed Question: SRO Question # 85 The plant is operating at 30% power when the following sequence occurs:

Time (mm:ss) Condition 00:00 F3-4-6, FIRST STAGE BOWL PRESS LOW, is in alarm 01:00 Stator Water Cooling system temperature is 95°C and slowly riSing A Generator runback begins 03:05 Generator amps indicate 4400 04:00 Steady-state conditions are achieved Which one of the following describes the Reactor power response to this transient and the procedure that should be entered?

Reactor power is ... Enter ...

A. Higher due to Feedwater temperature changes SOP-31.1, Turbine Trip B. Lower due to Reactor pressure changes SOP-31.1, Turbine Trip C. Higher due to Feedwater temperature changes SOP-32, Generator Auxiliaries Failures D. Lower due to Reactor pressure changes SOP-32, Generator Auxiliaries Failures

Proposed Answer: C Explanation (Optional):

A. Incorrect -The cooler Feedwater temperatures entering the Reactor cause a rise in Reactor power. Because the runback has successfully lowered generator output to less than 4450 amps, no turbine trip is required.

B. Incorrect - The runback will allow a large volume of steam to bypass the turbine and be directed into the main condenser, which will result in a loss of Feedwater heating. The cooler Feedwater temperatures entering the Reactor cause a rise in Reactor power.

C. Correct - During a loss of Stator Water Cooling, the turbine control circuit will attempt to lower turbine input to the generator and hence generator output by closing the TCVs and opening the TBVs. This will allow a large volume of steam to bypass the turbine and be directed into the main condenser, which will result in a loss of Feedwater heating. The cooler Feedwater temperatures entering the Reactor cause a rise in Reactor power. Because the runback has successfully lowered generator output to less than 4450 amps, no turbine trip is required. SOP-32 provides the necessary guidance to deal with the runback.

D. Incorrect - The runback will allow a large volume of steam to bypass the turbine and be directed into the main condenser, which will result in a loss of Feedwater heating. The cooler Feedwater temperatures entering the Reactor cause a rise in Reactor power.

Because the run back has successfully lowered generator output to less than 4450 amps, no turbine trip is required. SOP-32 provides the necessary guidance to deal with the run back.

ARP A2-2-2 Technical Reference(s): SOP-32 (Attach if not previously provided)

SOP-1.5 Proposed References to be provided to applicants during examination: None Learning Objective: N1-253000-RBO-11 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X

Question History: Last NRC Exam: Not used on 2008 or 2009 exams Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 55.43 5 Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Comments:

Facility: Nine Mile Point Unit 1 Vendor: GE Exam Date: 2010 Exam Type: SRO ONLY Examination Outline Cross-reference: Level RO SRO Tier# 2 Group # 1 KIA # 259002 A2.07 Importance Rating 2.5 Ability to (a) predict the impacts of the following on the REACTOR WATER LEVEL CONTROL SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Loss of comparator bias signal Proposed Question: SRO Question # 86 The plant is operating at 100% power with the following:

  • RPV water level is 72" and stable Then, the total main steam flow signal to the steam flow-feed flow comparator fails to ZERO.

Which one of the following describes the impact of the RPV water level change with no operator action, and the required action in response to this magnitude of RPV water level change?

RPV water level lowers and ...

A. power operation continues. No Reactor scram occurs. Enter SOP-16.1, Feedwater System Failures, and direct manual control of FWLC.

B. power operation continues. No Reactor scram occurs. Enter SOP-16.1, Feedwater System Failures, and direct an emergency power reduction.

C. the Reactor scrams on low level. Enter SOP-1, Reactor Scram, and SOP-16.1, Feedwater System Failures. Direct control of FWLC from the Control Room.

D. the Reactor scrams on low level. Enter SOP-1, Reactor Scram, and SOP-16.1, Feedwater System Failures. Direct control of FWLC by pinning Feedwater FCVs.

Proposed Answer: C Explanation (Optional):

A. Incorrect - The magnitude of the steam-feed flow bias in three-element FWLC will result in a 40" level change with total loss of steam flow signal. This will cause a low level scram at 53".

B. Incorrect - The magnitude of the steam-feed flow bias in three-element FWLC will result in a 40" level change with total loss of steam flow signal. This will cause a low level scram at 53".

C. Correct - Three-Element Mode of Control is used when Reactor Power is greater than 25%. The Feedwater Control System sums the individual feed flow and steam flow signals to develop Total Steam Flow and Total Feedwater Flow signals. The total steam and feed flow signals are then sent to a comparator to determine if there is mismatch between steam flow and feed flow. This produces a steam flow/feed flow error signal that is used to anticipate changes in Reactor level. With a lower steam flow, the FWLC system senses a lower demand in FW flow, closes FW FCV and level lowers.

Because of the failure of the total steam flow output signal to zero, the magnitude of the RPV level change demanded is approximately 40 inches. Level will lower below the low level scram setpoint of 53". The reactor will scram on the low level, requiring entry into SOP-1 and SOP-16.1. Only automatic 3-element control of FWLC is affected by the loss, therefore manual control is available from the Control Room. Additionally, after the scram main steam flow is so low that even 3-element control with 11 or 12 FW FCV would be fairly accurate with the failed signal.

D. Incorrect - Only automatic 3-element control of FWLC is affected by the loss, therefore manual control is available from the Control Room.

N1-0P-16 N1-S0P-16.1 Technical Reference(s): (Attach if not previously provided)

N1-S0P-1 N1101259002C01 Proposed References to be provided to applicants during examination: None Learning Objective: N 1-259002-RBO-08 (As available)

Question Source: Bank #

'f' d B k # #339 from the I & C Mod lie (Note changes or attach parent) an Bank New

Question History: Last NRC Exam: Not used on the 2008 or 2009 exams Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 55.43 5 Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Comments:

Facility: Nine Mile Point Unit 1 Vendor: GE Exam Date: 2010 Exam Type: SRO ONLY Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 KIA # 211000 A2.07 Importance Rating 3.2 Ability to (a) predict the impacts of the following on the STANDBY LIQUID CONTROL SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Valve closures Proposed Question: SRO Question # 87 A failure to scram has occurred with the following:

  • Reactor power is 2% and slowly lowering
  • RPV water level is -65 inches and stable
  • Torus water temperature is 112"F and slowly rising
  • Drywell pressure is 2 psig and stable
  • The Liquid Poison (LP) selector switch has been taken to SYS 11, and then SYS 12
  • LP pump 12 is running
  • Both LP explosive valve white lights are energized
  • The RO is manually inserting control rods using EOP-3.1 Which one of the following describes required actions in response to these conditions?

A. Initiate Alternate Boron Injection using EOP-3.2. Continue inserting control rods using EOP-3.1.

B. Continue inserting control rods using EOP-3.1. Do NOT initiate Alternate Boron Injection using EOP-3.2.

C. Initiate Alternate Boron Injection using EOP-3.2. Terminate and prevent all RPV injection except Boron and CRD.

D. Terminate and prevent all RPV injection except Boron and CRD. Do NOT initiate Alternate Boron Injection using EOP-3.2.

Proposed Answer: A Explanation (Optional):

A. Correct - lAW EOP-3, with torus temperature above 11 OQF and the reactor not shutdown boron must be injected. With the explosive valves closed, as evidenced by the white lights still energized, the Liquid Poison system is NOT available and the crew must enter EOP-3.2 and start alternate liquid poison injection. Control rod insertion is continued in EOP-3.1 until a rod pattern is achieved that allows exit from EOP-3.

Q B. Incorrect - The crew must enter EOP-3.2 because torus temperature is above 110 F, the reactor not shutdown, and the LP explosive valves did not fire.

C. Incorrect - The conditions to re-terminate and prevent in EOP-3 are not met (Power above 6%, Level above -84", an ERVopen or OW press above 3.5#, and Torus temp above 110F).

O. Incorrect - The crew must enter EOP-3.2 because torus temperature is above 11O QF, the reactor not shutdown, and the LP explosive valves did not fire. The conditions to re terminate and prevent in EOP-3 are not met (Power above 6%, Level above -84", an ERV open or OW press above 3.5#, and Torus temp above 110F).

Technical Reference(s): EOP-3 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None Learning Objective: 01-0PS-006-344-1-03, EO-1.2 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam: Not used on the 2008 or 2009 exams Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41

55.43 5 Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Comments:

Facility: Nine Mile Point Unit 1 Vendor: GE Exam Date: 2010 Exam Type: SRO ONLY Examination Outline Cross-reference: Level RO SRO Tier# 2 Group # 1 KIA # 205000 2.2.12 Importance Rating 4.1 Equipment Control: Knowledge of surveillance procedures (Shutdown Cooling).

Proposed Question: SRO Question # 88 The plant is in cold shutdown during a refueling outage with the following:

  • N1-ST-C13, Reactor Shutdown Cooling System Valve Leakage Test, has been performed
  • The Shutdown Cooling System (SDC) was removed from service during the test, and is now back in service
  • Operations with the Potential to Drain the Reactor Vessel (OPDRVs) are in progress
  • Check Valve 38-12, SDC SYSTEM OUT IV 1 (OUTSIDE), is determined to have excess leakage
  • All other SDC valves tested satisfactorily Which one of the following describes the required action, if any, per Technical Specifications?

A. SDC is inoperable and all OPDRVs must be immediately stopped.

B. No action is required because 38-13, SDC SYSTEM OUT IV 1 (INSIDE), is operable.

C. No action is required because 38-12, SDC SYSTEM OUT IV 1 (OUTSIDE), is NOT a Containment Isolation Valve.

D. Verify closed 38-13, SDC SYSTEM OUT IV 1 (INSIDE), within four hours, or immediately initiate action to suspend all OPDRVs, or immediately initiate action to restore 38-12, SDC SYSTEM OUT IV 1 (OUTSIDE), to operable status.

Proposed Answer: D

Explanation (Optional):

A. Incorrect - OPDRVs must not be immediately stopped, T.S. 3.2.7.e permits activities to continue for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> prior to action being required.

B. Incorrect - With one SDC IV inop, 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> are allowed to close a valve in the common line. Just being "operable" is NOT sufficient, the valve must be closed.

C. Incorrect 12 is a Containment Isolation Valve.

D. Correct 12 is a Containment Isolation Valve. With coolant temperature less than 212°F, TS 3.2.7.d and e apply. With one SDC IV inop, 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> are allowed to close a valve in the common line. For 38-12 the common line valve is 38-13. If 38-13 is not closed within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, TS 3.2.7.f requires OPDRVs to be suspended or immediate action be initiated to restore the inop valve to operable status.

TS 3.2.7 Technical Reference(s): N1-ST-C13 (Attach if not previously provided)

C-18018-C Proposed References to be provided to applicants during examination: TS 3.2.7 Learning Objective: N1-205000-RBO-14 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam: Not used on 2008 or 2009 exams Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 2 Facility operating limitations in the technical specifications and their bases.

Comments:

Facility: Nine Mile Point Unit 1 Vendor: GE Exam Date: 2010 Exam Type: SRO ONLY Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 KIA # 264000 2.1.32 Importance Rating 4.0 Conduct of Operations: Ability to explain and apply all system limits and precautions (EDGs).

Proposed Question: SRO Question # 89 The plant is operating at 100% power with the following:

  • Power Control calls to report lowering grid voltage.
  • Power Control reports additional generation is being placed on the grid and that grid conditions are expected to improve shortly.
  • Power Control indicates the Load Flow Computer is unavailable.
  • Grid voltage is 112 KV and slowly lowering on all phases.

Which one of the following is required?

A. Enter N1-S0P-33A.1, Loss of 115 KV; ensure EDG 102 OR EDG 103 is running.

B. Enter N1-S0P-33A.1, Loss of 115 KV; ensure EDG 102 AND EDG 103 are running.

C. Enter N1-S0P-33A.3, Major 115 KV Grid Disturbances; ensure EDG 102 OR EDG 103 is running.

D. Enter N1-S0P-33A.3, Major 115 KV Grid Disturbances; ensure EDG 102 AND EDG 103 are running.

Proposed Answer: C Explanation (Optional):

A. Incorrect - SOP-33A.1 is for a total loss of 115KV.

B. Incorrect - SOP-33A.1 is for a total loss of 115KV. With both offsite lines inoperable due to low voltage, TS 3.6.3 requires only one EDG running.

C. Correct - N1-0P-33A P&L #7 discusses operability of 115 KV offsite power lines and references N1-S0P-33A. The stem contains the entry points to N1-S0P-33.A.3. With no load flow computer available and grid voltage less than 114 KV, both offsite lines must be declared inoperable. TS 3.6.3.e.(2) requires one EDG to be running in this condition.

D. Incorrect - TS 3.6.3.e.(2) requires one EDG to be running in this condition.

N1-0P-33A, P & L 7 Technical Reference(s): N1-S0P-33A.3 (Attach if not previously provided)

TS 3.6.3 Proposed References to be provided to applicants during examination: TS 3.6.3 Learning Objective: N 1-262000-RBO-1 0 (As available) 2008 Audit Exam Question Source: Bank #

  1. 88 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: Not used on 2008 or 2009 exams Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 55.43 5 Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Comments:

Facility: Nine Mile Point Unit 1 Vendor: GE Exam Date: 2010 Exam Type: SRO ONLY Examination Outline Cross-reference: Level RO SRO Tier# 2 Group # 1 KIA # 207000 2.2.25 Importance Rating 4.2 Equipment Control: Knowledge of bases in technical specifications for limiting conditions for operations and safety limits (Emergency Condensers).

Proposed Question: SRO Question # 90 The plant is operating at 100% power with the following:

  • Both Emergency Condenser (EC) Makeup Tank water levels are found to be below the minimum required value per Technical Specifications (TS)
  • Efforts to raise EC Makeup Tank water level are NOT successful
  • TS 3.1.3.e states "a normal orderly shutdown shall be initiated within one hour, and the reactor shall be in the cold shutdown condition within ten hours" Which one of the following describes when a manual Reactor scram must be inserted to meet the ten hour cold shutdown requirement and the basis for the EC Makeup Tank minimum water level?

Insert a manual Reactor scram at approximately (1) hours into the LCO. The EC Makeup Tank minimum water level provides approximately (2) hours of continuous system operation.

(1 ) (2)

A. three (3) eight (8)

B. three (3) forty-eight (48)

C. four (4) eight (8)

D. four (4) forty-eight (48)

Proposed Answer: A Explanation (Optional):

A. Correct - When required to shutdown and cool down to meet a 10-hour LCO requirement, it is necessary insert a manual scram approximately 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> into the LCO.

It is also necessary to have SDC permissives met approximately 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> into the LCO to achieve cold shutdown in 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />.

TS 3.1.3 bases state the EC Makeup Tank minimum water level provides about 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> of continuous system operation.

B. Incorrect - TS 3.1.3 bases state the EC Makeup Tank minimum water level provides about 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> of continuous system operation.

C. Incorrect - It is necessary insert a manual scram approximately 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> into the LCO.

D. Incorrect - It is necessary insert a manual scram approximately 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> into the LCO.

  • IR f T ec hmca () T.S. Bases for 3.1.3 (Att h'f t . I 'd d) e erence s: OP-43C P&L 19, OP-13 Section B ac I no prevIous y provi e Proposed References to be provided to applicants during examination: None Learning Objective: N 1-207000-RBO-14 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam: Not used on 2008 or 2009 exams Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 2 Facility operating limitations in the technical speCifications and their bases.

Comments:

Facility: Nine Mile Point Unit 1 Vendor: GE Exam Date: 2010 Exam Type: SRO ONLY Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 2 KIA # 286000 A2.08 Importance Rating 3.3 Ability to (a) predict the impacts of the following on the FIRE PROTECTION SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Failure to actuate when required Proposed Question: SRO Question # 91 The plant is operating at 100% power with the following:

  • A fire has been confirmed in Diesel Generator Room 102.
  • A Plant Operator has verified the CO2 storage tank is aligned to supply the fire suppression header.
  • An Operator attempted to manually discharge CO2 from the Main Fire Panel.
  • CO2 failed to initiate due to a loss of DC power.

Which one of the following describes the action required to activate CO2 and the level of evacuation that is required based on the CO2 discharge?

Direct an Operator to ...

A place the control switch at the Local Fire Panel to manual. Direct an evacuation of the Protected Area.

B. place the control switch at the Local Fire Panel to manual. Direct an evacuation of the Diesel Generator Room, only.

C. open the Hazard Block Valve using the EMPC manual control valve override. Direct an evacuation of the Protected Area.

D. open the Hazard Block Valve using the EMPC manual control valve override. Direct an evacuation of the Diesel Generator Room, only.

Proposed Answer: C Explanation (Optional):

A. Incorrect - The control switch at the Local Fire Panel will not operate without DC power.

B. Incorrect - The control switch at the Local Fire Panel will not operate without DC power.

A protected area evacuation is required since the C02 discharge leads to an Alert emergency classification.

C. Correct - With a loss of DC power, the Hazard Block Valve will not open via any initiation signal. Manual override of the corresponding EMPC is required. This discharge of C02 meets the threshold for Alert EAL 8.3.5. Per EPIP-EPP-18, a protected area evacuation will be directed at the Alert emergency level.

D. Incorrect - A protected area evacuation is required since the C02 discharge leads to an Alert emergency classification.

N1-0P-21C, Sect. H.2 Technical Reference(s): EPMP-EPP-0101 (Attach if not previously provided)

EPI P-EPP-18 att 1 fig 1 Proposed References to be provided to applicants during examination: EAL Matrix Learning Objective: N1-286000-RBO-08 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam: Not used on the 2008 or 2009 exams Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 5 Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Comments:

Facility: Nine Nile Point Unit 1 Vendor: GE Exam Date: 2010 Exam Type: SRO ONLY Examination Outline Cross-reference: Level RO SRO Tier# 2 Group # 2 KIA # 201003 2.4.31 Importance Rating 4.1 Emergency Procedures I Plan: Knowledge of annunciator alarms, indications, or response procedures (CRD).

Proposed Question: SRO Question # 92 The plant is conducting a startup when the following occur:

Which of the following describes the status of control rod 18-19 and the required action per Technical Specifications?

This alarm signifies that the control rod is ...

A. NOT coupled. If the control rod cannot be re-coupled, it shall be completely inserted and valved out of service.

B. NOT coupled. If the control rod cannot be re-coupled, the Reactor shall be placed in hot shutdown within 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />.

C. coupled, but stuck at position 48. If the control rod cannot be moved, the Reactor shall be placed in a shutdown condition within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

D. coupled, but stuck at position 48. If the control rod cannot be moved, it shall be valved out of service and no control rods in the nine-rod array may be inoperable.

Proposed Answer: A Explanation (Optional):

A. Correct - This alarm signifies that the control rod is uncoupled. TS 3.1. 'l.b.(1) directs that if the rod cannot be re-coupled, it shall be fully inserted and valved out of service.

B. Incorrect - TS 3.1.1.b.(1} allows an uncoupled rod to be fully inserted and valved out of service. Only if this cannot be done does TS 3.1.1.f required hot shutdown in 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />.

C. Incorrect - The control rod is uncoupled.

D. Incorrect - The control rod is uncoupled.

ARP F3-1-6 Technical Reference(s}: OP-5, Sect. F.7 (Attach if not previously provided) 1.S.3.1.1 Proposed References to be provided to applicants during examination: TS 3.1.1 Learning Objective: N1-201001-RBO-11 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam: Not used on 2008 or 2009 exams Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 5 Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Comments:

Facility: Nine Mile Point Unit 1 Vendor: GE Exam Date: 2010 Exam Type: SRO ONLY Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 2 KIA # 234000 2.1.32 Importance Rating 4.0 Conduct of Operations: Ability to explain and apply all system limits and precautions (Fuel Handling Equipment).

Proposed Question: SRO Question # 93 The plant is shutdown for a refueling outage and core alterations are about to begin. The SRMs indicate the following:

  • SRM 11 - 47 cps
  • SRM 12 - 50 cps
  • SRM 13 - 0 cps, with the DN SCL OR INOP light ON
  • SRM 14 - 48 cps Which one of the following describes the resulting restriction on core alterations, in accordance with Technical Specifications and OP-34, Refueling Procedure?

Movement of (1 ) is restricted in (2) core quadrant(s).

(1 ) (2)

A. fuel only only the SW B. fuel and replacement of control rods only the SW C. fuel only the SW, NW and SE D. fuel and replacement of control rods the SW, NW and SE Proposed Answer: B Explanation (Optional):

A. Incorrect - There are restrictions on the movement of fuel AND control rods.

B. Correct - lAW OP-34 P & L 15 and T.S. sect 3.5.3.b; two SRMs shall be operable during core alterations, one in and one adjacent to each quadrant where fuel or control rods are being moved.

C. Incorrect - SRMs are operable in the NWand SE quadrants with SRMs operable in the adjacent quadrant (NE) so there are no restrictions in these two quadrants. There are restrictions on the movement of fuel AND control rods.

D. Incorrect - SRMs are operable in the NW and SE quadrants with SRMs operable in the adjacent quadrant (NE) so there are no restrictions in these two quadrants.

Technical Reference(s}: N1-0P-34, P & L 15 (Attach if not previously provided)

T.S. sect 3.5.3.b Proposed References to be provided to applicants during examination: None Learning Objective: N1-234000-RBO-14 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New x Question History: Last NRC Exam: Not used on 2008 or 2009 exam (2009

  1. 89 similar)

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 2 55.43 Facility operating limitations in the technical specifications and their bases.

Comments:

Facility: Nine Mile Point Unit 1 Vendor: GE Exam Date: 2010 Exam Type: SRO ONLY Examination Outline Cross-reference: Level RO SRO Tier # 3 Group # 1 KIA # G1 2.1.40 Importance Rating 3.9 Conduct of Operations: Knowledge of refueling administrative requirements Proposed Question: SRO Question # 94 The Reactor core is being reloaded with the following sequence of events:

  • A fuel assembly was latched and raised in the Spent Fuel Pool. All indications on the Refuel Bridge for lifting the fuel assembly were correct.
  • The fuel assembly was transferred from the Spent Fuel Pool to the Reactor core.
  • The fuel assembly is now fully raised above the location in which it will be placed.
  • The Control Room reports the Rod Block Monitor Panel REFUEL INTERLOCK indicator light is OFF.
  • The FUEL HOIST INTLK light on the Refuel Bridge is OFF.

Which one of the following describes how to proceed with fuel movements and the reason for continuing or stopping fuel movements, in accordance with FHP-25, General Description of Fuel Moves?

A. Stop fuel movement. The over-the-core limit switch failed to indicate the bridge was over the core.

B. Stop fuel movement. At least one control rod is not fully inserted, affecting the all control rods fully inserted signal.

C. Continue fuel movement. The rod block is not received until the main hoist is lowered from the normal-up position.

D. Continue fuel movement. The rod block is not received until the fuel assembly is approaching the core top guide.

Proposed Answer: A Explanation (Optional):

A. Correct - The rod block should be received when the bridge is loaded and over the reactor core even with the main hoist normal up. The limit switch failed to recognize the bridge over the reactor core. lAW FHP-25, Att. 4, the loss of any refuel bridge interlocks requires stopping fuel movement.

B. Incorrect - If one rod was out with the bridge over the reactor core and loaded, then a bridge reverse motion stop, fuel hoist motion block, and rod block would be received, not just a rod block. Additionally rod blocks, travel blocks, and fuel hoist interlocks are generated independent of the REFUEL INTERLOCK light and would be indicated by the FUEL HOIST INTLK light being ON.

C. Incorrect - Rod block failed to actuate because the over-the-core limit switch failed.

D. Incorrect - Rod block failed to actuate because the over-the-core limit switch failed.

  • IR f Techmca () N1-0P-34, Sect H.4.0 (Attach if not previously provided) e erence s: N1-FHP-25 , Att 4 Proposed References to be provided to applicants during examination: None Learning Objective: N1-234000-RBO-9 (As available)

ID: LC1 03-01 AUDIT Question Source: Bank #

SR018 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: Not used on 2008 or 2009 exams Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 55.43 7 - Fuel handling facilities and procedures.

Comments:

Facility: Nine Mile Point Unit 1 Vendor: GE Exam Date: 2010 Exam Type: SRO ONLY Examination Outline Cross-reference: Level RO SRO Tier # 3 Group # 2 KIA # G2 2.2.42 Importance Rating 4.6 Equipment Control: Ability to recognize system parameters that are entry-level conditions for Technical Specifications.

Proposed Question: SRO Question # 95 The plant is operating at 100% power with the following:

  • I&C has just discovered that the Liquid Poison (LP) system temperature indicating controller and alarm temperature switches have a calibration error
  • LP tank temperature is indicating 90°F
  • Actual LP tank temperature is 60°F
  • LP tank volume is 1341 gallons
  • LP solution concentration is 15.9%
  • LP solution boron-10 enrichment is 45.9%

Which one of the following identifies the operability of the Liquid Poison system and describes the required action?

A. Operable. Restore tank temperature per Section H of N1-0P-12.

B. Inoperable. Restore tank temperature or dilute the LP tank concentration to 12.5%.

C. Inoperable. Restore tank temperature or commence a normal shutdown within one hour.

D. Inoperable. Restore tank temperature or commence a rapid shutdown within one hour and be in cold shutdown within 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />.

Proposed Answer: C

Explanation (Optional):

A. Incorrect - LP system is inoperable. The system is considered inoperable from the time of the initial discovery that the temperature requirements of TS 3.1.2.d are not met.

B. Incorrect - This concentration meets the temperature requirements for TS 3.1.2.d, but is below the minimum concentration for TS 3.1.2.c.

C. Correct - Per Tech Spec 3.1.2.d, the minimum allowable temperature for 15.9%

concentration is -70oF. Liquid Poison is inoperable. Either temperature must be restored or a normal shutdown must be commenced within one hour per TS 3.1.2.e.

D. Incorrect - A rapid shutdown is not required.

Technical Reference(s): Tech Spec 3.1.2 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: Tech Spec 3.1.2 Learning Objective: N1-211000-RBO-14 (As available)

Question Source: Bank # Rx Pwr Sys #463 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: Not used on 2008 or 2009 exams Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 55.43 1 Conditions and limitations in the facility license Comments:

Facility: Nine Mile Point Unit 1 Vendor: GE Exam Date: 2010 Exam Type: SRO ONLY Examination Outline Cross-reference: Level RO SRO Tier # 3 Group # 3 KIA # G3 2.3.11 Importance Rating 4.3 Radiation Control: Ability to control radiation releases.

Proposed Question: SRO Question # 96 The plant is operating at 100% power with the following:

  • An Operator reports that the EQUIP FAIL light on the Service Water Rad Monitor is ON
  • Chemistry has been notified and determines that Service Water release rates are normal Which one of the following actions is required to allow Service Water operation to continue?

A. Verify the other Service Water Radiation Monitor is operable within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

B. Collect and analyze Service Water effluent grab samples at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

C. Sample Reactor Building and Turbine Building Service Water return lines alternately every 15 minutes.

D. Estimate the Service Water flow at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and collect and analyze effluent grab samples at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

Proposed Answer: B Explanation (Optional):

A. Incorrect - There is only the one SW radiation monitor.

B. Correct - lAW ODMC section 3.6.14, with the one Service Water System Effluent Line Monitor inoperable, take the actions specified in Table 3.6.14-1. With no channels

OPERABLE, effluent releases via this pathway may continue provided that, at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> during the release, grab samples are collected and analyzed.

C. Incorrect - Note (i) of the table "Monitoring will be conducted continuously by alternately sampling the reactor building and turbine building service water return lines for approximately 15-minute intervals." applies to the normal sampling during all modes of operation.

D. Incorrect - Pumps curves or rated capacity will be utilized to estimate flow for the Liquid Radwaste Effluent line only.

N1-ARP-H1 Technical Reference(s}: N1-0P-50B (Attach if not previously provided)

ODCM 3.6.14 Proposed References to be provided to applicants during examination: ODCM 3.6.14 Learning Objective: N 1-272000-RBO-14 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New x Question History: Last NRC Exam: Not used on 2008 or 2009 exams, similar to #50 on 2008 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 55.43 1 Conditions and limitations in the facility license Comments:

Facility: Nine Mile Point Unit 1 Vendor: GE Exam Date: 2010 Exam Type: SRO ONLY Examination Outline Cross-reference: Level RO SRO Tier # 3 Group # 4 KIA # G4 2.4.26 Importance Rating 3.6 Emergency Procedures I Plan: Knowledge of facility protection requirements, including fire brigade and portable fire fighting equipment usage.

Proposed Question: SRO Question # 97 The plant is operating at 100% power with the following:

  • The Fire Brigade Leader has been dispatched to respond to a fire alarm in the Powerboard 102 room
  • The Fire Brigade Leader reports:

o A fire is out of control in the Powerboard 102 room o The fire has spread into the Powerboard 103 room and the EDG 102 room

  • Control of EDG 102 has been lost Which one of the following actions is required per SOP-21.1, Fire in Plant?

A. Immediately scram the Reactor and enter SOP-1, Reactor Scram, only.

B. Immediately scram the Reactor and enter SOP-1, Reactor Scram and EOP-2, RPV Control.

C. If the fire is NOT under control within 15 minutes, scram the Reactor and enter SOP-1 ,

Reactor Scram, only.

D. If the fire is NOT under control within 15 minutes scram the Reactor and enter SOP-1, Reactor Scram and EOP-2, RPV Control.

Proposed Answer: B

Explanation (Optional):

A. Incorrect - SOP-21.1 directs the crew to enter SOP-1, Reactor Scram and EOP-2, RPV Control.

B. Correct - SOP-21.1 directs a reactor scram and entry into SOP-1 and EOP-2 when ANY of the following conditions exist due to fire:

  • Spurious valve operation,
  • Loss of equipment control,
  • Fire NOT under control within 15 minutes,
  • Fire endangers Safe Shutdown capability, In this case, the fire endangers safe shutdown capability (both 4160 KVemergency busses) and has resulted in the loss of equipment control (EDG 102), requiring an immediate scram and entry into EOP-2.

C. Incorrect - The scram cannot be delayed 15 minutes and SOP-21.1 directs the crew to enter SOP-1, Reactor Scram and EOP-2, RPV Control.

D. Incorrect - The scram cannot be delayed 15 minutes.

Technical Reference(s): SOP-21.1 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None Learning Objective: 01-0PS-006-342-1-01, EO-1.2 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam: Not used on the 2008 or 2009 exams Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 5 Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Comments:

Facility: Nine Mile Point Unit 1 Vendor: GE Exam Date: 2010 Exam Type: SRO ONLY Examination Outline Cross-reference: Level RO SRO Tier # 3 Group # 1 KJA# G1 2.1.13 Importance Rating 3.2 Conduct of Operations: Knowledge of facility requirements for controlling vitalI controlled access.

Proposed Question: SRO Question # 98 The plant is operating at 100% power with the following:

  • A credible insider threat has resulted in activation of the "Two Person Rule" in accordance with EPIP-EPP-10, Security Contingency Event
  • The Security investigation is ongoing, and the "Two Person Rule" will NOT be terminated for 2-3 hours
  • The operating CRD pump has tripped
  • An Operator requests an exception to the "Two Person Rule" so they may inspect the CRD pumps
  • There are no other Operators currently available Which one of the following describes the correct response to this request, in accordance with EPIP-EPP-10?

An exception to the "Two Person Rule" ...

A. may be approved by the Security Shift Supervisor.

B. may be approved by the Control Room Supervisor.

C. may NOT be approved, but a Security Officer may accompany the Operator.

D. may NOT be approved, and the Operator must wait until another Operator becomes available.

Proposed Answer: C Explanation (Optional):

A. Incorrect - There is no provision in EPIP-EPP-10 for an exception to the two person rule.

B. Incorrect - There is no provision in EPIP-EPP-10 for an exception to the two person rule.

C. Correct - The two man rule requires:

1. Equal task qualification levels are not necessary.
2. Partner must remain within line of sight the entire time you are in the vital area.
3. Partner must have access to the vital area.

The security guard meets the requirements for the two man rule.

D. Incorrect - The requirements for the two man rule are only that another person qualified to be in the vital area is within line of site. Equal qualifications are not required.

Technical Reference(s): EPIP-EPP-10 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None Learning Objective: 03-OPS-006-350-3-31, EO-1.3 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New x Question History: Last NRC Exam: not used on 2008 or 2009 exams (Same KIA as 2008, #94)

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41

55.43 5 Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Comments:

Facility: Nine Mile Point Unit 1 Vendor: GE Exam Date: 2010 Exam Type: SRO ONLY Examination Outline Cross-reference: Level RO SRO Tier # 3 Group # 4 KIA # G4 2.4.17 Importance Rating 4.3 Emergency Procedures I Plan: Knowledge of EOP terms and definitions.

Proposed Question: SRO Question # 99 The plant is experiencing a transient with the following:

  • Torus pressure is 20 psig and rising slowly
  • Torus level is 11 feet and rising slowly

Containment Spray is defined to be in service when ...

A. the first Containment Spray pump is running and up to rated flow. Next, an RPV Slowdown must be performed.

S. two Containment Spray pumps are running and up to rated flow. Next, an RPV Slowdown must be performed.

C. the first Containment Spray pump is running and up to rated flow. Next, an evaluation of Torus pressure in relation to the Pressure Suppression Pressure curve must be performed.

D. two Containment Spray pumps are running and up to rated flow. Next, an evaluation of Torus pressure in relation to the Pressure Suppression Pressure curve must be performed.

Proposed Answer: C

Explanation (Optional):

A. Incorrect - An RPV Blowdown will only be directed after evaluating Torus pressure in relation to the Pressure Suppression Pressure. If Containment Spray successfully lowers Torus pressure below the PSP limit before this evaluation is performed, a blowdown will not be required.

B. Incorrect - Containment Spray should be considered to be in service after the first pump has been started and is up to rated flow. An RPV Blowdown will only be directed after evaluating Torus pressure in relation to the Pressure Suppression Pressure. If Containment Spray successfully lowers Torus pressure below the PSP limit before this evaluation is performed, a blowdown will not be required.

C. Correct - When the Torus pressure is greater that the Pressure Suppression Pressure (PSP) upon entry into the Primary Containment Control EOP, all steps up to evaluating proximity to PSP are to be implemented prior to evaluating Torus Pressure against the PSP curve. Containment sprays are considered to be "in service" when one train of Containment Spray is initiated (The other additional loop of Containment Spray that is started is for Appendix J Water Seal requirements). Evaluation of the Torus pressure in relation to the Pressure Suppression Pressure (PSP) curve is expected to occur once Containment Sprays are "in service."

D. Incorrect - Containment Spray should be considered to be in service after the first pump has been started and is up to rated flow.

N1-EOP-4 Technical Reference(s): GAI-OPS-20 (Attach if not previously provided)

Primary Containment Press Proposed References to be provided to applicants during examination:

Leg of EOP-4 and the PSP Learning Objective: 01-0PS-006-344-1-04 EO-1.2 (As available)

NRC 2006 SRO 2 Question Source: Bank #

SYSI D: 21302 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: Not used on 2008 or 2009 exams

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 5 Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Comments:

Facility: Nine Mile Point Unit 1 Vendor: GE Exam Date: 2010 Exam Type: SRO ONLY Examination Outline Cross-reference: Level RO SRO Tier# 3 Group # 2 KIA # G2 2.2.40 Importance Rating 4.7 Equipment Control: Ability to apply technical specifications for a system.

Proposed Question: SRO Question # 100 A plant startup is in progress with the following:

  • The Reactor is critical and reached the point of adding heat at time 0715
  • The following heatup data has been recorded:

Time Reactor pressure (psig) Recirc loop suction (hhmm) temperature tF) 0730 0 98 0800 0 118 0830 0 169 0900 0 187 0930 0 211 1000 36 232 1030 53 264 1100 74 292 Which one of the following describes the Technical Specification implications of this heatup with respect to heatup rate and minimum Reactor vessel coolant temperature?

Minimum Reactor Vessel Heatup Rate Limit Coolant Temperature Limit A. Not Exceeded Not Exceeded B. Not Exceeded Exceeded C. Exceeded Not Exceeded

D. Exceeded Exceeded Proposed Answer: B Explanation (Optional):

A. Incorrect - Minimum reactor vessel coolant temperature curve is violated at time 0730 since the reactor is critical with temperature less than 100°F.

B. Correct - Heatup rate limit is not exceeded because no one hour temperature change is above 100°F. Minimum reactor vessel coolant temperature curve is violated at time 0730 since the reactor is critical with temperature less than 100°F.

C. Incorrect - Heatup rate limit is not exceeded because no one hour temperature change is above 100°F. Minimum reactor vessel coolant temperature curve is violated at time 0730 since the reactor is critical with temperature less than 100°F.

D. Incorrect - Heatup rate limit is not exceeded because no one hour temperature change is above 100°F.

Technical Reference(s): TS 3.2.1 and 3.2.2, PTLR (Attach if not previously provided)

TS 3.2.1 and 3.2.2, Proposed References to be provided to applicants during examination:

PTLR Learning Objective: N1-1 01 001-RBO-14 (As available)

ID: N1-101001-RBO-Question Source: Bank #

14-Q-01 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: Not used on 2008 or 2009 exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x

10 CFR Part 55 Content: 55.41 55.43 2 Facility operating limitations in the technical specifications and their bases.

Comments:

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